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Sample records for gamma radiation shieldings

  1. Concrete for. gamma. radiation shielding

    Energy Technology Data Exchange (ETDEWEB)

    de Azevedo e Souza, A.C. (Rio de Janeiro Univ. (Brazil). Inst. de Quimica); Rogers, J D [Rio de Janeiro Univ. (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia

    1980-06-01

    The attenuation characteristics of ..gamma.. radiation in concrete slabs, considering their mechanical resistence and densities were determined. One heavy concrete which was used, was prepared using as additives iron ore and Fe/sub 2/ O/sub 3/ pellets in various grain sizes. Fortran programs were used for analysing data and determining the absorption coefficients and attenuation factors.

  2. 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

  3. Dosimetry and Shielding of X and Gamma Radiation

    International Nuclear Information System (INIS)

    Oncescu, M.; Panaitescu, I.

    1992-01-01

    This book covers the following problems: 1. X and Gamma radiations, 2. Interaction of X-ray and gamma radiations with matter, 3. Interaction of electrons with matter, 4. Principles and basic concepts of dosimetry, 5. Ionization dosimetry, 6. Calorimetric chemical and photographic dosimetry, 7. Solid state dosimetry, 8. Computation of dosimetric quantities, 9. Dosimetry in radiation protection, 10. Shielding of X and gamma radiations. The authors, well-known Romanian experts in Radiation Physics and Engineering, gave an up-dated, complete and readable account of this subject matter. The analyses of physical principles and concepts, of materials and instruments and of computational methods and applications are all well balanced to meat the needs of a broad readership

  4. 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

  5. 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

  6. Gamma radiation shielding materials improved with burning resistance

    International Nuclear Information System (INIS)

    Nakamura, Michio; Nakamura, Ken-ichi; Yukawa, Katsunori.

    1985-01-01

    Purpose: To obtain gamma irradiation shielding materials excellent in workability and resistant to burning by using a two component type room temperature vulcanizing silicon rubber composition as the base material. Method: Silicon rubber comprising a diorganopolysiloxane polymer, an alkyl silicate as a crosslinker and a suitable sulfurdizing catalyst, for example, a carboxylate is mixed with iron powder and silicon oxide powder as reinforcing and flame retardant material and applied with molding. The iron powder and the silica rocks powder have grain size of 50 - 150 μm and 1 - 70 μm and charged by the amount of from 55 to 60 % by weight and from 20 to 25 % by weight respectively. The fluidizing property is impaired if the particle size of the silica rocks powder is less than 1 μm and, while on the other hand, no desired specific gravity of a predetermined value can be obtained for the molding product if the filled amount of the iron powder is less than 55 %. The oxygen index of the molding product is 45 to improve the burning resistance. The materials are excellent in the air-tightness, gamma radiation shielding performance, elasticity and workability required for the cable penetrations in a nuclear power plant and they generate noxious gases neither. (Kawakami, Y.)

  7. PKI, Gamma Radiation Reactor Shielding Calculation by Point-Kernel Method

    International Nuclear Information System (INIS)

    Li Chunhuai; Zhang Liwu; Zhang Yuqin; Zhang Chuanxu; Niu Xihua

    1990-01-01

    1 - Description of program or function: This code calculates radiation shielding problem of gamma-ray in geometric space. 2 - Method of solution: PKI uses a point kernel integration technique, describes radiation shielding geometric space by using geometric space configuration method and coordinate conversion, and makes use of calculation result of reactor primary shielding and flow regularity in loop system for coolant

  8. Radiation shielding

    International Nuclear Information System (INIS)

    Yue, D.D.

    1979-01-01

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

  9. Evaluation of the gamma radiation shielding parameters of bismuth modified quaternary glass system

    Science.gov (United States)

    Kaur, Parminder; Singh, K. J.; Thakur, Sonika

    2018-05-01

    Glasses modified with heavy metal oxides (HMO) are an interesting area of research in the field of gamma-ray shielding. Bismuth modified lithium-zinc-borate glasses have been studied whereby bismuth oxide is added from 0 to 50 mol%. The gamma ray shielding properties of the glasses were evaluated at photon energy 662 keV with the help of XMuDat computer program by using the Hubbell and Seltzer database. Various gamma ray shielding parameters such as attenuation coefficient, shield thickness in terms of half and tenth value layer, effective atomic number have been studied in this work. A useful comparison of this glass system has been made with standard radiation shielding concretes viz. ordinary, barite and iron concrete. The glass samples containing 20 to 50 mol% bismuth oxide have shown better gamma ray shielding properties and hence have the potential to become good radiation absorbers.

  10. Spectroscopic Study of Radiation around the Leksell Gamma Knife for Room Shielding Applications

    OpenAIRE

    Hubert, Alexis

    2017-01-01

    Any center planning to install a Gamma Knife radiosurgery unit has to provide for an efficient shielding of the treatment room, to protect the patient, the staff and the public, against undesired radiation. The shielding barrier design is controlled by national and international recommendations; the reference documents for gamma ray radiotherapy facilities are the National Council on Radiation Protection and Measurements (NCRP) reports 49 and 151. However, some facts highlighted in this thesi...

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  12. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

    Aigbosuria, E F [Department of Computer Electronics/Physics, Lead City University, Ibadan (Nigeria)

    2011-10-24

    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{sup -1}, 0.165cm{sup -1}, 0.163cm{sup -1}, 0.156cm{sup -1}, 0.149cm{sup -1}, 0.143cm{sup -1}, 0.133cm{sup -1}, 0.132cm{sup -1}, 0.127cm{sup -1}, 0.124cm{sup -1}, 0.085cm{sup -1}, 0.123cm{sup -1}, 0.122cm{sup -1}, 0.113cm{sup -1}, 0.101cm{sup -1}, 0.088cm{sup -1}, 0.087cm{sup -1}, 0.086cm{sup -1}, 0.082cm{sup -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

  14. Radiation shielding lead shield

    International Nuclear Information System (INIS)

    Dei, Shoichi.

    1991-01-01

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

  15. Durability and shielding performance of borated Ceramicrete coatings in beta and gamma radiation fields

    Energy Technology Data Exchange (ETDEWEB)

    Wagh, Arun S., E-mail: asw@anl.gov [Environmental Science Division, Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, IL 60439 (United States); Sayenko, S.Yu.; Dovbnya, A.N.; Shkuropatenko, V.A.; Tarasov, R.V.; Rybka, A.V.; Zakharchenko, A.A. [National Science Center, Kharkov Institute of Physics and Technology, Kharkov (Ukraine)

    2015-07-15

    Highlights: • It incorporates all suggestions by the reviewers. • Explanation to each new term is provided and suitable references are given. • Sample identities have been streamlined by revising the text and the tables. • Some figures have been redrawn. - Abstract: Ceramicrete™, a chemically bonded phosphate ceramic, was developed for nuclear waste immobilization and nuclear radiation shielding. Ceramicrete products are fabricated by an acid–base reaction between magnesium oxide and mono potassium phosphate. Fillers are used to impart desired properties to the product. Ceramicrete’s tailored compositions have resulted in several commercial structural products, including corrosion- and fire-protection coatings. Their borated version, called Borobond™, has been studied for its neutron shielding capabilities and is being used in structures built for storage of nuclear materials. This investigation assesses the durability and shielding performance of borated Ceramicrete coatings when exposed to gamma and beta radiations to predict the composition needed for optimal shielding performance in a realistic nuclear radiation field. Investigations were conducted using experimental data coupled with predictive Monte Carlo computer model. The results show that it is possible to produce products for simultaneous shielding of all three types of nuclear radiations, viz., neutrons, gamma-, and beta-rays. Additionally, because sprayable Ceramicrete coatings exhibit excellent corrosion- and fire-protection characteristics on steel, this research also establishes an opportunity to produce thick coatings to enhance the shielding performance of corrosion and fire protection coatings for use in high radiation environment in nuclear industry.

  16. Radiation Build-Up Of High Energy Gamma In Shielding Of High Atomic Number

    International Nuclear Information System (INIS)

    Yuliati, Helfi; Akhadi, Mukhlis

    2000-01-01

    Research to observe effect of radiation build-up factor (b) in iron (Fe) and lead (Pb) for high energy gamma shielding from exp.137 Cs (E gamma : 662 keV) and exp.60 Co (E gamma : 1332 keV) sources has been carried out. Research was conducted bt counting of radiation intensity behind shielding with its thickness vary from 1 to 5 times of half value thickness (HVT). NaI (TI) detector which connected to multi channel analyzer (MCA) was used for the counting. Calculation result show that all of b value are near to 1 (b∼1) both for Fe and Pb. Without inserting b in calculation, from the experiment it was obtained HVT value of Fe for high gamma radiation of 662 and 1332 keV were : (12,94 n 0,03) mm and (17,33 n 0,01) mm with their deviation standards were 0,2% and 0,06% respectively. Value of HVT for Pb with the same energy were : (6,31 n 0,03) mm and (11,86 n 0,03) mm with their deviation standars were : 0,48% and 0,25% respectively. HVL concept could be applied directly to estimate shielding thickness of high atomic number of high energy gamma radiation, without inserting correction of radiation build-up factor

  17. 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

  18. 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

  19. Effects of increased shielding on gamma-radiation levels within spacecraft

    Science.gov (United States)

    Haskins, P. S.; McKisson, J. E.; Weisenberger, A. G.; Ely, D. W.; Ballard, T. A.; Dyer, C. S.; Truscott, P. R.; Piercey, R. B.; Ramayya, A. V.; Camp, D. C.

    The Shuttle Activation Monitor (SAM) experiment was flown on the Space Shuttle Columbia (STS-28) from 8 - 13 August, 1989 in a 57°, 300 km orbit. One objective of the SAM experiment was to determine the relative effect of different amounts of shielding on the gamma-ray backgrounds measured with similarly configured sodium iodide (NaI) and bismuth germante (BGO) detectors. To achieve this objective twenty-four hours of data were taken with each detector in the middeck of the Shuttle on the ceiling of the airlock (a high-shielding location) as well as on the sleep station wall (a low-shielding location). For the cosmic-ray induced background the results indicate an increased overall count rate in the 0.2 to 10 MeV energy range at the more highly shielded location, while in regions of trapped radiation the low shielding configuration gives higher rates at the low energy end of the spectrum.

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  1. A study of gamma shielding

    International Nuclear Information System (INIS)

    Roogtanakait, N.

    1981-01-01

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

  2. 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.

  3. Radiation shielding

    International Nuclear Information System (INIS)

    Aitken, D.

    1979-01-01

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

  4. Radiation Build-Up In Shielding Of Low Activity High Energia Gamma Source

    International Nuclear Information System (INIS)

    Helfi-Yuliati; Mukhlis-Akhadi

    2003-01-01

    Research to observe radiation build-up factor (b) in aluminium (Al), iron (Fe) and lead (Pb) for shielding of gamma radiation of high energy from 137 cs (E γ : 662 keV) source and 60 Co (E γ : 1332 keV) of low activity sources has been carried out. Al with Z =13 represent metal of low atomic number, Fe with Z =26 represent metal of medium atomic number, and Pb with Z = 82 represent metal of high atomic number. Low activity source in this research is source which if its dose rate decrease to 3 % of its initial dose rate became safe for the workers. Research was conducted by counting of radiation intensity behind shielding with its thickness vary from 1 to 5 times of half value thickness (HVT). NaI(TI) detector which connected to multi channel analyzer (MCA) was used for the counting. Calculation result show that all of b value are close to 1 (b ∼ 1) for all kinds of metals. No radiation build-up factor is required in estimating the shielding thickness from several kinds of metals for low activity of high energy gamma source. (author)

  5. Study of local Agregate for Gamma radiation concrete shield; Studi pemakaian Agregat lokal pada pembuatan beton perisai radiasi Gamma

    Energy Technology Data Exchange (ETDEWEB)

    Tochrul-Binowo,; Endro-Kismolo,; Darsono, [Yogyakarta Nuclear Research Centre, National Atomic Energy Agency, Yogyakarta (Indonesia)

    1996-04-15

    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 {mu} barite concrete = 0.23071 cm{sup -1}, {mu} manganese concrete = 0.08401 cm{sup -1} and {mu} normal concrete = 0.1669 cm{sup -1}.

  6. 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...... 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...

  7. GARLIC, a shielding program for GAmma Radiation from Line- and Cylinder-sources

    Energy Technology Data Exchange (ETDEWEB)

    Roos, Matts

    1959-07-15

    GARLIC is a program for computing the gamma ray flux or dose rate at a shielded idotropic point detector, due to a line source or the line equivalent of a cylindrical source. The source strength distribution along the line must be either uniform or an arbitrary part of the positive half-cycle of a cosine function. The line source can be oriented arbitrarily with respect to the main shield and the detector, except that the detector must not be located on the line source or on its extension. The main source is a homogeneous plane slab in which scattered radiation is accounted for by multiplying each point element of the line source by a point source build-up factor inside the integral over the point elements. Between, the main shield and the line source additional shields can be introduced, which are either plane slabs, parallel to the main shield, or cylindrical rings, coaxial with the line source. Scattered radiation in the additional shields can only be accounted for by constant build-up factors outside the integral. GARLIC-xyz is an extended version particularly suited for the frequently met problem of shielding a room containing a large number of line sources in different positions. The program computes the angles and linear dimensions of a problem for GARLIC when the positions of the detector point and the end points of the line source are given as points in an arbitrary rectangular coordinate system. As an example the isodose curves in water are presented for a monoenergetic cosine-distributed line source at several source energies and for an operating fuel element of the Swedish reactor R3.

  8. 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)

  9. Study of gamma radiation shielding properties of ZnO-TeO_2 glasses

    International Nuclear Information System (INIS)

    Issa, Shama A.M.; Sayyed, M.I.; Kurudirek, Murat

    2017-01-01

    Mass attenuation coefficient (μm), half value layer (HVL) and mean free path (MFP) for xZnO-(100-x)TeO_2, where x=10, 15, 20, 25, 30, 35 and 40 mol%, have been measured for 0.662, 1.173 and 1.33 MeV photons emitted from "1"3"7Cs and "6"0Co using a 3 x 3 inch NaI (Tl) detector. Some relevant parameters such as effective atomic numbers (Z_e_f_f) and electron densities (Nel) of glass samples have been also calculated in the photon energy range of 0.015-15 MeV. Moreover, gamma-ray energy absorption buildup factor (EABF) and exposure buildup factor (EBF) were estimated using a five-parameter Geometric Progression (GP) fitting approximation, for penetration depths up to 40 MFP and in the energy range 0.015-15 MeV. The measured mass attenuation coefficients were found to agree satisfactorily with the theoretical values obtained through WinXcom. Effective atomic numbers (Z_e_f_f) and electron densities (N_e_l) were found to be the highest for 40ZnO-60TeO_2 glass in the energy range 0.04-0.2 MeV. The 10ZnO-90TeO_2 glass sample has lower values of gamma-ray EBFs in the intermediate energy region. The reported new data on radiation shielding characteristics of zinc tellurite glasses should be beneficial from the point of proper gamma shield designs when intended to be used as radiation shields. (author)

  10. Crystal glass used for X ray and gamma radiation shielding - Part two

    International Nuclear Information System (INIS)

    Antonio Filho, Joao

    2007-01-01

    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. However, properties of the radiation attenuation of crystal glass commercially available in Brazil, for the different types of energy are not known. For this reason, this work was carried out aiming to determine the radiation attenuation, transmission curves and Half Value Layer. In this work, ten plates of crystal glass, with dimensions of 20 cm x 20 cm and range of thicknesses from 0.5 to 2.0 cm, were used. The plates were X-ray irradiated with potential constants of 60, 80, 110, 150 kV and gamma radiation of 60 Co. Analysis in the properties of the 60 Co radiation attenuation of barite plaster and barite concrete commercially available in Brazil were also carried out. 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. The thickness equivalent of a half value layer and deci value layer of crystal glass for all types of radiation and energies studied was also determined. (author)

  11. Shielding effect of snow cover on indoor exposure due to terrestrial gamma radiation

    International Nuclear Information System (INIS)

    Fujimoto, Kenzo; Kobayashi, Sadayoshi

    1988-01-01

    Many people in the world live in high latitude region where it snows frequently in winter. When snow covers the ground, it considerably reduces the external exposure from the radiation sources in the ground. Therefore, the evaluation of snow effect on exposure due to terrestrial gamma radiation is necessary to obtain the population dose as well as the absorbed dose in air in snowy regions. Especially the shielding effect on indoor exposure is essentially important in the assessment of population dose since most individuals spend a large portion of their time indoors. The snow effect, however, has been rather neglected or assumed to be the same both indoors and outdoors in the population dose calculation. Snow has been recognized only as a cause of temporal variation of outdoor exposure rate due firstly to radon daughters deposition with snow fall and secondly to the shielding effect of snow cover. This paper describes an approach to the evaluation of shielding effect of snow cover on exposure and introduces population dose calculation as numerical example for the people who live in wooden houses in Japan

  12. Crystal glass and barite used for x ray and gamma radiation shielding

    International Nuclear Information System (INIS)

    Antonio Filho, Joao

    2008-01-01

    Full text: Crystal glass, barite plaster and barite concrete has been widely used as shielding material in gamma radiation sources as well as x-ray generating equipment to replace the plumbiferous glass and in the wall covering, in order to minimize exposure to individuals. However, properties of the radiation attenuation of crystal glass commercially available in Brazil, for the different types of energy are not known. For this reason, this work was carried out aiming to determine the radiation attenuation, transmission curves and Half Value Layer. In this work, ten plates of crystal glass, with dimensions of 20 cm x 20 cm and range of thicknesses from 0.5 to 2.0 cm, and ten plates of barite plaster and five plates of barite concrete, with dimensions of 20 x 20 cm 2 and range of thicknesses from 1,0 to 5,0 cm, were used. The plates were X-ray irradiated with potential constants of 60, 80, 110, 150 kV and gamma radiation of 60 Co. Analysis in the properties of the 60 Co radiation attenuation of barite plaster and barite concrete commercially available in Brazil were also carried out. The curves of attenuation and of transmission were obtained for crystal glass, barite plaster and barite concrete (mGy/m A.min) at 1 meter as a function of thickness. The thickness equivalent of a half value layer and deci value layer of crystal glass for all types of radiation and energies studied was also determined. Although their use permits the dimensioning of the armor covering for external x-radiation whit precision and safety without elevating the cost of protection. (author)

  13. A Monte Carlo Method for the Analysis of Gamma Radiation Transport from Distributed Sources in Laminated Shields

    International Nuclear Information System (INIS)

    Leimdoerfer, M.

    1964-02-01

    A description is given of a method for calculating the penetration and energy deposition of gamma radiation, based on Monte Carlo techniques. The essential feature is the application of the exponential transformation to promote the transport of penetrating quanta and to balance the steep spatial variations of the source distributions which appear in secondary gamma emission problems. The estimated statistical errors in a number of sample problems, involving concrete shields with thicknesses up to 500 cm, are shown to be quite favorable, even at relatively short computing times. A practical reactor shielding problem is also shown and the predictions compared with measurements

  14. A Monte Carlo Method for the Analysis of Gamma Radiation Transport from Distributed Sources in Laminated Shields

    Energy Technology Data Exchange (ETDEWEB)

    Leimdoerfer, M

    1964-02-15

    A description is given of a method for calculating the penetration and energy deposition of gamma radiation, based on Monte Carlo techniques. The essential feature is the application of the exponential transformation to promote the transport of penetrating quanta and to balance the steep spatial variations of the source distributions which appear in secondary gamma emission problems. The estimated statistical errors in a number of sample problems, involving concrete shields with thicknesses up to 500 cm, are shown to be quite favorable, even at relatively short computing times. A practical reactor shielding problem is also shown and the predictions compared with measurements.

  15. 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)

  16. Radiation shielding analysis

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  17. Optical absorption and gamma-radiation-shielding parameter studies of Tm3+-doped multicomponent borosilicate glasses

    Science.gov (United States)

    Lakshminarayana, G.; Sayyed, M. I.; Baki, S. O.; Lira, A.; Dong, M. G.; Kaky, Kawa M.; Kityk, I. V.; Mahdi, M. A.

    2018-05-01

    Different concentrations (0.1‒2.0 mol%) of Tm3+-doped multicomponent borosilicate glasses with 10 mol% Li2O (alkali) or MgO (alkaline) have been synthesized and their optical absorption and radiation shielding features were studied. For both Li2O and MgO series 0.5 mol% Tm3+-doped glass samples, the evaluated Ωλ ( λ = 2, 4, and 6) Judd-Ofelt (JO) intensity parameters from experimental oscillator strengths were used in estimating the radiative transition probabilities ( A R), branching ratios ( β R), and radiative lifetimes ( τ R) for several emission transitions. Using the XCOM software, the mass attenuation coefficients ( µ/ ρ) for all the fabricated glasses were evaluated within the 0.015‒10 MeV energy range. Also, the ( µ/ ρ) values were calculated at 0.356, 0.662, 1.173, and 1.33 MeV photon energies by MCNP5 simulation code and the results were compared with those obtained by XCOM. The ( µ/ ρ) values for Li2O, as well as MgO series glasses, increase with the addition of Tm2O3 and these values for MgO series glasses are slightly higher with respect to Li2O series glasses. From the ( µ/ ρ) values, effective atomic number ( Z eff), half-value layer (HVL), and mean free path (MFP) were calculated and the HVL and MFP results revealed that high-energy photons have more penetration into a glass sample compared to low-energy photons. Further, geometric progression (GP) fitting method was utilized to calculate the exposure buildup factor (EBF) within the 0.015‒15 MeV energy range. The 2.0 mol% Tm2O3-doped glasses show a better ability to attenuate gamma-rays in comparison to other glass samples, so the addition of Tm2O3 content leads to improvement of the shielding efficiency of the prepared glasses.

  18. 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)

  19. 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.)

  20. EFFECTS OF INTERFACES ON GAMMA SHIELDING

    Energy Technology Data Exchange (ETDEWEB)

    Clifford, C. E.

    1963-06-15

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

  1. 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.

  2. 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

  3. Radiation shielding plate

    International Nuclear Information System (INIS)

    Kobayashi, Torakichi; Sugawara, Takeo.

    1983-01-01

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

  4. Radiation shielding quality assurance

    Science.gov (United States)

    Um, Dallsun

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

  5. Shielding effect of clinical x-ray protector and lead glass against annihilation radiation and gamma rays of 99mTc

    International Nuclear Information System (INIS)

    Fukuda, Atsushi; Takahashi, Masaaki; Kitabayashi, Keitarou; Koshida, Kichiro; Matsubara, Kousuke; Noto, Kimiya; Nakagawa, Hiroto; Kawabata, Chikako

    2004-01-01

    Various pharmaceutical companies in Japan are making radioactive drugs available for positron emission tomography (PET) in hospitals without a cyclotron. With the distribution of these drugs to hospitals, medical check-ups and examinations using PET are expected to increase. However, the safety guidelines for radiation in the new deployment of PET have not been adequately improved. Therefore, we measured the shielding effect of a clinical X-ray protector and lead glass against annihilation radiation and gamma rays of 99m Tc. We then calculated the shielding effect of a 0.25 mm lead protector, 1 mm lead, and lead glass using the EGS4 (Electron Gamma Shower Version 4) code. The shielding effects of 22-mm lead glass against annihilation radiation and gamma rays of 99m Tc were approximately 31.5% and 93.3%, respectively. The clinical X-ray protector against annihilation radiation approximately doubled the skin-absorbed dose. (author)

  6. [Shielding effect of clinical X-ray protector and lead glass against annihilation radiation and gamma rays of 99mTc].

    Science.gov (United States)

    Fukuda, Atsushi; Koshida, Kichiro; Yamaguchi, Ichiro; Takahashi, Masaaki; Kitabayashi, Keitarou; Matsubara, Kousuke; Noto, Kimiya; Kawabata, Chikako; Nakagawa, Hiroto

    2004-12-01

    Various pharmaceutical companies in Japan are making radioactive drugs available for positron emission tomography (PET) in hospitals without a cyclotron. With the distribution of these drugs to hospitals, medical check-ups and examinations using PET are expected to increase. However, the safety guidelines for radiation in the new deployment of PET have not been adequately improved. Therefore, we measured the shielding effect of a clinical X-ray protector and lead glass against annihilation radiation and gamma rays of (99m)Tc. We then calculated the shielding effect of a 0.25 mm lead protector, 1 mm lead, and lead glass using the EGS4 (Electron Gamma Shower Version 4) code. The shielding effects of 22-mm lead glass against annihilation radiation and gamma rays of (99m)Tc were approximately 31.5% and 93.3%, respectively. The clinical X-ray protector against annihilation radiation approximately doubled the skin-absorbed dose.

  7. Gamma radiation, an aid to geologic mapping on the Arabian shield, Kingdom of Saudi Arabia

    International Nuclear Information System (INIS)

    Flanigan, V.J.

    1972-01-01

    Aerial gamma-radiation measurements in the Jabal Al Qarah quadrangle, Saudi Arabia, correlate with and complement magnetic data in distinguishing common Precambrian rock types. Areas of lower total-count gamma radiation correlate with areas of more intense magnetic patterns, which in turn correlate with areas of mafic rocks, as suggested from the study of the geophysical and geologic data of the Jabal Ishmas and Jabal Yafikh quadrangles adjacent to the north. In contrast, areas that reflect a lower magnetic intensity tend to show considerable variation in the radiation intensity and can be interpreted as being underlain by granitic rocks. On the basis of extrapolation of geophysical-geologic relationships established previously, selected radiation levels may be used to identify mappable rock units. Thus, radioactivity levels of 2,000 to 4,000 cpM suggest mafic rocks, levels of 4,000 to 6,000 cpM represent metavolcanic and metasedimentary rocks, levels of 7,000 to 10,000 cpM are representative of granodiorite gneiss, and levels of more than 11,000 cpM typify granitic rocks. The spectral gamma-radiation data are used to evaluate total-count anomalies and indicators of geologic processes of enrichment, and estimating the amount of isotopes of U, Th, and K within each of the lithologic units. (U.S.)

  8. Radiation shielding concrete

    International Nuclear Information System (INIS)

    Kunishima, Shigeru.

    1990-01-01

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

  9. 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.

  10. Radiation shielding curtain

    International Nuclear Information System (INIS)

    Winkler, N.T.

    1976-01-01

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

  11. Concrete radiation shielding

    International Nuclear Information System (INIS)

    Kaplan, M.F.

    1989-01-01

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

  12. 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 .

  13. Radiation shielding bricks

    International Nuclear Information System (INIS)

    Crowe, G.J.W.

    1983-01-01

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

  14. 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)

  15. Radiation shielding cloth

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  16. Radiation shielding material

    International Nuclear Information System (INIS)

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

    1976-01-01

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

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

    Indian Academy of Sciences (India)

    2017-07-25

    Jul 25, 2017 ... as industry, agriculture and medicine. It is very important ...... prominent reaction between the studied glass samples barriers and gamma rays is the ..... buildup factors for engineering materials. [20] Singh V P, Badiger N M ...

  18. Radiation shielding wall structure

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  19. 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.)

  20. Radiation shielding glass

    International Nuclear Information System (INIS)

    Kido, Kazuhiro; Ueda, Hajime.

    1997-01-01

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

  1. Highly heat removing radiation shielding material

    International Nuclear Information System (INIS)

    Asano, Norio; Hozumi, Masahiro.

    1990-01-01

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

  2. Multilayer radiation shield

    Science.gov (United States)

    Urbahn, John Arthur; Laskaris, Evangelos Trifon

    2009-06-16

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

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

    International Nuclear Information System (INIS)

    Matsumoto, Seiichiro; Aoyama, Saburo; Tashiro, Shingo; Nagai, Shiro

    1984-06-01

    The Waste Safety Testing Facility (WASTEF) was planned in 1978 to test the safety performance of HLW vitrified forms under the simulated conditions of long term storage and disposal, and completed in August 1981. The designed feature of the facility is to treat the vitrified forms contain actual high-level wastes of 5 x 10 4 Ci in maximum with 5 units of concrete shilded hot cells (3 units : Bate-Gamma cells, 2 units : Alpha-Gamma cells) and one units of Alpha-Gamma lead shielded cell, and to store radioactivity of 10 6 Ci in maximum. The safety performance of this facility is fundamentally maintained with confinement of radioactivity and shielding of the radiation. This report describes the method of gamma-ray shielding design, evaluation of the shielding test performed by using sealded gamma-ray sources(Co-60). (author)

  4. 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.

  5. 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...

  6. 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)

  7. 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)

  8. 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 ...

  9. 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

  10. Delayed radiation injury of gut-exposed and gut-shielded mice. I. The decrement in resistance to continuous gamma-ray stress

    International Nuclear Information System (INIS)

    Spalding, J.F.; Archuleta, R.F.; London, J.E.; Prine, J.R.

    1977-02-01

    Two mouse strains (RF/J and C57B1/6J) were exposed to x-ray doses totaling 400, 800, or 1200 rad. Total doses were given in 200-rad fractions at 7-day intervals to the whole body, gut only, or gut shielded. Animals treated as above (conditioned) were divided into 2 groups to form a two-part investigation. X-ray-conditioned and control mice were subjected to a continuous gamma-ray stress (challenge exposure) 28 days after the last x-ray dose. Delayed injury was measured as a reduction in mean after-survival (MAS) time and was observed in whole-body, gut-conditioned, and gut-shielded groups. The cause of death was attributed to hemopoietic hypoplasia in all groups. MAS reduction in all conditioned groups in both strains was linear with dose within the dose range used. Delayed injury per volume dose (measured as a reduction in MAS) was independent of the tissue initially conditioned with an acute dose of x rays. Thus, delayed injury per unit weight of gut tissue exposed was equal to that of either whole-body or gut-shielded radiation injury. Comparative weight loss observations during the continuous gamma-ray challenge exposure revealed a decrement in metabolic processes associated with body weight maintenance. This decrement was seen in all x-ray-conditioned groups

  11. Radiation shield for nuclear reactors

    International Nuclear Information System (INIS)

    Weissenfluh, J.A.

    1978-01-01

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

  12. Design experience: CRBRP radiation shielding

    International Nuclear Information System (INIS)

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

    1978-11-01

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

  13. Radiation shielding for neutron guides

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  14. Radiation shielding for neutron guides

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  15. 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.)

  16. Radiation shielding for fusion reactors

    International Nuclear Information System (INIS)

    Santoro, R.T.

    2000-01-01

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

  17. 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...

  18. EBT-P gamma-ray-shielding analysis

    International Nuclear Information System (INIS)

    Gohar, Y.

    1983-01-01

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

  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. Shielding walls against ionizing radiation

    International Nuclear Information System (INIS)

    1993-05-01

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

  3. Concrete mix design for X-and gamma shielding

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  4. Radiation shield for nuclear reactors

    International Nuclear Information System (INIS)

    Weissenfluh, J.A.

    1980-01-01

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

  5. 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

  6. Radiation shielding apparatus

    International Nuclear Information System (INIS)

    McCullagh, R.J.

    1977-01-01

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

  7. Radiation Attenuation and Stability of ClearView Radiation Shielding TM-A Transparent Liquid High Radiation Shield.

    Science.gov (United States)

    Bakshi, Jayeesh

    2018-04-01

    Radiation exposure is a limiting factor to work in sensitive environments seen in nuclear power and test reactors, medical isotope production facilities, spent fuel handling, etc. The established choice for high radiation shielding is lead (Pb), which is toxic, heavy, and abidance by RoHS. Concrete, leaded (Pb) bricks are used as construction materials in nuclear facilities, vaults, and hot cells for radioisotope production. Existing transparent shielding such as leaded glass provides minimal shielding attenuation in radiotherapy procedures, which in some cases is not sufficient. To make working in radioactive environments more practicable while resolving the lead (Pb) issue, a transparent, lightweight, liquid, and lead-free high radiation shield-ClearView Radiation Shielding-(Radium Incorporated, 463 Dinwiddie Ave, Waynesboro, VA). was developed. This paper presents the motivation for developing ClearView, characterization of certain aspects of its use and performance, and its specific attenuation testing. Gamma attenuation testing was done using a 1.11 × 10 Bq Co source and ANSI/HPS-N 13.11 standard. Transparency with increasing thickness, time stability of liquid state, measurements of physical properties, and performance in freezing temperatures are reported. This paper also presents a comparison of ClearView with existing radiation shields. Excerpts from LaSalle nuclear power plant are included, giving additional validation. Results demonstrated and strengthened the expected performance of ClearView as a radiation shield. Due to the proprietary nature of the work, some information is withheld.

  8. 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.

  9. Radiation shielding for fission reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-03-01

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

  10. Radiation shielding in dental radiography

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  11. 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

  12. Radiation-shielding transparent material

    International Nuclear Information System (INIS)

    Kusumeki, Asao.

    1983-01-01

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

  13. 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

  14. 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

  15. 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.

  16. 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…

  17. Radiation shielding and safety design

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-15

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

  18. Evaluation of radiation-shielding properties of the composite material

    International Nuclear Information System (INIS)

    Pavlenko, V.I.; Chekashina, N.I.; Yastrebinskij, R.N.; Sokolenko, I.V.; Noskov, A.V.

    2016-01-01

    The paper presents the evaluation of radiation-shielding properties of composite materials with respect to gamma-radiation. As a binder for the synthesis of radiation-shielding composites we used lead boronsilicate glass matrix. As filler we used nanotubular chrysotile filled with lead tungstate PbWO4. It is shown that all the developed composites have good physical-mechanical characteristics, such as compressive strength, thermal stability and can be used as structural materials. On the basis of theoretical calculation we described the graphs of the gamma-quanta linear attenuation coefficient depending on the emitted energy for all investigated composites. We founded high radiation-shielding properties of all the composites on the basis of theoretical and experimental data compared to materials conventionally used in the nuclear industry - iron, concrete, etc

  19. Radiation Shielding Properties of Some Marbles in Turkey

    International Nuclear Information System (INIS)

    Guenoglu, K.; Akkurt, I.

    2011-01-01

    Especially after development of technology, radiation started to be used in a large fields such as medicine, industry and energy. Using radiation in those fields bring hazardous effect of radiation into humancell. Thus radiation protection becomes important in physics. Although there are three ways for radiation protection, shielding of the radiation is the most commonly used method. Natural Stones such as marble is used as construction material especially in critical building and thus its radiation shielding capability should be determined.In this study, gamma ray shielding properties of some different types of marble mined in Turkey, have been measured using a NaI(Tl) scintillator detector. The measured results were also compared with the theoretical calculations XCOM.

  20. 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.

  1. 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.

  2. Shielding properties of the ordinary concrete loaded with micro- and nano-particles against neutron and gamma radiations.

    Science.gov (United States)

    Mesbahi, Asghar; Ghiasi, Hosein

    2018-06-01

    The shielding properties of ordinary concrete doped with some micro and nano scaled materials were studied in the current study. Narrow beam geometry was simulated using MCNPX Monte Carlo code and the mass attenuation coefficient of ordinary concrete doped with PbO 2 , Fe 2 O 3 , WO 3 and H 4 B (Boronium) in both nano and micro scales was calculated for photon and neutron beams. Mono-energetic beams of neutrons (100-3000 keV) and photons (142-1250 keV) were used for calculations. The concrete doped with nano-sized particles showed higher neutron removal cross section (7%) and photon attenuation coefficient (8%) relative to micro-particles. Application of nano-sized material in the composition of new concretes for dual protection against neutrons and photons are recommended. For further studies, the calculation of attenuation coefficients of these nano-concretes against higher energies of neutrons and photons and different particles are suggested. Copyright © 2018 Elsevier Ltd. All rights reserved.

  3. Radiation Shielding Properties of Some Marbles in Turkey

    Science.gov (United States)

    Günoǧlu, K.; Akkurt, I.

    2011-12-01

    Especially after development of technology, radiation started to be used in a large fields such as medicine, industry and energy. Using radiation in those fields bring hazordous effect of radition into humancell. Thus radiation protection becomes important in physics. Although there are three ways for radiation protection, shielding of the radiation is the most commonly used method. Natural Stones such as marble is used as construction material especially in critical building and thus its radiation shielding capability should be determined. In this study, gamma ray shielding properties of some different types of marble mined in Turkey, have been measured using a NaI(Tl) scintillator detector. The measured results were also compared with the theoretical calculations XCOM.

  4. 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)

  5. Shielding of the GERDA experiment against external gamma background

    International Nuclear Information System (INIS)

    Barabanov, I.; Bezrukov, L.; Demidova, E.; Gurentsov, V.; Kianovsky, S.; Knoepfle, K.T.; Kornouhkov, V.; Schwingenheuer, B.; Vasenko, A.

    2009-01-01

    The GERmanium Detector Array (GERDA) experiment will search for neutrinoless double beta decay of 76 Ge and is currently under construction at the INFN Laboratori Nazionali del Gran Sasso (LNGS) in Italy. The basic design of GERDA is the use of cryogenic liquid and water of high purity as a superior shield against the hitherto dominant background from external gamma radiation. In this paper we show by Monte Carlo simulations and analytical calculations how GERDA was designed to suppress this background at Q ββ ( 76 Ge)=2039keV to a level of about 10 -4 cts/(keVkgy).

  6. 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.

  7. 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)

  8. Shielding of gamma field in residential houses

    Energy Technology Data Exchange (ETDEWEB)

    Smejkal, Z; Pavlata, M [Univ. Pardubice, (Czech Republic); Pokorna, I; Urban, M [Institute of CO CR, 53341 Lazne Bohdanec (Czech Republic)

    1996-12-31

    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. (Abstract Truncated)

  9. 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)

  10. 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

  11. 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.

  12. 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

  13. Radiation shielding application of lead glass

    International Nuclear Information System (INIS)

    Nathuram, R.

    2017-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

  15. Radiation shielding calculation using MCNP

    International Nuclear Information System (INIS)

    Masukawa, Fumihiro

    2001-01-01

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

  16. Glasses impregnated with lead for radiation shielding

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  17. 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

  18. 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.

  19. Flexible shielding material sheet for radiations

    International Nuclear Information System (INIS)

    Kokan, Susumu; Fukuoka, Masasuke.

    1976-01-01

    Object: To provide a soft sheet of shielding material for radioactive rays without involving no problem such as environmental contamination, without generating intense second radioactive rays such as conventional cadmium. Structure: 100 weight parts of boron compound (boron carbide, boric acid anhydride) and 5 to 60 weight parts of low molecular-weight polyethylene resin, of which average molecular weight is less than 8000, are agitated in a mixer and during agitation are increased in temperature to a level above a softening temperature of the polyethylene resin to obtain a mixture in which the boron compound is coated with the low molecular-weight polyethylene. Next, 3 to 200 weight parts of the resultant mixture and 100 weight parts of olefin group resin (ethylene-vinyl acetate copolymer, styrene-butadiene random copolymer) are evenly mixed within an agitator such as a tumbler to form a sheet having the desired thickness and dimension. The thus obtained shielding material generates no capture gamma radiation. (Kamimura, M.)

  20. 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...

  1. 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...

  2. 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)

  3. Heavy metal oxide glasses as gamma rays shielding material

    International Nuclear Information System (INIS)

    Kaur, Preet; Singh, Devinder; Singh, Tejbir

    2016-01-01

    The gamma rays shielding parameters for heavy metal oxide glasses and concrete samples are comparable. However, the transparent nature of glasses provides additional feature to visualize inside the shielding material. Hence, different researchers had contributed in computing/measuring different shielding parameters for different configurations of heavy metal oxide glass systems. In the present work, a detailed study on different heavy metal (_5_6Ba, _6_4Gd, _8_2Pb, _8_3Bi) oxide glasses has been presented on the basis of different gamma rays shielding parameters as reported by different researchers in the recent years. It has been observed that among the selected heavy metal oxide glass systems, Bismuth based glasses provide better gamma rays shielding. Hence, Bismuth based glasses can be better substitute to concrete walls at nuclear reactor sites and nuclear labs.

  4. Heavy metal oxide glasses as gamma rays shielding material

    Energy Technology Data Exchange (ETDEWEB)

    Kaur, Preet; Singh, Devinder; Singh, Tejbir, E-mail: dr.tejbir@gmail.com

    2016-10-15

    The gamma rays shielding parameters for heavy metal oxide glasses and concrete samples are comparable. However, the transparent nature of glasses provides additional feature to visualize inside the shielding material. Hence, different researchers had contributed in computing/measuring different shielding parameters for different configurations of heavy metal oxide glass systems. In the present work, a detailed study on different heavy metal ({sub 56}Ba, {sub 64}Gd, {sub 82}Pb, {sub 83}Bi) oxide glasses has been presented on the basis of different gamma rays shielding parameters as reported by different researchers in the recent years. It has been observed that among the selected heavy metal oxide glass systems, Bismuth based glasses provide better gamma rays shielding. Hence, Bismuth based glasses can be better substitute to concrete walls at nuclear reactor sites and nuclear labs.

  5. 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)

  6. Comparison of MCNP4C and experimental results on neutron and gamma ray shielding effects for materials

    Energy Technology Data Exchange (ETDEWEB)

    Cha, Kyoon Ho; Lee, Eun Ki [KEPRI, Taejon (Korea, Republic of)

    2004-07-01

    MCNP code is a general-purpose Monte Carlo radiation transport code that can numerically simulate neutron, photon, and electron transport. Increasing the speed of computing machine is making numerical transport simulation more attractive and has led to the widespread use of such code. This code can be used for general radiation shielding and criticality accident alarm system related dose calculations, so that the version 4C2 of this code was used to evaluate the shielding effect against neutron and gamma ray experiments. The Ueki experiments were used for neutron shielding effects for materials, and the Kansas State University (KSU) photon skyshine experiments of 1977 were tested for gamma ray shielding effects.

  7. Preliminary radiation shielding design for BOOMERANG

    International Nuclear Information System (INIS)

    Donahue, Richard J.

    2002-01-01

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

  8. 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.)

  9. Nuclear data for radiation shielding

    International Nuclear Information System (INIS)

    Miyasaka, Shunichi; Takahashi, Hiroshi.

    1976-01-01

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

  10. Measurement of radiation shielding properties of polymer composites by using HPGe detector

    International Nuclear Information System (INIS)

    Gupta, Anil; Pillay, H.C.M.; Kale, P.K.; Datta, D.; Suman, S.K.; Gover, V.

    2014-01-01

    Lead is the most common radiation shield and its composite with polymers can be used as flexible radiation shields for different applications. However, lead is very hazardous and has been found to be associated with neurological disorders, kidney failure and hematotoxicity. Lead free radiation shield material has been developed by synthesizing radiation cross linked PDMS/Bi 2 O 3 polymer composites. In order to have a lead free radiation shield the relevant shielding properties such as linear attenuation, half value thickness (HVT) and tenth value thickness (TVT) have been measured by using HPGe detector. The present study describes the methodology of measurement of the shielding properties of the lead free shield material. In the measurement gamma energies such as 59.537 keV ( 241 Am), 122.061 keV and 136.474 keV ( 57 Co) are taken into consideration

  11. Multifunctional BHL Radiation Shield, Phase I

    Data.gov (United States)

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

  12. Radiation shielding 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

  13. Radiation shielding for TFTR DT diagnostics

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  14. 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

  15. A conceptual gamma shield design using the DRP model computation

    Energy Technology Data Exchange (ETDEWEB)

    Ahmad, E E [Reactor Department, Nuclear Research Center, Atomic Energy Authority, Cairo (Egypt); Rahman, F A [National Center of Nuclear Safety and Radiation Control, Atomic Energy Authority, Cairo (Egypt)

    1997-12-31

    The purpose of this investigation is to assess basic areas of concern in the development of reactor shielding conceptual design calculations. A spherical shield model composed of low carbon steel and lead have been constructed to surround a Co-60 gamma point source. two alternative configurations have been considered in the model computation. The numerical calculations have been performed using both the ANISN code and DRP model computation together with the DLC 75-Bugle 80 data library. A resume of results for deep penetration in different shield materials with different packing densities is presented and analysed. The results showed that the gamma fluxes attenuation is increased with increasing distribution the packing density of the shield material which reflects its importance of considering it as a safety parameter in shielding design. 3 figs.

  16. Polyvinyl Alcohol-Lead Nitrate Paint for Gamma Radiation Installations

    International Nuclear Information System (INIS)

    EI-Ahdal, M.A.

    2007-01-01

    Dealing with gamma ray installations represents an important problem for radiation protection workers. Radiation shielding is used to avoid the risk resulting from these gamma sources. This study suggested the use of polyvinyl alcohol (PVA) solution that contains lend nitrate (with lead metal/PVA= 1.72) to lower the gamma radiation intensity and reduce its risk to workers. This can be achieved by painting the radiation shielding with this solution Temperature relief of the irradiated solution shows the degradation of the polymer content up to 50 degree C, which starts to crosslink increasing the protection capability of this solution

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

    International Nuclear Information System (INIS)

    Satty, H. E. M.

    2014-10-01

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

  19. 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)

  20. Structural Shielding Design and Evaluation for Megavoltage X- and Gamma-Ray Radiotherapy Facilities.

    Science.gov (United States)

    Glasgow, Glenn P

    2006-09-01

    Structural Shielding Design and Evaluation for Megavoltage X- and Gamma-Ray Radiotherapy Facilities NCRP Report No. 151, 2005, 246 pp. (Hardcover $100). National Council on Radiation Protection and Measurements, 7910 Woodmont Avenue, Suite 400, Bethesda, MD 20814-3095. ISBN-10 0-0929600-87-8; http://www.NCRPonline.org. © 2006 American Association of Physicists in Medicine.

  1. Structure shielding from cloud and fallout gamma ray sources for assessing the consequences of reactor accidents

    International Nuclear Information System (INIS)

    Burson, Z.G.; Profio, A.E.

    1975-12-01

    Radiation shielding provided by transportation vehicles and structures typical of where people live and work were estimated for cloud and fallout gamma-ray sources resulting from a hypothetical reactor accident. Dose reduction factors are recommended for a variety of situations for realistically assessing the consequences of reactor accidents

  2. Practical radiation shielding for biomedical research

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  3. 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. Discussion on the standardization of concrete composition for radiation shielding design 2. Evaluation of the effect of the composition variance on the shielding property

    International Nuclear Information System (INIS)

    Ogata, Tomohiro; Kimura, Ken-ichi; Nakata, Mikihiro; Okuno, Koichi; Ishikawa, Tomoyuki

    2017-01-01

    Radiation Shielding Material Standardization Working Group of AESJ has been organized to establish Japanese standard concrete composition for radiation shielding design. We have collected concrete composition data to organize a representative concrete composition data. Neutron and Gamma dose rates penetrated through several concrete compositions are calculated by one dimensional discrete ordinate code ANISN. Effects of the variation of concrete composition on the neutron and gamma dose are evaluated. In this paper, recent standardization activity is summarized. (author)

  5. Radiation Resistance Test of Wireless Sensor Node and the Radiation Shielding Calculation

    Energy Technology Data Exchange (ETDEWEB)

    Li, Liqan; Sur, Bhaskar [Atomic Energy of Canada Limited, Ontario (Canada); Wang, Quan [University of Western Ontario, Ontario (Canada); Deng, Changjian [The University of Electronic Science and Technology, Chengdu (China); Chen, Dongyi; Jiang, Jin [Applied Physics Branch, Ontario (Korea, Republic of)

    2014-08-15

    A wireless sensor network (WSN) is being developed for nuclear power plants. Amongst others, ionizing radiation resistance is one essential requirement for WSN to be successful. This paper documents the work done in Chalk River Laboratories of Atomic Energy of Canada Limited (AECL) to test the resistance to neutron and gamma radiation of some WSN nodes. The recorded dose limit that the nodes can withstand before being damaged by the radiation is compared with the radiation environment inside a typical CANDU (CANada Deuterium Uranium) power plant reactor building. Shielding effects of polyethylene, cadmium and lead to neutron and gamma radiations are also analyzed using MCNP simulation. The shielding calculation can be a reference for the node case design when high dose rate or accidental condition (like Fukushima) is to be considered.

  6. Shielding walls against ionizing radiation. Lead bricks

    International Nuclear Information System (INIS)

    1993-04-01

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

  7. gamma. radiation of ionium

    Energy Technology Data Exchange (ETDEWEB)

    Curie, I

    1948-12-08

    Following the work of Ward (Proc Cambridge Phil Soc 35 322(1939)), the ..gamma..-radiation of ionium (from an IoTh preparation) was studied with the aid of Ta and W screens, and an aluminum counter. The screen measurements confirmed Ward's findings of two radiations, of 68 keV and of about 200 keV. The number of quanta per second of each radiation was determined with the counter, which has been calibrated on certain L lines of radium. The global quanta number of L lines of ionium was also determined. The results were as follows: 0.7 quanta ..gamma.. of 68 keV for 100 ..cap alpha..-particles; 0.2 quanta ..gamma.. of 200 keV for 100 ..cap alpha..-particles; 10 quanta L for 100 ..cap alpha..-particles. These data, which show an important internal conversion, agree with the findings of Teillac (Compt Rend 227 1227 (1948)), who investigated the ..beta..-radiation of ionium. It is the radiation 68 keV which is highly converted. On the other hand, these results do no agree with the data on the fine structure of ionium found by Rosenblum, Valadares, and Vial (Compt Rend 227 1088(1948)).

  8. Radiation shielding fiber and its manufacturing method

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, Koji; Ono, Hiroshi.

    1988-08-17

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

  9. The angular gamma flux in an iron slab shield

    International Nuclear Information System (INIS)

    Penkuhn, H.

    1975-08-01

    The angular distribution of the photon energy and dose rate flux in a plane iron shield is investigated assuming an isotropic volume source. Near the shield axis (cos phi approximately 1, with phi=angle between shield axis and gamma direction) the angular spectrum is strongly space-dependent. For large phi, space-independent fits are given. Source energies from 0.662 to 6 MeV and penetrations from 6 to 60 cm are treated and the results are compared with a similar investigation on normal concrete. The differences iron-concrete are appreciable only for the lowest source energy

  10. Methods for calculating radiation attenuation in shields

    Energy Technology Data Exchange (ETDEWEB)

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

    1964-10-01

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

  11. 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.)

  12. 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

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

    International Nuclear Information System (INIS)

    1981-01-01

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

  14. 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.

  15. Verification of radiation exposure using lead shields

    International Nuclear Information System (INIS)

    Hayashida, Keiichi; Yamamoto, Kenyu; Azuma, Masami

    2016-01-01

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

  16. INTOR radiation shielding for personnel access

    International Nuclear Information System (INIS)

    Gohar, Y.; Abdou, M.

    1981-01-01

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

  17. 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

  18. Gamma radiation in apartments

    International Nuclear Information System (INIS)

    Grindborg, J.-E.

    1983-05-01

    This investigation forms the basis for the description of methods for the detection of gamma radiation. The aim is to control that the dose limit will not exceed 50 μR/h in a room where people reside. The distribution of dose rates in different rooms has been calculated and the results have been compared with experimental data. Various instruments have been calibrated and their specifications are discussed. (G.B.)

  19. Analysis of portable gamma flaw detectors concerning radiation hygiene

    International Nuclear Information System (INIS)

    Makarova, T.V.

    1982-01-01

    Design and shields of gamma flaw detectors as one of the main factors responsible for personnel dose were studied. The analysis was conducted using the results of radiation hygienic surveys of gamma flaw detection laboratories functioning constantly in Estonia. It is shown that recently the replacement of GUP apparatuses by flaw detectors of RID and ''Gamma-RID'' (types which have design and shielding advantages is observed. However personnel doses have not reduced considerably for the last 10 years. This fact is attributed to design disadvantages of the RID and ''Gamma-RID'' apparatuses the removing of which will give the decreasing of annual personnel dose by 80 %

  20. 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.

  1. 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)

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-12-15

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

  4. Radiation shielding design for DECY-13 cyclotron using Monte Carlo method

    International Nuclear Information System (INIS)

    Rasito T; Bunawas; Taufik; Sunardi; Hari Suryanto

    2016-01-01

    DECY-13 is a 13 MeV proton cyclotron with target H_2"1"8O. The bombarding of 13 MeV protons on target H_2"1"8O produce large amounts of neutrons and gamma radiation. It needs the efficient radiation shielding to reduce the level of neutrons and gamma rays to ensure safety for workers and public. Modeling and calculations have been carried out using Monte Carlo method with MCNPX code to optimize the thickness for the radiation shielding. The calculations were done for radiation shielding of rectangular space room type with the size of 5.5 m x 5 m x 3 m and thickness of 170 cm made from lightweight concrete types of portland. It was shown that with this shielding the dose rate outside the wall was reduced to 1 μSv/h. (author)

  5. Gamma ray shielding: a web based interactive program

    International Nuclear Information System (INIS)

    Subbaiah, K.V.; Senthi Kumar, C.; Sarangapani, R.

    2005-01-01

    A web based interactive computing program is developed using java for quick assessment of Gamma Ray shielding problems. The program addresses usually encountered source geometries like POINT, LINE, CYLINDRICAL, ANNULAR, SPHERICAL, BOX, followed by 'SLAB' shield configurations. The calculation is based on point kernel technique. The source points are randomly sampled within the source volume. From each source point, optical path traversed in the source and shield media up to the detector location is estimated to calculate geometrical and material attenuations, and then corresponding buildup factor is obtained, which accounts for scattered contribution. Finally, the dose rate for entire source is obtained by summing over all sampled points. The application allows the user to select one of the seven regular geometrical bodies and provision exist to give source details such as emission energies, intensities, physical dimensions and material composition. Similar provision is provided to specify shield slab details. To aid the user, atomic numbers, densities, standard build factor materials and isotope list with respective emission energies and intensity for ready reference are given in dropdown combo boxes. Typical results obtained from this program are validated against existing point kernel gamma ray shielding codes. Additional facility is provided to compute fission product gamma ray source strengths based on the fuel type, burn up and cooling time. Plots of Fission product gamma ray source strengths, Gamma ray cross-sections and buildup factors can be optionally obtained, which enable the user to draw inference on the computed results. It is expected that this tool will be handy to all health physicists and radiological safety officers as it will be available on the internet. (author)

  6. 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.)

  7. 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

  8. Development of EASYQAD version β. A visualization code system for gamma and neutron shielding calculations

    International Nuclear Information System (INIS)

    Kim, Jae Cheon; Kim, Soon Young; Lee, Hwan Soo; Ha, Pham Nhu Viet; Kim, Jong Kyung

    2008-01-01

    EASYQAD version β was developed by MATLAB GUI (Graphical User Interface) as a visualization code system based on QAD-CGGP-A point-kernel code for convenient shielding calculations of gammas and neutrons. It consists of four graphic interface modules including GEOMETRY, INPUT, OUTPUT, and SHIELD. These modules were compiled in C++ programming language by using the MATLAB Compiler Toolbox to form a stand-along code system that can be run on the Windows XP operating system without MATLAB installation. In addition, EASYQAD version β has user-friendly graphical interfaces and, additionally, many useful functions in comparison with QAD- CGGP-A such as common material library, line and grid detectors, and multi-group energy calculations so as to increase its applicability in the field of radiation shielding analysis. It is a powerful tool for non-experts to analyze easily the shielding problems without special training. Therefore, EASYOAD version β is expected to contribute effectively to the development of radiation shielding analysis by providing users in medical and industrial fields with an efficient radiation shielding code. (author)

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

    International Nuclear Information System (INIS)

    Sekiguchi, Noboru; Naito, Yoshitaka

    1986-04-01

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

  10. Concrete for γ radiation shielding

    International Nuclear Information System (INIS)

    Azevedo e Souza, A.C. de; Rogers, John Douglas

    1980-01-01

    The attenuation characteristics of γ radiation in concrete slabs, considering their mechanical resistence and densities were determined. One heavy concrete which was used, was prepared using as additives iron ore and Fe 2 O 3 pellets in various grain sizes. Fortran programs were used for analysing data and determining the absorption coefficients and attenuation factors. (Author) [pt

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  12. 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)

  13. Thermal Degradation of Lead Monoxide Filled Polymer Composite Radiation Shields

    International Nuclear Information System (INIS)

    Harish, V.; Nagaiah, N.

    2011-01-01

    Lead monoxide filled Isophthalate resin particulate polymer composites were prepared with different filler concentrations and investigated for physical, thermal, mechanical and gamma radiation shielding characteristics. This paper discusses about the thermo gravimetric analysis of the composites done to understand their thermal properties especially the effect of filler concentration on the thermal stability and degradation rate of composites. Pristine polymer exhibits single stage degradation whereas filled composites exhibit two stage degradation processes. Further, the IDT values as well as degradation rates decrease with the increased filler content in the composite.

  14. 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)

  15. Radiation and shielding around beam absorbers

    International Nuclear Information System (INIS)

    Hurkmans, A.; Maas, R.

    1978-12-01

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

  16. 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

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  18. 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)

  19. Radiation shielding properties of a novel cement–basalt mixture for nuclear energy applications

    Energy Technology Data Exchange (ETDEWEB)

    Ipbüker, Cagatay; Nulk, Helena; Gulik, Volodymyr [University of Tartu, Institute of Physics (Estonia); Biland, Alex [HHK Technologies, Houston (United States); Tkaczyk, Alan Henry, E-mail: alan@ut.ee [University of Tartu, Institute of Physics (Estonia)

    2015-04-01

    Highlights: • Basalt fiber is a relatively cheap material that can be used as reinforcement. • Gamma-ray attenuation remains relatively stable with addition of basalt fiber. • Neutron attenuation remains relatively stable with addition of basalt fiber. • Cement–basalt mixture has a good potential for use in nuclear energy applications. - Abstract: The radiation shielding properties of a new proposed building material, a novel cement–basalt fiber mixture (CBM), are investigated. The authors analyze the possibility of this material to be a viable substitute to outgoing materials in nuclear energy applications, which will lead to a further sustained development of nuclear energy in the future. This computational study involves four types of concrete with various amounts of basalt fiber in them. The gamma-ray shielding characteristics of proposed CBM material are investigated with the help of WinXCom program, whereas the neutron shielding characteristics are computed by the Serpent code. For gamma-ray shielding, we find that the attenuation coefficients of concretes with basalt fibers are not notably influenced by the addition of fibers. For neutron shielding, additional basalt fiber in mixture presents negligible effect on neutron radiation shielding. With respect to radiation shielding, it can be concluded that basalt fibers have good potential as an addition to heavyweight concrete for nuclear energy applications.

  20. Radiation shielding properties of a novel cement–basalt mixture for nuclear energy applications

    International Nuclear Information System (INIS)

    Ipbüker, Cagatay; Nulk, Helena; Gulik, Volodymyr; Biland, Alex; Tkaczyk, Alan Henry

    2015-01-01

    Highlights: • Basalt fiber is a relatively cheap material that can be used as reinforcement. • Gamma-ray attenuation remains relatively stable with addition of basalt fiber. • Neutron attenuation remains relatively stable with addition of basalt fiber. • Cement–basalt mixture has a good potential for use in nuclear energy applications. - Abstract: The radiation shielding properties of a new proposed building material, a novel cement–basalt fiber mixture (CBM), are investigated. The authors analyze the possibility of this material to be a viable substitute to outgoing materials in nuclear energy applications, which will lead to a further sustained development of nuclear energy in the future. This computational study involves four types of concrete with various amounts of basalt fiber in them. The gamma-ray shielding characteristics of proposed CBM material are investigated with the help of WinXCom program, whereas the neutron shielding characteristics are computed by the Serpent code. For gamma-ray shielding, we find that the attenuation coefficients of concretes with basalt fibers are not notably influenced by the addition of fibers. For neutron shielding, additional basalt fiber in mixture presents negligible effect on neutron radiation shielding. With respect to radiation shielding, it can be concluded that basalt fibers have good potential as an addition to heavyweight concrete for nuclear energy applications

  1. Gamma radiation in dwellings

    International Nuclear Information System (INIS)

    Mjoenes, L.

    1981-08-01

    A nationwide investigation has been made into the gamma radiation in Swedish dwellings. The measurements were made with small detectors containing thermoluminescent dosimeters. The detectors were sent to the selected participants by mail. 1300 dwellings were included in the investigation. In each dwelling three measurements were made: one detector was placed in the kitchen, one in the living-room and one in the bedroom. The mean annual absorbed tissue dose in dwellings in Sweden was found to be 0.65 mGy (corresponding to an exposition rate of 12 μR/h) when the contribution from cosmic radiation had been subtracted. That represents an annual collective dose of about 4000 mansieverts to the population of Sweden. From a previous investigation we have calculate the mean value for the gamma radiation in Swedish dwellings for 1950 to be 0.4 mGy/a (8μR/h). The reason for the relatively large increase in the mean value is an increased use of building materials on stone, particularly of lightweight concrete based on alum shale, from 1940 to middle 1960s. The production of this type of lightweight concrete was discontinued in 1975 and the use of other stone-based building materials has decreased. The mean value of gamma radiation in Swedish dwellings is therefore expected to decrease slowly in the future if this tendency holds. Sweden has some 3.5 million dwellings. About 10 % of them have mean values of 1 mGy/a (19 μR/h) or more, 0.2 % have 3 mGy/a (57 μR/h) or more and a couple of hundred 5 mGy/a (95 μR/h or more. The mean value for detached houses was found to be 0.43 mGy/a (8 μR/h) and for dwellings in multi-family houses 0.80 mGy/a (15 μR/h). The investigation dwellings have also been classified according to the building materials, the year of construction and the degree of urbanization of the area. (author)

  2. 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.

  3. 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.

  4. ICRS1, Proceedings of the First Radiation Shielding Symposium, Cambridge, UK 1958

    International Nuclear Information System (INIS)

    Goebelbecker, Hans-Juergen

    2008-01-01

    Description: The papers of the European Atomic Energy Society Symposium VI-58 on radiation shielding (ICRS1) held at Caius College, Cambridge England from 26 to 29 August 1958 are collected here for the first time in electronic form. This symposium was organised in connection with the Second Atoms for Peace Conference held in Geneva Held in Geneva from 1 to 13 September 1958. The Topics discussed covered gamma rays and neutron radiation; the Methods discussed were analytical approaches, semi-empirical Methods, simple computer codes, Monte Carlo method. Little quality nuclear data for shielding calculations was available and the presentations would concentrate on removal cross-sections and build-up factors. Experimental techniques in support to estimate the effective shielding properties of materials were discussed such as general experimental shielding techniques and experiments on neutron attenuation in different materials and on concrete as shield. Foil detectors for spectra measurements and determination of dose rates were mainly used. The typical issues addressed were gamma-heating, gamma spectra, neutron induced gammas, fission products gamma spectra, skyshine radiation and neutron ducts - streaming. Most participants were researchers from the naval and aeronautics sector

  5. Radiation breeding researches in gamma field. Results of researches

    International Nuclear Information System (INIS)

    Morishita, Toshikazu

    2006-01-01

    Abstract of radiation breeding researches and outline of gamma field in IRB (Institute of Radiation Breeding) are described. The gamma field is a circular field of 100 m radius with 88.8TBqCo-60 source at the center. The field is surrounded by a shielding dike of 8 m in height. The effects of gamma irradiation on the growing plants, mutant by gamma radiation and plant molecular biological researches using mutant varieties obtained by the gamma field are explained. For examples, Japanese pear, chrysanthemum, Cytisus, Eustoma grandiflorum, Manila grass, tea and rose are reported. The mutant varieties in the gamma field, nine mutant varieties of flower colors in chrysanthemum, evergreen mutant lines in Manila grass, selection of self-compatible mutants in tea plant, and the plants of the gamma field recently are shown. (S.Y.)

  6. 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)

  7. 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

  8. 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.

  9. 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

  10. 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.

  11. Gamma radiation and chickens

    International Nuclear Information System (INIS)

    Toropilova, D.; Takac, L.; Toropila, M.; Tomko, M. M.

    2014-01-01

    In our work, we focused the effect of low doses of gamma radiation on metabolic parameters in chickens. In the first group of chickens we monitor changes of the concentration in glucose and cholesterol after whole body irradiation dose of chicken (3 Gy). In the second group of chickens we studied the combined effect of radiation and intraperitoneal application solution of zinc chloride to changes of the concentration in glucose and total cholesterol. In the tissues of organisms are found only in a very small amount of microelements however are of particular importance in a number of enzymatic catalytic and regulatory processes. Zinc is found in all cells of the body. However, it is the highest percentage of zinc contained in muscle and bone cells. Resorption takes place in the small intestine, especially in the duodenum. For both groups of chickens, we performed analyzes on the 3 rd , 7 th , 14 th , 21 st and 30 day. Results and an overview of the work can be helpful in the peaceful uses of nuclear energy and in preventing diseases from exposure to radiation, but also in the case of the consequences after nuclear accidents. (authors)

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

    International Nuclear Information System (INIS)

    Almeida Junior, Airton T.; Araujo, F.G.S.; Nogueira, M.S.; Santos, M.A.P.

    2013-01-01

    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 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 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)

  13. Evaluation of gamma ray durability and its application of shielded RF tags

    International Nuclear Information System (INIS)

    Teraura, Nobuyuki; Ito, Kunio; Kobayashi, Daisuke; Sakurai, Kouichi

    2015-01-01

    In this study, the RF (Radio Frequency) tag with radiation shield is developed and its gamma ray durability is evaluated. RFID (RF Identification) is a radio-wave-based identification technology that can be used for various items. RF tags find use in many applications, including item tracing, access control, etc. RF tags can be classified as active RF tags, which have inbuilt voltaic cells, and passive RF tags without these cells. Passive RF tags, known for their low price and durability, are used in various fields. For instance, they are used for equipment maintenance in factories and thermal power plants. Several frequencies are used for RF tags. Further, RF tagging on the UHF (Ultra High Frequency) frequencies allows a communication range of approximately 10 m, and thus, remote reading is possible. When used in radiation environments such as in nuclear power plants, remote reading can contribute to the reduction of radiation exposure. However, because semiconductors are the primary elements used in the manufacture of RF tags, they can be damaged by radiation, and operational errors can occur. Therefore, this technology has not been used in environments affected by relatively high radiation levels. Therefore, in nuclear power plants, the use of RF tags is limited in areas of low radiation levels. In our study, we develop and manufacture a new RF tag with a radiation shield cover that provides error correction functionality. It is expected that radiation shielded RF tags will improve the radiation-proof feature, and its application range will be expanded. Using the radiation-proof RF tag, we have conducted radiation durability tests. These tests are of two types: one using low energy gamma ray, and the other using high-energy gamma ray. Experimental results are then analyzed. The number of applications for radiation shielded RF tags is considerably increasing, because it can be used in various radiation environments other than nuclear power plants as well, such as

  14. 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.

  15. 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.)

  16. Development of EASYQAD version β: A Visualization Code System for QAD-CGGP-A Gamma and Neutron Shielding Calculation Code

    International Nuclear Information System (INIS)

    Kim, Jae Cheon; Lee, Hwan Soo; Ha, Pham Nhu Viet; Kim, Soon Young; Shin, Chang Ho; Kim, Jong Kyung

    2007-01-01

    EASYQAD had been previously developed by using MATLAB GUI (Graphical User Interface) in order to perform conveniently gamma and neutron shielding calculations at Hanyang University. It had been completed as version α of radiation shielding analysis code. In this study, EASYQAD was upgraded to version β with many additional functions and more user-friendly graphical interfaces. For general users to run it on Windows XP environment without any MATLAB installation, this version was developed into a standalone code system

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

    CERN Multimedia

    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...

  18. 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

  19. Gamma rays shielding parameters for white metal alloys

    Science.gov (United States)

    Kaur, Taranjot; Sharma, Jeewan; Singh, Tejbir

    2018-05-01

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

  20. Design, fabrication, and properties of a continuous carbon-fiber reinforced Sm_2O_3/polyimide gamma ray/neutron shielding material

    International Nuclear Information System (INIS)

    Wang, Peng; Tang, Xiaobin; Chai, Hao; Chen, Da; Qiu, Yunlong

    2015-01-01

    Highlights: • Sm_2O_3 is used for neutron absorber instead of B_4C, and Sm_2O_3 has a good photon-shielding effect. • Carbon-fiber cloth and polyimide were used to enhance shielding materials’ mechanical behavior and thermal behavior. • Both Monte Carlo method and shielding test were used to evaluate shielding performance of the novel shielding material. - Abstract: The design and fabrication of shielding materials with good heat-resistance and mechanical properties is a major problem in the radiation shielding field. In this paper, based on gamma ray and neutron shielding theory, a continuous carbon-fiber reinforced Sm_2O_3/polyimide gamma ray/neutron shielding material was fabricated by hot-pressing method. The material's application behavior was subsequently evaluated using neutron shielding, photon shielding, mechanical tensile, and thermogravimetric analysis–differential scanning calorimetry tests. The results show that the tensile strength of the novel shielding material exceeds 200 MPa, which makes it of similar strength to aluminum alloy. The material does not undergo crosslinking and decomposition reactions at 300 °C and it can be used in such environments for long periods of time. The continuous carbon-fiber reinforced Sm_2O_3/polyimide material has a good shielding performance with respect to gamma rays and neutrons. The material thus has good prospects for use in fusion reactor system and nuclear waste disposal applications.

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

    International Nuclear Information System (INIS)

    Sartori, Enrico; Vaz, Pedro

    2000-01-01

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

  2. 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.

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

    International Nuclear Information System (INIS)

    Zorla, Eyüp; Ipbüker, Cagatay; Biland, Alex; Kiisk, Madis; Kovaljov, Sergei; Tkaczyk, Alan H.; Gulik, Volodymyr

    2017-01-01

    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. 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)

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

    International Nuclear Information System (INIS)

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

    2017-01-01

    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. - Highlights: • A new method of recycling glass of Cathode Ray Tubes (CRTs) has been presented. • The glass from CRTs used as raw materials for radiation shielding glass. • The resulted glass have good optical properties and stability against radiations.

  6. 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...

  7. Lunar soil as shielding against space radiation

    Energy Technology Data Exchange (ETDEWEB)

    Miller, J. [Lawrence Berkeley National Laboratory, MS 83R0101, 1 Cyclotron Road, Berkeley, CA 94720 (United States)], E-mail: miller@lbl.gov; Taylor, L. [Planetary Geosciences Institute, Department of Earth and Planetary Sciences, University of Tennessee, Knoxville, TN 37996 (United States); Zeitlin, C. [Southwest Research Institute, Boulder, CO 80302 (United States); Heilbronn, L. [Department of Nuclear Engineering, University of Tennessee, Knoxville, TN 37996 (United States); Guetersloh, S. [Department of Nuclear Engineering, Texas A and M University, College Station, TX 77843 (United States); DiGiuseppe, M. [Northrop Grumman Corporation, Bethpage, NY 11714 (United States); Iwata, Y.; Murakami, T. [National Institute of Radiological Sciences, Chiba 263-8555 (Japan)

    2009-02-15

    We have measured the radiation transport and dose reduction properties of lunar soil with respect to selected heavy ion beams with charges and energies comparable to some components of the galactic cosmic radiation (GCR), using soil samples returned by the Apollo missions and several types of synthetic soil glasses and lunar soil simulants. The suitability for shielding studies of synthetic soil and soil simulants as surrogates for lunar soil was established, and the energy deposition as a function of depth for a particular heavy ion beam passing through a new type of lunar highland simulant was measured. A fragmentation and energy loss model was used to extend the results over a range of heavy ion charges and energies, including protons at solar particle event (SPE) energies. The measurements and model calculations indicate that a modest amount of lunar soil affords substantial protection against primary GCR nuclei and SPE, with only modest residual dose from surviving charged fragments of the heavy beams.

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

    International Nuclear Information System (INIS)

    Kugel, H.W.; Ascione, G.; Elwood, S.; Gilbert, J.; Hwang, D.; Lewis, M.; Levine, J.; Ku, L.P.; Rule, K.; Hajnal, F.

    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

  9. Development and production of radiation shielding window (RSW) glass: Indian scenario

    International Nuclear Information System (INIS)

    Phani, K.K.

    2006-01-01

    Nuclear energy/power and its peaceful applications play an ever increasing role in India. Irradiated nuclear fuels, irradiated structural materials from reactors, nuclear wastes and radio-isotopes emit high energy gamma radiations which are extremely health hazardous. These materials are handled remotely by manipulators inside the hot cells, which are constructed by shielding materials such as lead and concrete walls. The direct visual control of processes in the hot cells during operation demands the windows in the radiation shielding walls. These windows must provide the clear viewing but yet ensure the good protection to the working personnel from the high energy radiation

  10. Determination of material and its thickness for Cs-137 gamma source shielding

    International Nuclear Information System (INIS)

    Tukiman

    2008-01-01

    Its has been determined the shielding material and its thickness necessarily conducted due to every material will have different half-thickness characteristics, and by the selection a suitable material and its thickness will be obtained. Half-thickness of any material is the ability of the material at a certain thickness to absorb any radiation intensity so that the intensity becomes half of its source. Sample materials to be used are concrete, wood, and lead with their thickness varied. From experiment data and theoretical computation can be concluded that lead is the suitable material for shielding with the value of HVT for gamma radiation 0,732 cm. For wood and concrete will give half-thickness of 11,0 cm and 3,164 cm respectively. (author)

  11. 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

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

    International Nuclear Information System (INIS)

    Trubey, D.K.

    1975-11-01

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

  13. Important aspects of radiation shielding for fusion reactor tokamaks

    International Nuclear Information System (INIS)

    Abdou, M.A.

    1977-01-01

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

  14. 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...

  15. 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...

  16. 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.

  17. PEP radiation shielding tests in SLAC A Beam

    International Nuclear Information System (INIS)

    Ash, W.; DeStaebler, H.; Harris, J.; Jenkins, T.; Murray, J.

    1977-09-01

    Radiation shielding tests designed to simulate possible conditions in and around the PEP experimental halls were conducted. The SLAC A Beam was targeted in the block tunnel at a point about midway between End Station A and Beam Dump East. At that site it was relatively easy to rearrange the concrete block structure to simulate the various shielding configurations under consideration for PEP. Extensive surveys of neutron and ionizing radiation were made. Complete results of the shielding tests are given

  18. Experimental investigation of clay fly ash bricks for gamma-ray shielding

    Energy Technology Data Exchange (ETDEWEB)

    Mann Harjinder Singh; Mudahar, Gumel Singh [Dept. of Physics, Punjabi University, Patiala (India); Brar, Gurdarshan Singh [Dept. of Higher Education, Additional Project Director, Chandigarh (India); Mann, Kulwinder Singh [Dept. of Applied Sciences, I.K. Gujral Punjab Technical University, Jalandhar (India)

    2016-10-15

    This study aims to determine the effect of fly ash with a high replacing ratio of clay on the radiation shielding properties of bricks. Some interaction parameters (mass attenuation coefficients, half value layer, effective atomic number, effective electron density, and absorption efficiency) of clay fly ash bricks were measured with a NaI(Tl) detector at 661.6 keV, 1,173.2 keV, and 1,332.5 keV. For the investigation of their shielding behavior, fly ash bricks were molded using an admixture to clay. A narrow beam transmission geometry condition was used for the measurements. The measured values of these parameters were found in good agreement with the theoretical calculations. The elemental compositions of the clay fly ash bricks were analyzed by using an energy dispersive X-ray fluorescence spectrometer. At selected energies the values of the effective atomic numbers and effective electron densities showed a very modest variation with the composition of the fly ash. This seems to be due to the similarity of their elemental compositions. The obtained results were also compared with concrete, in order to study the effect of fly ash content on the radiation shielding properties of clay fly ash bricks. The clay fly ash bricks showed good shielding properties for moderate energy gamma rays. Therefore, these bricks are feasible and eco-friendly compared with traditional clay bricks used for construction.

  19. LSHINSE, Air Scattering Neutron and Gamma Dose rates for Complex Shielding Geometry

    International Nuclear Information System (INIS)

    Baran, A.; Gruen, M.; Leicht, R.

    1991-01-01

    1 - Description of program or function: The program LSHINSE is used to calculate the flux and the dose rate caused by gamma radiation emanating from a point source and being scattered in surrounding air. The program considers all forms of single scattering. Multiple scattering is taken into account in an approximate way by use of buildup factors. 2 - Method of solution: The program LSHINSE solves the equations for skyshine by use of Simpson integration. The integration limits are chosen such that the partial shielding is approximated by rectangular walls around the source. In addition, the attenuation of the primary radiation by a room ceiling can be calculated for several materials. By giving the height of the ceiling, the scattering in the air of the room can be calculated. By specifying energy groups the spectrum of the scattered radiation can be obtained. Valid energy range is 0.1 - 0.2 MeV, where the lower limit is due to uncertainties in the buildup factors. 3 - Restrictions on the complexity of the problem: The program is restricted to rectangular shielding problems involving gamma radiation in the range of 0.1 to 2.0 MeV

  20. 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

  1. 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

  2. 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)

  3. Perfecting of shielding calculation technique against the gamma rays arising from a Tokamak with the TFR experience. Application to the conceptual design Tokamak TORE 2 SUPRA

    International Nuclear Information System (INIS)

    Diop, Cheikh M'Backe.

    1980-09-01

    The conception of the necessary shielding around a conceptual design Tokamak requires to execute an estimated calculation of the doses due to the different radiation sources arising from the machine: the thermonuclear neutron source and the gamma ray source emitted during the interaction of the runaway electrons with the diaphragm. In this study, we propose a theorical method to calculate this gamma source. We tackle also the shielding problem of the conceptual design Tokamak: TORE 2 SUPRA [fr

  4. 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

  5. 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.)

  6. 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

  7. 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

  8. Observation of galactic gamma radiation

    International Nuclear Information System (INIS)

    Paul, J.A.

    1982-09-01

    A complete and deep survey of the galactic high-energy gamma radiation is now available, thanks to the gamma-ray telescopes on board of the SAS-2 and COS-B spacecrafts. A comparison of the COS-B gamma-ray survey with a fully sampled CO survey together with an Hsub(I) survey is used to show that a simple model, in which uniformly distributed cosmic rays interact with the interstellar gas, can account for almost all the gamma-ray emission observed in the first galactic quadrant. At medium galactic latitudes, it is shown that a relationship exists between the gamma radiation and the interstellar absorption derived from galaxy counts. Therefore gamma rays from the local galactic environment can be used as a valuable probe of the content and structure of the local interstellar medium. The large scale features of the local interstellar gas are revealed, in particular wide concentrations of nearby molecular hydrogen. On a smaller scale, the detection of numerous localized gamma-ray sources focuses the attention on some particular phases of clusters of young and massive stars where diffuse processes of gamma-ray emission may also be at work

  9. 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

  10. Influence on cell proliferation of background radiation or exposure to very low, chronic gamma radiation

    International Nuclear Information System (INIS)

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

    1987-01-01

    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

  11. 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

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

    International Nuclear Information System (INIS)

    Asano, Y.

    2006-01-01

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

  13. Multi-objective optimization design method of radiation shielding

    International Nuclear Information System (INIS)

    Yang Shouhai; Wang Weijin; Lu Daogang; Chen Yixue

    2012-01-01

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

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

    International Nuclear Information System (INIS)

    Singh, Harvinder; Singh, Kulwant; Gerward, Leif; Singh, Kanwarjit; Sahota, Hari Singh; Nathuram, Rohila

    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)B 2 O 3 (x=0.1-0.26) glasses at photon energies 511, 662, 1173 and 1332 keV and compared 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

  15. Radiation protection and shielding standards for the 1980s

    International Nuclear Information System (INIS)

    Trubey, D.K.

    1982-01-01

    The American Nuclear Society (ANS) is a standards-writing organization member of the American National Standards Institute (ANSI). The ANS Standards Committee has a subcommittee denoted ANS-6, Radiation Protection and Shielding, whose charge is to develop standards for radiation protection and shield design, to provide shielding information to other standards-writing groups, and to develop standard reference shielding data and test problems. This paper is a progress report of this subcommittee. Significant progress has been made since the last comprehensive report to the Society

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2018-02-15

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

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

    International Nuclear Information System (INIS)

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

    2018-01-01

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

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

    Science.gov (United States)

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

    2018-02-01

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

  19. Onboard radiation shielding estimates for interplanetary manned missions

    International Nuclear Information System (INIS)

    Totemeier, A.; Jevremovic, T.; Hounshel, D.

    2004-01-01

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

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  1. Gamma Radiation from Fission Fragments

    International Nuclear Information System (INIS)

    Higbie, Jack

    1969-10-01

    The gamma radiation from the fragments of the thermal neutron fission of 235 U has been investigated, and the preliminary data are presented here with suggestions for further lines of research and some possible interpretations of the data. The data have direct bearing on the fission process and the mode of fragment de-excitation. The parameters measured are the radiation decay curve for the time interval (1 - 7) x 10 -10 sec after fission, the photon yield, the total gamma ray energy yield, and the average photon energy. The last three quantities are measured as a function of the fragment mass

  2. Gamma Radiation from Fission Fragments

    Energy Technology Data Exchange (ETDEWEB)

    Higbie, Jack

    1969-10-15

    The gamma radiation from the fragments of the thermal neutron fission of {sup 235}U has been investigated, and the preliminary data are presented here with suggestions for further lines of research and some possible interpretations of the data. The data have direct bearing on the fission process and the mode of fragment de-excitation. The parameters measured are the radiation decay curve for the time interval (1 - 7) x 10{sup -10} sec after fission, the photon yield, the total gamma ray energy yield, and the average photon energy. The last three quantities are measured as a function of the fragment mass.

  3. Attenuation of 60Co gamma rays by barium acrylic resin composite shields

    International Nuclear Information System (INIS)

    Abdulla, Riaz; Fidha, Mariyam; Sripathi Rao, B.H.; Kudkuli, Jagadish; Rekha, P.D.; Sharma, S.D.

    2015-01-01

    Oral squamous cell carcinoma is the sixth most common cancer reported globally, with an annual incidence of over 300,000 cases, of which 62% arise in developing countries. Radiation therapy is a treatment modality that uses ionizing radiation as a therapeutic agent. It is widely employed in the treatment of head and neck cancer, as a primary therapy coupled with surgical procedure and chemotherapy or as a palliative treatment for advanced tumors. However, radiotherapy can cause a series of complications such as xerostomia, mucositis, osteoradionecrosis, and radiation caries. Composite circular disc containing different ratios of acrylic and barium sulfate (BaSO 4 ) were made in-house. The purpose of this study was to evaluate the percentage attenuation from these composite shields in 60 Co gamma rays. A maximum of 8% radiation attenuation was achieved using 1:4 ratio of acrylic-BaSO 4 composite shields. The study proposes BaSO 4 as one of the compounds in combination with acrylic resin or any other thermoplastic substances for making biocompatible radiation attenuating devices. (author)

  4. 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.

  5. Investigation of the use of Galena concrete in electromagnetic radiation shielding

    International Nuclear Information System (INIS)

    Egwuonwu, G. N.; Bukar, P. H.; Avaa, A.

    2011-01-01

    Galena samples, collected from Ishiagu, south-eastern Nigeria, were used to make high density concretes for experimental radiation shielding. The concretes were molded into cylindrical tablets of various densities and volumes in order to ascertain their attenuation capability to some electromagnetic radiations. Blue visible light and gamma-ray sourced from cobalt-60, were transmitted through the concretes and detected with the aid of Op-Amp and digital Geiger-Muller Counter respectively. The absorption coefficients of the samples of thicknesses in the range of 1.00 - 5.00 cm were determined. Results show that for a typical galena concrete of average density 2.33gcm -3 , the absorption coefficient is about 1.186 cm -1 for the blue light and 0.495cm -1 for gamma-ray. For this density, 4.45cm of the galena concrete reduces the gamma-ray intensity by 90% and its half value layer thickness is 1.40cm. The investigation however, suggests the shielding properties of the galena sourced from Ishiagu. A database of shielding strength for the in situ galena was established hence, can serve as suitable platform for quality and quantity control in radiation shielding technology in radiotherapy treatment rooms and nuclear reactors.

  6. Induced radioactivity in Bevatron concrete radiation shielding blocks

    International Nuclear Information System (INIS)

    Moeller, G.C.; Donahue, R.J.

    1994-07-01

    The Bevatron accelerated protons up to 6.2 GeV and heavy ions up to 2.1 GeV/amu. It operated from 1954 to 1993. Radioactivity was induced in some concrete radiation shielding blocks by prompt radiation. Prompt radiation is primarily neutrons and protons that were generated by the Bevatron's primary beam interactions with targets and other materials. The goal was to identify the gamma-ray emitting nuclides (t 1/2 > 0.5 yr) that could be present in the concrete blocks and estimate the depth at which the maximum radioactivity presently occurs. It is shown that the majority of radioactivity was produced via thermal neutron capture by trace elements present in concrete. The depth of maximum thermal neutron flux, in theory, corresponds with the depth of maximum induced activity. To estimate the depth at which maximum activity occurs in the concrete blocks, the LAHET Code System was used to calculate the depth of maximum thermal neutron flux. The primary beam interactions that generate the neutrons are also modeled by the LAHET Code System

  7. 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

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

    International Nuclear Information System (INIS)

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

    1980-08-01

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

  9. The construction of radiation shielding for baby ebm

    International Nuclear Information System (INIS)

    Mohd Rizal Md Chulan; Leo Kwee Wah; Lee Chee Huei; Muhamad Zahidee Taat; Fadzlie Nordin; Abu Bakar Mhd Ghazali; Mohd Yusof Ali; Mohd Rizal Mamat Ibrahim; Syed Nasaruddin Syed Idris; Mahmud Hamid; Mohd Khairi Mohd Said

    2005-01-01

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

  10. Experiment and analysis of CASTOR type model cask for verification of radiation shielding

    Energy Technology Data Exchange (ETDEWEB)

    Hattori, Seiichi; Ueki, Kohtaro.

    1988-08-01

    The radiation shielding system of CASTOR type cask is composed of the graphite cast iron and the polyethylene lod. The former fomes the cylndrical body of the cask to shield gamma rays and the latter is embeded in the body to shield neutrons. Characteristic of radiation shielding of CASTOR type cask is that zigzag arrangement of the polyethylene lod is adopted to unify the penetrating dose rate. It is necessary to use the three-dimensional analysis code to analyse the shielding performance of the cask with the complicated shielding system precisely. However, it takes too much time as well as too much cost. Therefore, the two-dimensional analysis is usually applied, in which the three-dimensional model is equivalently transformed into the two-dimensional calculation. The reseach study was conducted to verify the application of the two-dimensional analysis, in which the experiment and the analysis using CASTOR type model cask was perfomed. The model cask was manufactured by GNS campany in West Germany and the shielding ability test facilities in CRIEPI were used. It was judged from the study that the two-dimensional analysis is useful means for the practical use.

  11. Radiation protection and shielding design - Strengthening the link

    International Nuclear Information System (INIS)

    Hobson, J.; Cooper, A.

    2005-01-01

    The improvement in quality and flexibility of shielding methods and data has been progressive and beneficial in opening up new opportunities for optimising radiation protection in design. The paper describes how these opportunities can best be seized by taking a holistic view of radiation protection, with shielding design being an important component part. This view is best achieved by enhancing the role of 'shielding assessors' so that they truly become 'radiation protection designers'. The increase in speed and efficiency of shielding calculations has been enormous over the past decades. This has raised the issue of how the assessor's time now can be best utilised; pursuing ever greater precision and accuracy in shielding/dose assessments, or improving the contribution that shielding assessment makes to radiological protection and cost-effective design. It is argued in this paper that the latter option is of great importance and will give considerable benefits. Shielding design needs to form part of a larger radiation protection perspective based on a deep understanding/appreciation of the opportunities and constraints of operators and designers, enabling minimal design iterations, cost optimisation of alternative designs (with a 'lifetime' perspective) and improved realisation of design intent in operations. The future of shielding design development is argued to be not in improving the 'tool-kit', but in enhanced understanding of the 'product' and the 'process' for achieving it. The holistic processes being developed in BNFL to realise these benefits are described in the paper and will be illustrated by case studies. (authors)

  12. 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

  13. A study of shielding properties of x-ray and gamma in barium compounds

    Energy Technology Data Exchange (ETDEWEB)

    Seenappa, L.; Manjunatha, H. C. [Dept. of of Physics, Government College for women, Karnataka (India); Chandrika, B. M. [PC Extension, St. Annes School, Karnataka (India); Chikka, Hanumantharayappa [Vivekananda Degree College, Karnataka (India)

    2017-03-15

    Ionizing radiation is known to be harmful to human health. The shielding of ionizing radiation depends on the attenuation which can be achieved by three main rules, i.e. time, distance and absorbing material. The mass attenuation coefficient, linear attenuation coefficient, Half Value Layer (HVL) and Tenth Value Layer (TVL) of X-rays (32 keV, 74 keV) and gamma rays (662 keV) are measured in Barium compounds. The measured values agree well with the theory. The effective atomic numbers (Z{sub eff}) and electron density (Ne) of Barium compounds have been computed in the wide energy region 1 keV to 100 GeV using an accurate database of photon-interaction cross sections and the WinXCom program. The mass attenuation coefficient and linear attenuation coefficient for BaCO{sub 3} is higher than the BaCl{sub 2}, Ba(No{sub 3}){sub 2} and BaSO{sub 4}. HVL, TVL and mean free path are lower for BaCO{sub 3} than the BaCl{sub 2}, Ba(No{sub 3}){sub 2} and BaSO{sub 4}. Among the studied barium compounds, BaCO{sub 3} is best material for x-ray and gamma shielding.

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

    International Nuclear Information System (INIS)

    Ofori, K.

    2011-07-01

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

  15. 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.

  16. 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.

  17. 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.

  18. Estimation of gamma dose rate from hulls and shield design for the hull transport cask of Fuel Reprocessing Plant (FRP)

    International Nuclear Information System (INIS)

    Chandrasekaran, S.; Rajagopal, V.; Jose, M.T.; Venkatraman, B.

    2012-01-01

    In Fuel Reprocessing Plant (FRP), un-dissolved clad of fuel pins known as hulls are the major sources of high level solid waste. Safe handling, transport and disposal require the estimation of radioactivity as a consequent of gamma dose rate from hulls in fast reactor fuel reprocessing plant in comparison with thermal reactor fuel. Due to long irradiation time and low cooling of spent fuel, the evolution of activation products 51 Cr, 58 Co, 54 Mn and 59 Fe present as impurities in the fuel clad are the major sources of gamma radiation. Gamma dose rate from hull container with hulls from Fuel Sub Assembly (FSA) and Radial Sub Assembly (RSA) of Fuel Reprocessing Plant (FRP) was estimated in order to design the hull transport cask. Shielding computations were done using point kernel code, IGSHIELD. This paper describes the details of source terms, estimation of dose rate and shielding design of hull transport cask in detail. (author)

  19. Development of advanced, non-toxic, synthetic radiation shielding aggregate

    Energy Technology Data Exchange (ETDEWEB)

    Mudgal, Manish; Chouhan, Ramesh Kumar; Verma, Sarika; Amritphale, Sudhir Sitaram; Das, Satyabrata [CSIR-Advanced Materials and Processes Research Institute, Bhopal (India); Shrivastva, Arvind [Nuclear Power Corporation of India Ltd. (NPCIL), Mumbai (India)

    2018-04-01

    For the first time in the world, the capability of red mud waste has been explored for the development of advanced synthetic radiation shielding aggregate. Red mud, an aluminium industry waste consists of multi component, multi elemental characteristics. In this study, red mud from two different sources have been utilized. Chemical formulation and mineralogical designing of the red mud has been done by ceramic processing using appropriate reducing agent and additives. The chemical analysis, SEM microphotographs and XRD analysis confirms the presence of multi-component, multi shielding and multi-layered phases in both the different developed advance synthetic radiation shielding aggregate. The mechanical properties, namely aggregate impact value, aggregate crushing value and aggregate abrasion value have also been evaluated and was compared with hematite ore aggregate and found to be an excellent material useful for making advanced radiation shielding concrete for the construction of nuclear power plants and other radiation installations.

  20. 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...

  1. 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...

  2. 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...

  3. Studying the shielding properties of lead glass composites using neutrons and gamma rays

    International Nuclear Information System (INIS)

    Osman, A.M.; El-Sarraf, M.A.; Abdel-Monem, A.M.; El-Sayed Abdo, A.

    2015-01-01

    Highlights: • Samples of sodalime silica glass loaded with different ratios of PbO were prepared. • Leaded glass composites were investigated for radiation shielding. • Experimental and theoretical attenuation parameters were studied. • Experimental and theoretical (MCNP5) results were in good agreement. - Abstract: The present work deals with the shielding properties of lead glass composites to find out its integrity for practical shielding applications and radiological safety. Composites of different lead oxide ratios (x = 0, 5, 10, 15 and 25 wt.%) have been prepared by the Nasser Glass and Crystal Company (Egypt). Attenuation measurements have been carried out using a collimated emitted beam from a fission 252 Cf (100 μg) neutron source, and the neutron–gamma spectrometer with stilbene scintillator. The pulse shape discriminating (P.S.D.) technique based on the zero cross-over method was used to discriminate between neutron and gamma-ray pulses. Thermal neutron fluxes were measured using the BF3 detector and thermal neutron detection system. The attenuation relations were used to evaluate fast neutron macroscopic effective removal cross-section Σ R-Meas (cm −1 ), gamma rays total attenuation coefficient μ (cm −1 ) and thermal neutron macroscopic cross-section Σ Meas (cm −1 ). Theoretical calculations have been achieved using MCNP5 code to calculate the same two parameters. Also, MERCSF-N program was used to calculate fast neutron macroscopic removal cross-section Σ R-MER (cm −1 ). Measured and MCNP5 calculated results have been compared and were found to be in reasonable agreement

  4. 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.

  5. Transparent Metal-Salt-Filled Polymeric Radiation Shields

    Science.gov (United States)

    Edwards, David; Lennhoff, John; Harris, George

    2003-01-01

    "COR-RA" (colorless atomic oxygen resistant -- radiation shield) is the name of a transparent polymeric material filled with x-ray-absorbing salts of lead, bismuth, cesium, and thorium. COR-RA is suitable for use in shielding personnel against bremsstrahlung radiation from electron-beam welding and industrial and medical x-ray equipment. In comparison with lead-foil and leaded-glass shields that give equivalent protection against x-rays (see table), COR-RA shields are mechanically more durable. COR-RA absorbs not only x-rays but also neutrons and rays without adverse effects on optical or mechanical performance. The formulation of COR-RA with the most favorable mechanical-durability and optical properties contains 22 weight percent of bismuth to absorb x-rays, plus 45 atomic percent hydrogen for shielding against neutrons.

  6. Fabrication and replacement work of beryllium frame and gamma-ray shield

    International Nuclear Information System (INIS)

    Watahiki, Shunsuke; Hanawa, Yoshio; Asano, Norikazu; Hiyama, Kazuhisa; Ito, Sachito; Tsuboi, Kazuaki; Fukasaku, Akitomi

    2012-03-01

    This replacement work was carried out under refurbishment plan of JMTR for beryllium distortion draw to acceptable limit. And gamma-ray shield refurbishment was carried out the view point of prevention maintenance in consideration of operation plan. Fabrication of beryllium frame and gamma-ray shield was spent for two years it was finished in February, 2010. It took five months to replacement work from January 2010. In this report is presented fabrication and replacement work of beryllium frame and gamma-ray shield. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-15

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

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

    International Nuclear Information System (INIS)

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

    2017-01-01

    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. - Highlights: • Studies of decorative building materials for shielding of ionizing radiation. • High energy photon beam were used to obtain various interaction properties. • Marble stone ‘Carrara’ from Italy shows suitability to be used as shielding material.

  9. 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.)

  10. Light-refractory radiation shielding materials using diatomites and zeolites

    International Nuclear Information System (INIS)

    Murakami, Hideki

    2005-01-01

    It has been recently shown that diatomites and zeolites have some useful characteristics for radiation shielding materials. In this study, the availability of these materials for unexpected accidents in the nuclear sites is examined. The diatomites and zeolites, compared to existing shielding materials, have superior characteristics; low density and light weight, low in radiation-induced problem, high-heat resistance, remain unaltered by the addition of an acid except hydrofluoric acid, porous and large specific surface area, and also excellent water-absorbing property. These porous materials could also expand the shielding energy range applied and be used for fast- and thermal-neutrons, and γ ray. In addition, these materials are easy to store for long periods of time against emergency because of their natural rocks. From the examinations, it is cleared that diatomites and zeolites have excellent properties as radiation shielding materials for emergency use. (author)

  11. Combination thermal and radiation shield for well logging apparatus

    International Nuclear Information System (INIS)

    Wilson, B.F.

    1984-01-01

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

  12. Thick Galactic Cosmic Radiation Shielding Using Atmospheric Data

    Science.gov (United States)

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

    2013-01-01

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

  13. 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.

  14. Assessment of shielding integrity of Co-60 gamma irradiation-ray scanner at Aflao Border, Ghana

    International Nuclear Information System (INIS)

    Agbemafo, Edwin Capacity

    2016-07-01

    This study examines the current state of the shielding integrity of the 38.7 TBq Co-60 gamma ray scanner with an average energy of 1.25 MeV operated by NICK TC Scan Limited, which has been in use for destination inspection at Aflao Border of Ghana, for the past six years, (2010-2016). The facility uses a high energy ionizing radiation in its operation; therefore continuous adequacy of the installed biological shielding is critical to the protection and safety of the workers and the general public. The workload of the facility has increased since its commissioning, requiring the review of the status of the installed shielding. Theoretical calculations for dose rates and barrier thicknesses based on tenth – value layer (TVL) concept and NCRP 151, 2005 recommendations, were done around the scanning facility using the current operational data. The results were then compared with the measured dose rates and the shielding thickness constituted during the commissioning stage, and international standards. Calculated dose rate at commissioning state ranges from 0.6μSv/hr to 2.4μSv/hr with an average dose rate of 1.43μSv/hr and that of the current operational state ranges from 1.1μSv/hr to 2.6μSv/hr with an average dose rate of 1.54μSv/hr, indicating an increase of 7.9%. Even though the dose rates were all below the recommended dose limit of 20μSvh"-"1 by NCRP, there has been an increase in dose to the staff and the general public. It has been observed that, the workload has increased three-fold from the commissioning stage to current operational state over the past six years. The assessment done on the installed shielding using the current operational data indicates that the shielding is inadequate in providing protection for the general public and the workers against X-ray radiation source of energy of at least 6MeV, and therefore the facility in its current state cannot be used to house a linear accelerator of energy up to 10MeV. (au)

  15. 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

  16. 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

  17. Theoretical analysis of infrared radiation shields of spacecraft

    Science.gov (United States)

    Shealy, D. L.

    1984-01-01

    For a system of N diffuse, gray body radiation shields which view only adjacent surfaces and space, the net radiation method for enclosures has been used to formulate a system of linear, nonhomogeneous equations in terms of the temperatures to the fourth power of each surface in the coupled system of enclosures. The coefficients of the unknown temperatures in the system of equations are expressed in terms of configuration factors between adjacent surfaces and the emissivities. As an application, a system of four conical radiation shields for a spin stabilized STARPROBE spacecraft has been designed and analyzed with respect to variations of the cone half angles, the intershield spacings, and emissivities.

  18. Application of the decoupling scheme on complex neutron-gamma shielding problems

    Energy Technology Data Exchange (ETDEWEB)

    Feher, S. [Institute of Nuclear Technology, Technical University of Budapest, Budapest (Hungary); Leege, P.F.A. de; Hoogenboom, J.E.; Kloosterman, J.L. [Interfaculty Reactor Institute, Delft University of Technology, Delft (Netherlands)

    2000-03-01

    Coupled neutron-gamma shielding calculations using S{sub n} transport theory can be time consuming, especially for two- and three-dimensional geometries. In general, the CPU time of these calculations increases stronger than linear with increasing number of neutron and gamma energy groups, and depends on the order of Legendre expansion and number of S{sub n} directions used. This fact induced the idea of the decoupling method, which seems applicable to accelerate coupled neutron-gamma shielding calculations. The data included in a combined neutron-gamma library can be readily separated into a library containing neutron data only and another library containing gamma data only. Separate calculations for neutrons and gammas are performed on complex geometries using a different Legendre order expansion for neutrons and gammas. CPU savings of 60 to 85% can be achieved for the two-dimensional DORT and three-dimensional TORT calculations respectively. (author)

  19. 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.

  20. Sewage sterilization through gamma radiation

    International Nuclear Information System (INIS)

    Vieira, D.V.; Teixeira, W.M.S.; Silva, F.C.A. da

    2017-01-01

    Lack of sewage treatment and adequate sanitation conditions can contribute to the proliferation of numerous parasitic and infectious diseases in addition to water degradation. Approximately fifty types of infections can be transmitted from a sick to a healthy person through different pathways involving human excreta. Untreated sewage can contaminate water, food, hands, soil, etc. Epidemics of certain diseases such as typhoid, cholera, dysenteries, etc., and countless cases of worms are responsible for high mortality rates in third world countries. In the work the different techniques of sewage treatment by disinfestation and sterilization were analyzed, highlighting the use and the advantages of the gamma radiation as well as the aspects of the radiological protection involved. The technique of sewage sterilization using gamma radiation is a method of controlling bacteria and microorganisms. It is estimated that more than 200 large irradiators are in operation worldwide, of which 5 are in Brazil, for general sterilization use. These facilities use a large amount of radioactive material, in the order of millions of Becquerel, for sterilization with high doses of radiation, which can generate lethal doses in a few minutes. These industrial facilities use Cobalt-60, being classified by the International Atomic Energy Agency - IAEA as Category 1 of high risk, and must possess a high level of radiological protection to carry out the sterilization, standing out the defense in depth. Specific legislation on radiological protection should be drafted for safe work and avoid future radiation accidents

  1. Reconfigurable Patch Antenna Radiations Using Plasma Faraday Shield Effect

    OpenAIRE

    Barro , Oumar Alassane; Himdi , Mohamed; Lafond , Olivier

    2016-01-01

    International audience; This letter presents a new reconfigurable antenna associated with a plasma Faraday shield effect. The Faraday shield effect is realized by using a fluorescent lamp. A patch antenna operating at 2.45 GHz is placed inside the lamp. The performance of the reconfigurable system is observed in terms of S11, gain and radiation patterns by simulation and measurement. It is shown that by switching ON the fluorescent lamp, the gain of the antenna decreases and the antenna syste...

  2. 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)

  3. Effect of Gamma Ray Energies and Steel Fiber addition by Weight on some Shielding Properties of Limestone Concrete

    International Nuclear Information System (INIS)

    Abd El-Latifa, A.A.; Ikraiam, F.A.; Abd El-Latifa, A.A.; Abd Elazziz, A.; Abd Elazziz, A.

    2010-01-01

    The mass attenuation coefficient , the build up factor , the half value thickness X 1/2 , and tenth value thickness X 1/10 of fiber concrete , 0% , 1% , 2%, 3%, and 4% by weight fiber content were measured at different gamma ray energies in MeV, 0.511,1.274 from Na-22 ,1.17 ,1.33 from Co-60 and 0.662 from Cs-137 . Appreciable variations were noted in the former nuclear parameters, due to the changes in the fiber content and gamma ray energies .A comparison of shielding properties of concrete with fiber content and reference sample(concrete without fiber ) have proven that the addition of steel fibers by weight to concrete have a potential application as a radiation shielding

  4. 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.

  5. Gamma ray shielding characteristic of BiZnBo-SLS and PbZnBo-SLS glass

    Science.gov (United States)

    Syuhada Ahmad, Nor; Shahrim Mustafa, Iskandar; Mansor, Ishak; Malik, Muhammad Fadhirul Izwan bin Abdul; Ain Nabilah Razali, Nur; Nordin, Sufiniza

    2018-05-01

    The radiation shielding and optical properties of x [RmOn] (0.5‑x) [ZnO] 0.2 [B2O3] 0.3 [SLS], where RmOn are Bi2O3 and PbO with x = 0.05, 0.10, 0.20, 0.30, 0.40, and 0.45 have been prepared by using the melt-quenching method at 1200 °C and was investigated on their physical, structural and gamma ray shielding properties. Field-emission scanning electron microscope (FESEM) data revealed that the particle morphologies is aggregated and irregular in shapes and size. Energy dispersive x-ray spectroscopy (EDS) elemental mapping data confirmed that all mentioned element all present on the prepared glass. Soda Lime Silica (SLS) that is mainly composed of SiO2 has been utilized in this study as the source of SiO2 for fabrication of glass system. From the result, the density and molar volume of both glass samples increased as Bi2O3 and PbO content increased. The gamma ray shielding properties, such as linear attenuation and mass attenuation coefficient, were increased while half value layer (HVL) and mean free path (MFP) were decreased as the increased in Bi2O3 and PbO concentrations. It is recognized that the mass attenuation coefficient value of Bi2O3 and PbO glass are slightly different. From this study, it can be concluded that from the non-toxicity and shielding point of view, the bismuth glass is a good shield to gamma radiation as compared to lead glass.

  6. Galactic structure and gamma radiation

    International Nuclear Information System (INIS)

    Casse, Michel; Cesarsky, Catherine; Paul Jacques

    1977-01-01

    A model of spiral structure of the Galaxy is outlined from radiosynchrotron and gamma observations. The most interesting observations in the galactic context, obtained by the SAS II American satellite are concerned with the distribution of the γ photoemission at energies higher than 10 8 eV, along the galactic equator. The model proposed is in quantitative agreement with the present ideas on the spiral structure of the Galaxy, the galactic magnetic field, and the confinement of cosmic rays by the magnetic field and of the magnetic field by matter. Following the American era, the European COS-B satellite opens the European phase towards an identification of the discrete gamma radiation sources [fr

  7. 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

  8. The evaluation of the radiation shielding ability of lead glass

    International Nuclear Information System (INIS)

    Tsuda, Keisuke; Fukushi, Masahiro; Myojoyama, Atsushi; Kitamura, Hideaki; Nakaya, Giichiro; Hassan, Nabil; Inoue, Kazumasa; Kimura, Junichi; Sawaguchi, Masato; Kinase, Sakae; Saito, Kimiaki

    2008-01-01

    Positron emission tomography (PET) scanning with the tracer 2-[F-18] Fluoro-2deoxy-D-glucose (FDG) is widely used in the clinical PET. However, the photon energy used in the PET scans is considerably higher than that of the X-rays traditionally used in the diagnoses. The radiation protection in the PET institution, therefore, is the remaining problem. Meanwhile, lead glass has attracted considerable attention as a radiation-shielding material for the PET institution. The aim of the present study was to evaluate the radiation-shielding ability of the lead glass against the positron emitters. The shielding ability evaluations were done both in the actual experiments and in the Monte Carlo simulation. The lead glass, the object of evaluation in this study, proved to have sufficient protective effect. The development and the spread of a thinner and lighter lead glass with the same effective dose transmission factor should be expected in the near future. (author)

  9. Development of Computer Program for Analysis of Irregular Non Homogenous Radiation Shielding

    International Nuclear Information System (INIS)

    Bang Rozali; Nina Kusumah; Hendro Tjahjono; Darlis

    2003-01-01

    A computer program for radiation shielding analysis has been developed to obtain radiation attenuation calculation in non-homogenous radiation shielding and irregular geometry. By determining radiation source strength, geometrical shape of radiation source, location, dimension and geometrical shape of radiation shielding, radiation level of a point at certain position from radiation source can be calculated. By using a computer program, calculation result of radiation distribution analysis can be obtained for some analytical points simultaneously. (author)

  10. 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...

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

    Science.gov (United States)

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

    2017-10-01

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

  12. 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

  13. 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.

  14. Bismuth silicate glass containing heavy metal oxide as a promising radiation shielding material

    Science.gov (United States)

    Elalaily, Nagia A.; Abou-Hussien, Eman M.; Saad, Ebtisam A.

    2016-12-01

    Optical and FTIR spectroscopic measurements and electron paramagnetic resonance (EPR) properties have been utilized to investigate and characterize the given compositions of binary bismuth silicate glasses. In this work, it is aimed to study the possibility of using the prepared bismuth silicate glasses as a good shielding material for γ-rays in which adding bismuth oxide to silicate glasses causes distinguish increase in its density by an order of magnitude ranging from one to two more than mono divalent oxides. The good thermal stability and high density of the bismuth-based silicate glass encourage many studies to be undertaken to understand its radiation shielding efficiency. For this purpose a glass containing 20% bismuth oxide and 80% SiO2 was prepared using the melting-annealing technique. In addition the effects of adding some alkali heavy metal oxides to this glass, such as PbO, BaO or SrO, were also studied. EPR measurements show that the prepared glasses have good stability when exposed to γ-irradiation. The changes in the FTIR spectra due to the presence of metal oxides were referred to the different housing positions and physical properties of the respective divalent Sr2+, Ba2+ and Pb2+ ions. Calculations of optical band gap energies were presented for some selected glasses from the UV data to support the probability of using these glasses as a gamma radiation shielding material. The results showed stability of both optical and magnetic spectra of the studied glasses toward gamma irradiation, which validates their irradiation shielding behavior and suitability as the radiation shielding candidate materials.

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

    International Nuclear Information System (INIS)

    1988-01-01

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

  16. Evaluating shielding effectiveness for reducing space radiation cancer risks

    International Nuclear Information System (INIS)

    Cucinotta, Francis A.; Kim, Myung-Hee Y.; Ren, Lei

    2006-01-01

    We discuss calculations of probability distribution functions (PDF) representing uncertainties in projecting fatal cancer risk from galactic cosmic rays (GCR) and solar particle events (SPE). The PDFs are used in significance tests for evaluating the effectiveness of potential radiation shielding approaches. Uncertainties in risk coefficients determined from epidemiology data, dose and dose-rate reduction factors, quality factors, and physics models of radiation environments are considered in models of cancer risk PDFs. Competing mortality risks and functional correlations in radiation quality factor uncertainties are included in the calculations. We show that the cancer risk uncertainty, defined as the ratio of the upper value of 95% confidence interval (CI) to the point estimate is about 4-fold for lunar and Mars mission risk projections. For short-stay lunar missions ( 180d) or Mars missions, GCR risks may exceed radiation risk limits that are based on acceptable levels of risk. For example, the upper 95% CI exceeding 10% fatal risk for males and females on a Mars mission. For reducing GCR cancer risks, shielding materials are marginally effective because of the penetrating nature of GCR and secondary radiation produced in tissue by relativistic particles. At the present time, polyethylene or carbon composite shielding cannot be shown to significantly reduce risk compared to aluminum shielding based on a significance test that accounts for radiobiology uncertainties in GCR risk projection

  17. Determination of self shielding factors and gamma attenuation effects for tree ring samples

    International Nuclear Information System (INIS)

    Dagistan Sahin; Kenan Uenlue

    2012-01-01

    Determination of tree ring chemistry using Neutron Activation Analysis (NAA) is part of an ongoing research between Penn State University (PSU) and Cornell University, The Malcolm and Carolyn Wiener Laboratory for Aegean and Near Eastern Dendrochronology. Tree-ring chemistry yields valuable data for environmental event signatures. These signatures are a complex function of elemental concentration. To be certain about concentration of signature elements, it is necessary to perform the measurements and corrections with the lowest error and maximum accuracy possible. Accurate and precise values of energy dependent neutron flux at dry irradiation tubes and detector efficiency for tree ring sample are calculated for Penn State Breazeale Reactor (PSBR). For the calculation of energy dependent and self shielding corrected neutron flux, detailed model of the TRIGA Mark III reactor at PSU with updated fuel compositions was prepared using the MCNP utility for reactor evolution (MURE) libraries. Dry irradiation tube, sample holder and sample were also included in the model. The thermal flux self-shielding correction factors due to the sample holder and sample for were calculated and verified with previously published values. The Geant-4 model of the gamma spectroscopy system, developed at Radiation Science and Engineering Center (RSEC), was improved and absolute detector efficiency for tree-ring samples was calculated. (author)

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

    Directory of Open Access Journals (Sweden)

    Robert L. Wilby

    2013-10-01

    Full Text Available Temperature sensors are potentially susceptible to errors due to heating by solar radiation. Although this is well known for air temperature (Ta, significance to continuous water temperature (Tw monitoring is relatively untested. This paper assesses radiative errors by comparing measurements of exposed and shielded Tinytag sensors under indirect and direct solar radiation, and in laboratory experiments under controlled, artificial light. In shallow, still-water and under direct solar radiation, measurement discrepancies between exposed and shielded sensors averaged 0.4 °C but can reach 1.6 °C. Around 0.3 °C of this inconsistency is explained by variance in measurement accuracy between sensors; the remainder is attributed to solar radiation. Discrepancies were found to increase with light intensity, but to attain Tw differences in excess of 0.5 °C requires direct, bright solar radiation (>400 W m−2 in the total spectrum. Under laboratory conditions, radiative errors are an order of magnitude lower when thermistors are placed in flowing water (even at velocities as low as 0.1 m s−1. Radiative errors were also modest relative to the discrepancy between different thermistor manufacturers. Based on these controlled experiments, a set of guidelines are recommended for deploying thermistor arrays in water bodies.

  19. Neutron and gamma-ray spectra measurement on the model of the KS-150 reactor radial shielding

    International Nuclear Information System (INIS)

    Holman, M.; Hogel, J.; Marik, J.; Kovarik, K.; Franc, L.; Vespalec, R.

    1977-01-01

    A shortened model of the peripheral region of the KS-150 reactor core consisting of two rows of fuel elements and a reflector was constructed from the peripheral fuel elements of the KS-150 reactor core in an experiment on the TR-0 reactor. The mockup of the thermal shield (10 cm of steel), the pressure vessel (15 cm of steel) and the inner wall of the water biological shielding (2 cm of steel) of the KS-150 reactor were erected outside the TR-0 vessel. Fast neutron and gamma spectra were measured with a stilbene crystal scintillation spectrometer. The resonance neutron spectra were measured with 197 Au, 63 Cu and 23 Na resonance activation detectors. Fast neutron spectra inside the reactor were measured with a 10 mm diameter by 10 mm thick stilbene crystal spectrometer, outside the reactor with a 10 mm diameter by 10 mm thick and a 20 mm diameter by 20 mm thick stilbene crystal spectrometer. Neutron spectra in the energy regions of 1 eV to 3 keV and 0.6 MeV to 0.8 MeV were obtained on the core periphery, on the reflector half-thickness and in front of and behind the reactor thermal shield. Gamma spectra were obtained in front of and behind the thermal shield. It was found that the attenuation of neutron fluxes by the reflector and the thermal shield increased with increasing energy while gamma radiation attenuation decreased with increasing energy. It was not possible to obtain the neutron spectrum in the 10 to 600 keV energy range because suitable detection instrumentation was not available. (J.P.)

  20. 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...

  1. Radiation shielding design for a hot repair facility

    International Nuclear Information System (INIS)

    Courtney, J.C.; Dwight, C.C.

    1991-01-01

    A new repair and decontamination area is being built to support operations at the demonstration fuel cycle facility for the Integral Fast Reactor program at Argonne National Laboratory's site at the Idaho National Engineering Laboratory. Provisions are made for remote, glove wall, and contact maintenance on equipment removed from hot cells where spent fuel will be electrochemically processed and recycled to the Experimental Breeder Reactor-II. The source for the shielding design is contamination from a mix of fission and activation products present on items removed from the hot cells. The repair facility also serves as a transfer path for radioactive waste produced by processing operations. Radiation shields are designed to limit dose rates to no more than 5 microSv h-1 (0.5 mrem h-1) in normally occupied areas. Point kernel calculations with buildup factors have been used to design the shielding and to position radiation monitors within the area

  2. Barium-borate-flyash glasses: As radiation shielding materials

    International Nuclear Information System (INIS)

    Singh, Sukhpal; Kumar, Ashok; Singh, Devinder; Thind, Kulwant Singh; Mudahar, Gurmel S.

    2008-01-01

    The attenuation coefficients of barium-borate-flyash glasses have been measured for γ-ray photon energies of 356, 662, 1173 and 1332 keV using narrow beam transmission geometry. The photon beam was highly collimated and overall scatter acceptance angle was less than 3 o . Our results have an uncertainty of less than 3%. These coefficients were then used to obtain the values of mean free path (mfp), effective atomic number and electron density. Good agreements have been observed between experimental and theoretical values of these parameters. From the studies of the obtained results it is reported here that from the shielding point of view the barium-borate-flyash glasses are better shields to γ-radiations in comparison to the standard radiation shielding concretes and also to the ordinary barium-borate glasses

  3. 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.

  4. 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

  5. Synthesis of mullite (3Al2O32SiO2) from local kaolin for radiation shielding

    Science.gov (United States)

    Ripin, Azuhar; Mohamed, Faizal; Aman, Asyraf

    2018-04-01

    Raw kaolin from Kota Tinggi, Johor was used in this study to produce ceramic mullite (3Al2O22SiO2) for radiation shielding materials. In this work, an attempt was made to study the potential of local minerals to be used as a shielding barrier for diagnostic radiology radiation facilities in hospitals and medical centers throughout Malaysia. The conventional ceramic processing route was employed in the study using different pressing strength and sintering time. The obtained samples were characterized using X-ray diffractometer (XRD) for phase identification of each of the samples. The lead equivalent (LE) test was carried out using 15.05 mCi Cobalt-57 with gamma energy of 122 keV to compute the abilities of the mullite ceramic samples to attenuate the radiation. XRD patterns of prepared ceramics revealed the presence of orthorhombic mullite, hexagonal quartz and orthorhombic sillimanite structures. Furthermore, the radiation test displayed the ability of ceramics to shield of 70 % of gamma radiation at the distance of 60 cm from the radiation source. The highest lead equivalent thickness is 1.0 mm Pb and the lowest is about 0.06 mm Pb. From the result, it is shown that the ceramic has the potential to use as a shielding barrier in diagnostic radiology facilities due to the ability of reducing the radiation dose up to 70 % from its initial value.

  6. 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.)

  7. Method for the diffraction of terrestrial radiation, GAMMA radiation, etc

    International Nuclear Information System (INIS)

    Busenbach, P.

    1975-01-01

    The patent claim concerns shielding against so-called earth radiation. These rays originate from water veins etc., cannot be defined any closer, and are supposed to cause injury to health. The author claims that shielding is possible with the aid of welding electrodes which have cores of nickel. (ORU/LN) [de

  8. Measurement of TFTR D-T radiation shielding efficiency

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  9. Radiation shielding for the Super Collider West Utility region

    International Nuclear Information System (INIS)

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

    1994-02-01

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

  10. Non-combustible nuclear radiation shields with high hydrogen content

    International Nuclear Information System (INIS)

    Hall, W.C.; Peterson, J.M.

    1978-01-01

    The invention relates to compositions, methods of production, and uses of non-combustible nuclear radiation shields, with particular emphasis on those containing a high concentration of hydrogen atoms, especially effective for moderating neutron energy by elastic scatter, dispersed as a discontinuous phase in a continuous phase of a fire resistant matrix

  11. Radiation shield analysis for a manned Mars rover

    International Nuclear Information System (INIS)

    Morley, N.J.; ElGenk, M.S.

    1991-01-01

    Radiation shielding for unmanned space missions has been extensively studied; however, designs of man-rated shields are minimal. Engle et al.'s analysis of a man-rated, multilayered shield composed of two and three cycles (a cycle consists of a tungsten and a lithium hydride layer) is the basis for the work reported in this paper. The authors present the results of a recent study of shield designs for a manned Mars rover powered by a 500-kW(thermal) nuclear reactor. A train-type rover vehicle was developed, which consists of four cars and is powered by an SP-100-type nuclear reactor heat source. The maximum permissible dose rate (MPD) from all sources is given by the National Council on Radiation Protection and Measurements as 500 mSv/yr (50 rem/yr) A 3-yr Mars mission (2-yr round trip and 1-yr stay) will deliver a 1-Sv natural radiation dose without a solar particle event, 450 mSv/yr in flight, and an additional 100 mSv on the planet surface. An anomalously large solar particle event could increase the natural radiation dose for unshielded astronauts on the Martian surface to 200 mSv. This limits the MPD to crew members from the nuclear reactor to 300 mSv

  12. 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

  13. Laboratory tests on neutron shields for gamma-ray detectors in space

    CERN Document Server

    Hong, J; Hailey, C J

    2000-01-01

    Shields capable of suppressing neutron-induced background in new classes of gamma-ray detectors such as CdZnTe are becoming important for a variety of reasons. These include a high cross section for neutron interactions in new classes of detector materials as well as the inefficient vetoing of neutron-induced background in conventional active shields. We have previously demonstrated through Monte-Carlo simulations how our new approach, supershields, is superior to the monolithic, bi-atomic neutron shields which have been developed in the past. We report here on the first prototype models for supershields based on boron and hydrogen. We verify the performance of these supershields through laboratory experiments. These experimental results, as well as measurements of conventional monolithic neutron shields, are shown to be consistent with Monte-Carlo simulations. We discuss the implications of this experiment for designs of supershields in general and their application to future hard X-ray/gamma-ray experiments...

  14. 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

  15. 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...

  16. 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...

  17. The angular gamma flux in an iron shield due to a thin slab source

    International Nuclear Information System (INIS)

    Penkuhn, H.

    1977-04-01

    The angular spectra of the gamma energy fluxes and dose rates in iron shields due to thin and thick sources are compared. The anisotropicity increases with increasing source thickness. But the changes can be ignored near the forward direction (shield axis) and moreover for all directions at deep penetrations. At low source energies the changes are smaller than at higher ones (at equal penetrations in cm)

  18. 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)

  19. 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...

  20. Barium and calcium borate glasses as shielding materials for x rays and gamma rays

    DEFF Research Database (Denmark)

    Singh, H.; Singh, K.; Sharma, G.

    2003-01-01

    Values of the gamma-ray, mass attenuation coefficient and the effective atomic number have been determined experimentally for xBaO.(1-x) B2O3 (x=0.24, 0.30, 0.34,0.40 and 0.44) and xCaO. (I-x)B2O3 (x=0.30 and 0.40) glasses at photon energies 356, 511, 662, 1173, and 1332 keV It is pointed out tha...... that these glasses are potential radiation shielding materials. The specific volume of the glasses has been derived from density measurements and studied as a function of composition.......Values of the gamma-ray, mass attenuation coefficient and the effective atomic number have been determined experimentally for xBaO.(1-x) B2O3 (x=0.24, 0.30, 0.34,0.40 and 0.44) and xCaO. (I-x)B2O3 (x=0.30 and 0.40) glasses at photon energies 356, 511, 662, 1173, and 1332 keV It is pointed out...

  1. 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.)

  2. External dosimetry sources and shielding

    International Nuclear Information System (INIS)

    Calisto, Washington

    1994-01-01

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

  3. Comparative study of tungsten and lead as gamma ray shielding material

    International Nuclear Information System (INIS)

    Wang Jian; Zou Shuliang

    2011-01-01

    This article firstly compares the tungsten and lead's physical properties, price and environmental performance, then calculates the thickness of tungsten and lead with the gamma ray 10% transmission when the photon energy are 0.1 MeV, 0.2 MeV, 0.5, 1 MeV and 1.25 MeV, and makes a comparison chart. Finally, it establishes a commonly used shielding model, through which to validate whether the thickness of theoretical calculation can achieve an effective shielding effect by MCNP program. The results showers that tungsten as a new type of shielding material has a lot of advantages, which shielding ability is far higher than the lead. Thus it provides the reference to choose the suitable shielding materials in special occasions. (authors)

  4. 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.

  5. Shielding behavior of multi-transformation phase change materials (MTPCM) against nuclear radiations

    International Nuclear Information System (INIS)

    Kumar, Ravindra; Goplani, Deepak; Kumar, Rohitash; Das, Mrinal Kumar; Kumar, Pramod; Jodha, Ajay Singh; Misra, Manoj; Khatri, P.K.

    2008-01-01

    In nuclear hardened structures and AFV's, special shielding materials are being used to provide protection from radiations generated in nuclear blast. However, in blast an intense heat pulse is also generated along with radiation. Currently used shield does not take care of this heat pulse. Defence Laboratory, Jodhpur has developed multi transformation phase change materials (MTPCM) based cool panels for passive moderation of temperature in severe desert heat. The MTPCM contains light nuclei of hydrogen, carbon and oxygen, and thus can absorb good amount of neutrons. MTPCM can also absorb intense heat pulse along with heat generated by secondary fires during blast as its latent heat (160-170 J/g) without significant rise in temperature (melting point 36-38 deg. C). Thus MTPCM can provide protection against both radiation as well as heat pulse generated in a nuclear blast along with its designed regular function of passively moderating temperature below 40 deg C during severe desert summer. A study has been undertaken to explore multiple applications of MTPCM panel. Protection factor provided by standard MTPCM panels against neutron and gamma radiations (both initial and fall out) were measured and results compared with PF provided by special lining pad currently being used in AFV's and field structures for nuclear protection. It is observed that MTPCM provides good PF (2.17) against neutron which is better than currently used shield pads (PFP%1.8). Present paper discusses results of this study. (author)

  6. [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.

  7. Radiation shielding estimates for manned Mars space flight

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  8. Problems in radiation shielding calculations with Monte Carlo methods

    International Nuclear Information System (INIS)

    Ueki, Kohtaro

    1985-01-01

    The Monte Carlo method is a very useful tool for solving a large class of radiation transport problem. In contrast with deterministic method, geometric complexity is a much less significant problem for Monte Carlo calculations. However, the accuracy of Monte Carlo calculations is of course, limited by statistical error of the quantities to be estimated. In this report, we point out some typical problems to solve a large shielding system including radiation streaming. The Monte Carlo coupling technique was developed to settle such a shielding problem accurately. However, the variance of the Monte Carlo results using the coupling technique of which detectors were located outside the radiation streaming, was still not enough. So as to bring on more accurate results for the detectors located outside the streaming and also for a multi-legged-duct streaming problem, a practicable way of ''Prism Scattering technique'' is proposed in the study. (author)

  9. Challenges in commercial manufacture of radiation shielding glasses

    International Nuclear Information System (INIS)

    Gupta, R.K.

    2011-01-01

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

  10. 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)

  11. 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)

  12. 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.

  13. Radiation transport and shielding information, computer codes, and nuclear data for use in CTR neutronics research and development

    International Nuclear Information System (INIS)

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

    1976-01-01

    The activities of the Radiation Shielding Information Center (RSIC) of the Oak Ridge National Laboratory are being utilized in support of fusion reactor technology. The major activities of RSIC include the operation of a computer-based information storage and retrieval system, the collection, packaging, and distribution of large computer codes, and the compilation and dissemination of processed and evaluated data libraries, with particular emphasis on neutron and gamma-ray cross-section data. The Center has acquired thirteen years of experience in serving fission reactor, weapons, and accelerator shielding research communities, and the extension of its technical base to fusion reactor research represents a logical progression. RSIC is currently working with fusion reactor researchers and contractors in computer code development to provide tested radiation transport and shielding codes and data library packages. Of significant interest to the CTR community are the 100 energy group neutron and 21 energy group gamma-ray coupled cross-section data package (DLC-37) for neutronics studies, a comprehensive 171 energy group neutron and 36 energy group gamma-ray coupled cross-section data base with retrieval programs, including resonance self-shielding, that are tailored to CTR application, and a data base for the generation of energy-dependent atomic displacement and gas production cross sections and heavy-particle-recoil spectra for estimating radiation damage to CTR structural components

  14. 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.

  15. 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.

  16. Concrete Shielding For Radiation Safety And Unexpected Dangerous Inside Cobalt-60 Industrial Irradiator

    International Nuclear Information System (INIS)

    Keshk, A.B.; Aly, R.A.

    2011-01-01

    The study shows a proposed destruction inside one of three cobalt-60 industrial irradiators to determine and reduce the negative results, to improve and modify emergency plan to face terrorism works. The results show the performance of concrete shielding (walls and ceiling) contains the bad effect of dynamic pressures. The explosion forces are prevented to destructive by performance of their concrete shielding, which will contain the most components of devastated systems inside each irradiator after explosion. Shield penetration like electrical cable tunnels, pushers holes, hole with removable plug, product boxes openings, lens opening and ozone duct are affected badly by destruction. Through probability of transporting, some of devastated parts of broken radioactive cobalt- 60 pencils from inside radiation concreter room to outside (surrounded environment) are maintained and causing very danger radiation exposure by gamma rays outside irradiator. A necessity needs to modify emergency plan to prevent any explosive materials to enter inside the main building (irradiation sale) and also discovering any explosive materials which are placed inside the product boxes before passing to inside irradiator. The minimizing radiation exposure (2 mrem/h) inside underground radiation shelters are maintained by reducing radiation dose exerted from a nuclear explosion of 20 kT about 1 km away to a safe value, and calculating the protective factors of radiation main building basements are more than 40 (safety factor) as they are located under ground level, are surrounded by sandy soil and are constructed by concrete. The study shows the proposed basements of the main building maintain success to use as under ground safe radiation shelter (during emergency) with separate safe radiation trace. It begins from the main opening of irradiation sale and leads to underground proposed shelter through modified main stair

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

    International Nuclear Information System (INIS)

    Kaundal, Rajinder Singh

    2016-01-01

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

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

    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)

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

    OpenAIRE

    Kaundal, Rajinder Singh

    2016-01-01

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

  20. Evaluation of usability of the shielding effect for thyroid shield for peripheral dose during whole brain radiation therapy

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Myung Sic; Park, Ju Kyeong; Lee, Seung Hun; Kim, Yang Su; Lee, Sun Young; Cha, Seok Yong [Dept. of Radiation Oncology, Chonbuk National University Hospital, Jeonju (Korea, Republic of)

    2014-12-15

    To reduce the radiation dose to the thyroid that is affected to scattered radiation, the shield was used. And we evaluated the shielding effect for the thyroid during whole brain radiation therapy. To measure the dose of the thyroid, 300cGy were delivered to the phantom using a linear accelerator(Clinac iX VARIAN, USA.)in the way of the 6MV X-ray in bilateral. To measure the entrance surface dose of the thyroid, five glass dosimeters were placed in the 10th slice's surface of the phantom with a 1.5 cm interval. The average values were calculated by measured values in five times each, using bismuth shield, 0.5 mmPb shield, self-made 1.0 mmPb shield and unshield. In the same location, to measure the depth dose of the thyroid, five glass dosimeters were placed in the 10th slice by 2.5 cm depth of the phantom with a 1.5 cm interval. The average values were calculated by measured values in five times each, using bismuth shield, 0.5 mmPb shield, self-made 1.0 mmPb shield and unshield. Entrance surface dose of the thyroid were respectively 44.89 mGy at the unshield, 36.03 mGy at the bismuth shield, 31.03 mGy at the 0.5 mmPb shield and 23.21 mGy at a self-made 1.0 mmPb shield. In addition, the depth dose of the thyroid were respectively 36.10 mGy at the unshield, 34.52 mGy at the bismuth shield, 32.28 mGy at the 0.5 mmPb shield and 25.50 mGy at a self-made 1.0 mmPb shield. The thyroid was affected by the secondary scattering dose and leakage dose outside of the radiation field during whole brain radiation therapy. When using a shield in the thyroid, the depth dose of thyroid showed 11-30% reduction effect and the surface dose of thyroid showed 20-48% reduction effect. Therefore, by using the thyroid shield, it is considered to effectively protect the thyroid and can perform the treatment.

  1. Radiation shielding technology development for proton linear accelerator

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yong Ouk; Lee, Y. O.; Cho, Y. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Kim, M. H.; Sin, M. W.; Park, B. I. [Kyunghee Univ., Seoul (Korea, Republic of)] [and others

    2005-09-01

    This report was presented as an output of 2-year project of the first phase Proton Engineering Frontier Project(PEFP) on 'Radiation Shielding Technology Development for Proton Linear Accelerator' for 20/100 MeV accelerator beam line and facility. It describes a general design concept, provision and update of basic design data, and establishment of computer code system. It also includes results of conceptual and preliminary designs of beam line, beam dump and beam facilities as well as an analysis of air-activation inside the accelerator equipment. This report will guides the detailed shielding design and production of radiation safety analysis report scheduled in the second phase project.

  2. Validation of nuclear models used in space radiation shielding applications

    International Nuclear Information System (INIS)

    Norman, Ryan B.; Blattnig, Steve R.

    2013-01-01

    A program of verification and validation has been undertaken to assess the applicability of models to space radiation shielding applications and to track progress as these models are developed over time. In this work, simple validation metrics applicable to testing both model accuracy and consistency with experimental data are developed. The developed metrics treat experimental measurement uncertainty as an interval and are therefore applicable to cases in which epistemic uncertainty dominates the experimental data. To demonstrate the applicability of the metrics, nuclear physics models used by NASA for space radiation shielding applications are compared to an experimental database consisting of over 3600 experimental cross sections. A cumulative uncertainty metric is applied to the question of overall model accuracy, while a metric based on the median uncertainty is used to analyze the models from the perspective of model development by examining subsets of the model parameter space.

  3. 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

  4. Shielding system for the detection of radioisotopes gamma-rays emitters

    International Nuclear Information System (INIS)

    Nascimento Filho, V.F. do

    1983-01-01

    A shielding system for detection of radioisotopes gamma-rays emitters in samples of big volumes (plants, animals, soils) is presented. The detection between the beaker Marinelli and the glass tube (inside of the scintillator crystal well) is compared. The beaker Marinelli method allows a drastic reduction in the time detection of the sample. (M.A.C.) [pt

  5. Cosmical sources of gamma radiation

    Energy Technology Data Exchange (ETDEWEB)

    Kuchowicz, B [Warsaw Univ. (Poland)

    1974-01-01

    A brief historical outline of the X-ray and ..gamma..-ray astronomies is given first, then a summary of the recent status of X-ray astronomy follows. Further chapters include information on ..gamma..-ray sources in the solar system, in our Galaxy, and beyond it. In discussing linear gamma spectra attention is paid to the possibility of studying explosive nucleo-synthesis by observation of gamma lines from supernova remnants, etc. Questions of the isotropic gamma background are discussed at the end of the survey.

  6. 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

  7. A study on the apron shielding ratio according to electromagnetic radiation energy

    International Nuclear Information System (INIS)

    Jang, Dong Gun; Lee, Sang Ho; Choi, Hyung Seok; Son, Joo Chul; Yoon, Chang Yong; Ji, Yung Sik; Cho, Yong In; Lee, Hong Je; Yang, Seoung Oh

    2014-01-01

    The medical institution has been used electromagnetic radiation of various energy. But researchers are divided on whether using apron for radiation shielding will be effective or not. The purpose of present study was to analyze electromagnetic radiation shielding effect of apron by using Monte Carlo simulation. 1 MBq electromagnetic radiation was emitted from 10-500 keV at 10 keV increments in Monte Carlo simulation. Then shielded radiation dose difference was confirmed, when 0.25 mmPb shield use for shielding. As a results, shielding ratio was markedly decreased in high energy electromagnetic radiation. The radiation dose was inversely increased with 0.25 mmPb shielding

  8. A study on the apron shielding ratio according to electromagnetic radiation energy

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Dong Gun; Lee, Sang Ho; Choi, Hyung Seok; Son, Joo Chul; Yoon, Chang Yong; Ji, Yung Sik; Cho, Yong In; Lee, Hong Je; Yang, Seoung Oh [Dept. of Nuclear Medicine, Dongnam Institute of Radiological and Medical Sciences Cancer Center, Busan (Korea, Republic of)

    2014-12-15

    The medical institution has been used electromagnetic radiation of various energy. But researchers are divided on whether using apron for radiation shielding will be effective or not. The purpose of present study was to analyze electromagnetic radiation shielding effect of apron by using Monte Carlo simulation. 1 MBq electromagnetic radiation was emitted from 10-500 keV at 10 keV increments in Monte Carlo simulation. Then shielded radiation dose difference was confirmed, when 0.25 mmPb shield use for shielding. As a results, shielding ratio was markedly decreased in high energy electromagnetic radiation. The radiation dose was inversely increased with 0.25 mmPb shielding.

  9. Environmental Gamma Radiation Measurements in Baskil District

    International Nuclear Information System (INIS)

    Canbazoglu, C.

    2008-01-01

    In this study, we have determined environmental gamma radiation dose rate in Baskil district which has very high granite content in its geographical structure. Gamma radiation dose rate measurements were achieved by portable radiation monitoring equipment based on the energy range between 40 keV and 1.3 MeV. The measurements were performed on asphalt and soil surface level and also one meter above the ground surface. The gamma dose rate was also performed inside and outside of buildings over the district. The dose rates were found to be between 8.46μR/h and 34.66 μR/h. Indoor and outdoor effective dose rate of the gamma radiation exposure has been calculated to be 523μSv/y and 196μSv/y, respectively

  10. 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)

  11. 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

  12. Studying the ability to use basalt in preparing radiation shielding concrete and the properties of the resulted concrete

    International Nuclear Information System (INIS)

    Alhajali, S.; Yousef, S.; Kanbour, M.; Naoum, B.

    2010-12-01

    Basalt is widespread rocks in the lands of Syria. This kind of rocks has high density relatively, high insulation properties and, mechanical and heat resistance. In this work several kinds of basalt rocks, which were collected from several sites, were studied. The analyses which were done, shows that the basalt rocks collected from Shahba, Nba'a Al-Sakhr and Almana'a mountain are suitable for high efficient gamma radiation shielding, but with low efficiency for neutron shielding, especially for thermal and epithermal neutrons. (author)

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

    International Nuclear Information System (INIS)

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

    1980-01-01

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

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

    Science.gov (United States)

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

    2006-01-01

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

  15. 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.)

  16. Gamma-ray energy buildup factor calculations and shielding effects of some Jordanian building structures

    Science.gov (United States)

    Sharaf, J. M.; Saleh, H.

    2015-05-01

    The shielding properties of three different construction styles, and building materials, commonly used in Jordan, were evaluated using parameters such as attenuation coefficients, equivalent atomic number, penetration depth and energy buildup factor. Geometric progression (GP) method was used to calculate gamma-ray energy buildup factors of limestone, concrete, bricks, cement plaster and air for the energy range 0.05-3 MeV, and penetration depths up to 40 mfp. It has been observed that among the examined building materials, limestone offers highest value for equivalent atomic number and linear attenuation coefficient and the lowest values for penetration depth and energy buildup factor. The obtained buildup factors were used as basic data to establish the total equivalent energy buildup factors for three different multilayer construction styles using an iterative method. The three styles were then compared in terms of fractional transmission of photons at different incident photon energies. It is concluded that, in case of any nuclear accident, large multistory buildings with five layers exterior walls, style A, could effectively attenuate radiation more than small dwellings of any construction style.

  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. 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)

  19. Study of the gamma radiation of ionium

    Energy Technology Data Exchange (ETDEWEB)

    Curie, I

    1949-12-01

    A Geiger counter study has been made of the ..gamma.. radiation of ionium. Eleven quanta of the L radiation of radium were observed for every hundred ..cap alpha.. disintegrations, and three ..gamma.. rays were found with energies of 68, 140, and 240 keV at a rate of 0.85, 0.33, 0.05 quanta, respectively, for 100 disintegrations. It is noted that the radiation spectrum of ionium as a whole is difficult to interpret. In the course of this work, the author calculated the efficiency of a thin-walled aluminum counter, both for the L radiation of radium and for ..gamma.. rays of 68 keV. The author also measured, for soft radiation, the ratio between the efficiency of a thin-walled aluminum counter and that of a similar counter lined with 0.11 mm of lead.

  20. Sources of gamma radiation in a reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Roos, Matts

    1959-05-15

    In a thermal reactor the gamma ray sources of importance for shielding calculations and related aspects are 1) fission, 2) decay of fission products, 3) capture processes in fuel, poison and other materials, 4) inelastic scattering in the fuel and 5) decay of capture products. The energy release and the gamma ray spectra of these sources have been compiled or estimated from the latest information available, and the results are presented in a general way to permit application to any thermal reactor, fueled with a mixture of {sup 235}U and {sup 238}U. As an example the total spectrum and the spectrum of radiation escaping from a fuel rod in the Swedish R3-reactor are presented.

  1. 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

  2. Design and fabrication of shielding for gamma spectrometry

    International Nuclear Information System (INIS)

    Mariano H, E.

    1991-05-01

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

  3. Gamma-ray shielding effect of Gd3+ doped lead barium borate glasses

    Science.gov (United States)

    Kummathi, Harshitha; Naveen Kumar, P.; Vedavathi T., C.; Abhiram, J.; Rajaramakrishna, R.

    2018-05-01

    The glasses of the batch xPbO: 10BaO: (90-x)B2O3: 0.2Gd2O3 (x = 40,45,50 mol %) were prepared by melt-quench technique. The work emphasizes on gamma ray shielding effect on doped lead glasses. The role of Boron is significant as it acts as better neutron attenuator as compared with any other materials, as the thermal neutron cross-sections are high for Gadolinium, 0.2 mol% is chosen as the optimum concentration for this matrix, as higher the concentration may lead to further increase as it produces secondary γ rays due to inelastic neutron scattering. Shielding effects were studied using Sodium Iodide (NaI) - Scintillation Gamma ray spectrometer. It was found that at higher concentration of lead oxide (PbO) in the matrix, higher the attenuation which can be co-related with density. Infra-red (I.R.) spectra reveals that the conversion of Lose triangles to tight tetrahedral structure results in enhancement of shielding properties. The Differential Scanning Calorimeter (D.S.C.) study also reveals that the increase in glass forming range increases the stability which in-turn results in inter-conversion of BO3 to BO4 units such that the density of glass increases with increase in PbO content, resulting in much stable and efficient gamma ray shielding glasses.

  4. 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.

  5. 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

  6. Sterilization plants equipped with the isotopic gamma radiation sources

    International Nuclear Information System (INIS)

    Mehta, K.; Chmielewski, A.G.

    2007-01-01

    Presentation describes different isotopic gamma radiation sources applicable for sterilization of food and medical materials. Certain gamma pallet irradiators, mini gamma irradiators and different scale gamma tote irradiators are presented. It is concluded, that about two hundreds plants with gamma radiation sources operates in different countries. However, industrially developed countries must construct much more plants than operates now

  7. 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

  8. 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...

  9. Investigation of Radiation Protection Methodologies for Radiation Therapy Shielding Using Monte Carlo Simulation and Measurement

    Science.gov (United States)

    Tanny, Sean

    The advent of high-energy linear accelerators for dedicated medical use in the 1950's by Henry Kaplan and the Stanford University physics department began a revolution in radiation oncology. Today, linear accelerators are the standard of care for modern radiation therapy and can generate high-energy beams that can produce tens of Gy per minute at isocenter. This creates a need for a large amount of shielding material to properly protect members of the public and hospital staff. Standardized vault designs and guidance on shielding properties of various materials are provided by the National Council on Radiation Protection (NCRP) Report 151. However, physicists are seeking ways to minimize the footprint and volume of shielding material needed which leads to the use of non-standard vault configurations and less-studied materials, such as high-density concrete. The University of Toledo Dana Cancer Center has utilized both of these methods to minimize the cost and spatial footprint of the requisite radiation shielding. To ensure a safe work environment, computer simulations were performed to verify the attenuation properties and shielding workloads produced by a variety of situations where standard recommendations and guidance documents were insufficient. This project studies two areas of concern that are not addressed by NCRP 151, the radiation shielding workload for the vault door with a non-standard design, and the attenuation properties of high-density concrete for both photon and neutron radiation. Simulations have been performed using a Monte-Carlo code produced by the Los Alamos National Lab (LANL), Monte Carlo Neutrons, Photons 5 (MCNP5). Measurements have been performed using a shielding test port designed into the maze of the Varian Edge treatment vault.

  10. Poster - 11: Radiation barrier thickness calculations for the GammaPod

    International Nuclear Information System (INIS)

    La Russa, Daniel; Vandervoort, Eric; Wilkins, David

    2016-01-01

    A consortium of radiotherapy centers in North America is in the process of evaluating a novel new 60 Co teletherapy device, called the GammaPod™ (Xcision Medical Systems, Columbia Maryland), designed specifically for breast SBRT. The GammaPod consists of 36 collimated 60 Co sources with a total activity of 4320 Ci. The sources are housed in a hemispherical source carrier that rotates during treatment to produce a cylindrically symmetric cone of primary beam spanning 16° – 54° degrees from the horizontal. This unique beam geometry presents challenges when designing or evaluating room shielding for the purposes of meeting regulatory requirements, and for ensuring the safety of staff and the public in surrounding areas. Conventional methods for calculating radiation barrier thicknesses have been adapted so that barrier transmission factors for the GammaPod can be determined from a few relevant distances and characteristics of the primary beam. Simple formalisms have been determined for estimating shielding requirements for primary radiation (with a rotating and non-rotating source carrier), patient-scattered radiation, and leakage radiation. When making worst case assumptions, it was found that conventional barrier thicknesses associated with linac treatment suites are sufficient for shielding all sources of radiation from the GammaPod.

  11. Poster - 11: Radiation barrier thickness calculations for the GammaPod

    Energy Technology Data Exchange (ETDEWEB)

    La Russa, Daniel; Vandervoort, Eric; Wilkins, David [Radiation Medicine Program, The Ottawa Hospital (Canada)

    2016-08-15

    A consortium of radiotherapy centers in North America is in the process of evaluating a novel new {sup 60}Co teletherapy device, called the GammaPod™ (Xcision Medical Systems, Columbia Maryland), designed specifically for breast SBRT. The GammaPod consists of 36 collimated {sup 60}Co sources with a total activity of 4320 Ci. The sources are housed in a hemispherical source carrier that rotates during treatment to produce a cylindrically symmetric cone of primary beam spanning 16° – 54° degrees from the horizontal. This unique beam geometry presents challenges when designing or evaluating room shielding for the purposes of meeting regulatory requirements, and for ensuring the safety of staff and the public in surrounding areas. Conventional methods for calculating radiation barrier thicknesses have been adapted so that barrier transmission factors for the GammaPod can be determined from a few relevant distances and characteristics of the primary beam. Simple formalisms have been determined for estimating shielding requirements for primary radiation (with a rotating and non-rotating source carrier), patient-scattered radiation, and leakage radiation. When making worst case assumptions, it was found that conventional barrier thicknesses associated with linac treatment suites are sufficient for shielding all sources of radiation from the GammaPod.

  12. The radiation shielding potential of CI and CM chondrites

    Science.gov (United States)

    Pohl, Leos; Britt, Daniel T.

    2017-03-01

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

  13. 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.

  14. 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.

  15. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

  17. Radiation shielding design for the VISTA space craft

    Energy Technology Data Exchange (ETDEWEB)

    Pahyn, S.; Pahyn, H.M. [Gazi Univ., Teknik Eoitim Fakultesi, Ankara (Turkey)

    2001-07-01

    An innovative concept for the direct utilisation of fusion energy with laser ignited (D,T) capsules for propulsion is presented with the so called VISTA (Vehicle for Interplanetary Space Transport Applications) concept. VISTA's overall geometry is that of a 50 degrees-half-angle cone to avoid massive radioactive shielding. The 50 degrees-half-angle maximizes the jet efficiency, and is determined by selecting the optimum pellet firing position along the axis of the cone with respect to the plane of the magnet coil. The pellet firing position is in the vacuum. By a total fusion power production of 17 500 MW with a repetition rate of 5 Hz and 3 500 MJ per shot, the propulsion power in form of charged particles has been calculated as {approx} 7 000 MW, making {approx} 40 % of the total fusion power. About 60 % of the fusion energy is carried by the leaking neutrons out of the pellet. Most of them (96 %) escape into vacuum without striking the space ship. Only 4 % enter the frozen hydrogen exhaust cone (about 50 gr.). Total peak nuclear heat generation in the coils is calculated as 4.7 mW/cm{sup 3}. The peak neutron heating is 1.9 mW/cm{sup 3} and the peak {gamma}-ray heating density is 2.8 mW/cm{sup 3}. However, volume averaged nuclear heat generation in the coils is much lower. It is calculated as 0.18, 0.48 and 0.66 mW/cm{sup 3} for neutron, {gamma}-ray and total nuclear heating, respectively. Net shielding mass is found as 170 ton, making < 3 % of the vehicle mass. (authors)

  18. 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...

  19. 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

  20. Some factors affecting urokinase inactivation. [Gamma radiation

    Energy Technology Data Exchange (ETDEWEB)

    Iwata, Hiroo; Iketa, Yoshito

    1985-10-01

    The enzymatic activity of urokinase adsorbed on various polymer surfaces was measured to study the interaction between protein and polymers. The polymer films on which urokinase was adsorbed were exposed to either a high temperature or ..gamma..-radiation. The thermal inactivation rates were higher on hydrophobic polymers such as poly(ethylene terephthalate), nylon 6, and poly(vinylidene fluoride) than hydrophilic polymers like cellulose and ethylene-vinyl alcohol copolymer, indicating their substantial dependence on the interfacial free energy between the polymer and water. A similar dependence was also seen for the ..gamma..-radiation inactivation. Urokinase adsorbed on the hydrophobic polymers lost more easily its enzymatic activity by exposure to ..gamma..-radiation. The interfacial free energy seems to be one of the driving forces to denaturate proteins on polymers.

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

    International Nuclear Information System (INIS)

    1999-01-01

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

  2. Gamma response study of radiation sensitive MOSFETs for their use as gamma radiation sensor

    Energy Technology Data Exchange (ETDEWEB)

    Srivastava, Saurabh; Kumar, A. Vinod [Radiation Safety Systems Division, Bhabha Atomic Research Centre, Mumbai (India); Aggarwal, Bharti; Singh, Arvind; Topkar, Anita, E-mail: anita@barc.gov.in [Electronics Division, Bhabha Atomic Research Centre, Mumbai (India)

    2016-05-23

    Continuous monitoring of gamma dose is important in various fields like radiation therapy, space-related research, nuclear energy programs and high energy physics experiment facilities. The present work is focused on utilization of radiation-sensitive Metal-Oxide-Semiconductor Field Effect Transistors (MOSFETs) to monitor gamma radiation doses. Static characterization of these detectors was performed to check their expected current-voltage relationship. Threshold voltage and transconductance per unit gate to source voltage (K factor) were calculated from the experimental data. The detector was exposed to gamma radiation in both, with and without gate bias voltage conditions, and change in threshold voltage was monitored at different gamma doses. The experimental data was fitted to obtain equation for dependence of threshold voltage on gamma dose. More than ten times increase in sensitivity was observed in biased condition (+3 V) compared to the unbiased case.

  3. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-02-15

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

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

    Directory of Open Access Journals (Sweden)

    Reza Bagheri

    2017-02-01

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

  6. 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.

  7. 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

  8. 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

  9. Neutron Radiation Shielding For The NIF Streaked X-Ray Detector (SXD) Diagnostic

    International Nuclear Information System (INIS)

    Song, P; Holder, J; Young, B; Kalantar, D; Eder, D; Kimbrough, J

    2006-01-01

    The National Ignition Facility (NIF) at Lawrence Livermore National Laboratory (LLNL) is preparing for the National Ignition Campaign (NIC) scheduled in 2010. The NIC is comprised of several ''tuning'' physics subcampaigns leading up to a demonstration of Inertial Confinement Fusion (ICF) ignition. In some of these experiments, time-resolved x-ray imaging of the imploding capsule may be required to measure capsule trajectory (shock timing) or x-ray ''bang-time''. A capsule fueled with pure tritium (T) instead of a deutriun-tritium (DT) mixture is thought to offer useful physics surrogacy, with reduced yields of up to 5e14 neutrons. These measurements will require the use of the NIF streak x-ray detector (SXD). The resulting prompt neutron fluence at the planned SXD location (∼1.7 m from the target) would be ∼1.4e9/cm 2 . Previous measurements suggest the onset of significant background at a neutron fluence of ∼ 1e8/cm 2 . The radiation damage and operational upsets which starts at ∼1e8 rad-Si/sec must be factored into an integrated experimental campaign plan. Monte Carlo analyses were performed to predict the neutron and gamma/x-ray fluences and radiation doses for the proposed diagnostic configuration. A possible shielding configuration is proposed to mitigate radiation effects. The primary component of this shielding is an 80 cm thickness of Polyethylene (PE) between target chamber center (TCC) and the SXD diagnostic. Additionally, 6-8 cm of PE around the detector provide from the large number of neutrons that scatter off the inside of the target chamber. This proposed shielding configuration reduces the high-energy neutron fluence at the SXD by approximately a factor ∼50

  10. Neutron Radiation Shielding For The NIF Streaked X-Ray Detector (SXD) Diagnostic

    Energy Technology Data Exchange (ETDEWEB)

    Song, P; Holder, J; Young, B; Kalantar, D; Eder, D; Kimbrough, J

    2006-11-02

    The National Ignition Facility (NIF) at Lawrence Livermore National Laboratory (LLNL) is preparing for the National Ignition Campaign (NIC) scheduled in 2010. The NIC is comprised of several ''tuning'' physics subcampaigns leading up to a demonstration of Inertial Confinement Fusion (ICF) ignition. In some of these experiments, time-resolved x-ray imaging of the imploding capsule may be required to measure capsule trajectory (shock timing) or x-ray ''bang-time''. A capsule fueled with pure tritium (T) instead of a deutriun-tritium (DT) mixture is thought to offer useful physics surrogacy, with reduced yields of up to 5e14 neutrons. These measurements will require the use of the NIF streak x-ray detector (SXD). The resulting prompt neutron fluence at the planned SXD location ({approx}1.7 m from the target) would be {approx}1.4e9/cm{sup 2}. Previous measurements suggest the onset of significant background at a neutron fluence of {approx} 1e8/cm{sup 2}. The radiation damage and operational upsets which starts at {approx}1e8 rad-Si/sec must be factored into an integrated experimental campaign plan. Monte Carlo analyses were performed to predict the neutron and gamma/x-ray fluences and radiation doses for the proposed diagnostic configuration. A possible shielding configuration is proposed to mitigate radiation effects. The primary component of this shielding is an 80 cm thickness of Polyethylene (PE) between target chamber center (TCC) and the SXD diagnostic. Additionally, 6-8 cm of PE around the detector provide from the large number of neutrons that scatter off the inside of the target chamber. This proposed shielding configuration reduces the high-energy neutron fluence at the SXD by approximately a factor {approx}50.

  11. 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)

  12. 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)

  13. 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

  14. Evaluation of the computer code system RADHEAT-V4 by analysing benchmark problems on radiation shielding

    International Nuclear Information System (INIS)

    Sakamoto, Yukio; Naito, Yoshitaka

    1990-11-01

    A computer code system RADHEAT-V4 has been developed for safety evaluation on radiation shielding of nuclear fuel facilities. To evaluate the performance of the code system, 18 benchmark problem were selected and analysed. Evaluated radiations are neutron and gamma-ray. Benchmark problems consist of penetration, streaming and skyshine. The computed results show more accurate than those by the Sn codes ANISN and DOT3.5 or the Monte Carlo code MORSE. Big core memory and many times I/O are, however, required for RADHEAT-V4. (author)

  15. Effect of Steel Fiber Addition on Mechanical Properties and gamma-Ray Attenuation for Ordinary Concrete Used in El-Gabal El-Akhdar Area in Libya for Radiation Shielding Purposes

    International Nuclear Information System (INIS)

    Ikraiam, F.A.; Ali, J.M.; Abd El-Latif, A.; Abd ELazziz, A.

    2009-01-01

    This work deals with the study of ordinary concrete reinforced with steel fibers where the concrete was prepared from limestone ores as coarse aggregate, and sand as fine aggregate, in El-Gabal El-Akhdar in Libya in order to study some mechanical properties and gamma ray attenuation. For mechanical properties, compressive and tensile strengths have been studied, where tensile strength improved by a factor 1.189 at 3% steel fiber content in comparison with reference sample (0%), whereas the compressive strength improved by a factor of 1.012 at 1% steel fiber content. In relation to the concrete density, it is found that the optimum density was 2.217g/cm3 at 3% where the factor of improvement was 1.1. For gamma ray attenuation, the measurements have been obtained by a collimated beam of gamma ray from sources 60 Co, 22 Na and 137 Cs using a gamma ray spectrometer MCA cassy with inorganic scintillator Na(Tl). The total linear attenuation coefficient (μ cm - 1), mean free path length (λ), HVT (τ1/2) and TVT (τ1/10) are all evaluated in this study. The study reveals that concrete sample with 3% steel fiber content has the optimum values of all these nuclear parameters

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

    International Nuclear Information System (INIS)

    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)

  17. 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

  18. 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.

  19. Radiation transport and shielding information, computer codes, and nuclear data for use in CTR neutronics research and development

    International Nuclear Information System (INIS)

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

    1976-01-01

    The activities of the Radiation Shielding Information Center (RSIC) of the Oak Ridge National Laboratory are being utilized in support of fusion reactor technology. The major activities of RSIC include the operation of a computer-based information storage and retrieval system, the collection, packaging, and distribution of large computer codes, and the compilation and dissemination of processed and evaluated data libraries, with particular emphasis on neutron and gamma-ray cross-section data. The Center has acquired thirteen years of experience in serving fission reactor, weapons, and accelerator shielding research communities, and the extension of its technical base to fusion reactor research represents a logical progression. RSIC is currently working with fusion reactor researchers and contractors in computer code development to provide tested radiation transport and shielding codes and data library packages. Of significant interest to the CTR community are the 100 energy group neutron and 21 energy group gamma-ray coupled cross-section data package (DLC-37) for neutronics studies, a comprehensive 171 energy group neutron and 36 energy group gamma-ray coupled cross-section data base with retrieval programs, including resonance self-shielding, that are tailored to CTR application, and a data base for the generation of energy-dependent atomic displacement and gas production cross sections and heavy-particle-recoil spectra for estimating radiation damage to CTR structural components. Since 1964, the Center has been involved in the international exchange of information, encouraged and supported by both government and interagency agreements; and to achieve an equally viable and successful program in fusion research, the reciprocal exchange of CTR data and computing technology is encouraged and welcomed

  20. Gamma radiation utilization for polymers synthesis

    International Nuclear Information System (INIS)

    Santos, L.G. dos; Silva Filho, E.

    1990-01-01

    Samples of styrene monomer were irradiated with gamma rays from a sup(60)Co source, with doses in the range of zero to 32600 Gray (Gy). Viscosity analysis performed by using the method of Stokes showed an increase in the viscosity of the material. This result characterizes the occurrence of polymerization induced by the radiation. (author)

  1. Quality control of gamma radiation measuring systems

    International Nuclear Information System (INIS)

    Surma, M.J.

    2002-01-01

    The problem of quality control and assurance of gamma radiation measuring systems has been described in detail. The factors deciding of high quality of radiometric measurements as well as statistical testing and calibration of measuring systems have been presented and discussed

  2. 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 ...

  3. Gamma and electron radiation effects on straw

    International Nuclear Information System (INIS)

    Leonhardt, J.W.; Baer, M.; Huebner, G.

    1983-01-01

    Gamma and electron radiation effects on wheat straw, oat straw, barley straw and rye straw are reported. In vitro and in vivo studies show that the digestibility of these agricultural rough materials can be increased up to 80% and more at high doses. The increase of the digestibility is connected with a depolymerisation of cellulose and hemicellulose. (author)

  4. 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)

  5. Polyethylene/boron-containing composites for radiation shielding

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Ji Wook [Center for Materials Architecturing, Institute for Multi-Disciplinary Convergence of Materials, Korea Institute of Science and Technology (KIST), Seoul 136-791 (Korea, Republic of); Department of Chemical and Biological Engineering, Korea University, Seoul 136-701 (Korea, Republic of); Lee, Jang-Woo; Yu, Seunggun; Baek, Bum Ki; Hong, Jun Pyo [Center for Materials Architecturing, Institute for Multi-Disciplinary Convergence of Materials, Korea Institute of Science and Technology (KIST), Seoul 136-791 (Korea, Republic of); Seo, Yongsok [School of Materials Science and Engineering, Seoul National University, Seoul 151-744 (Korea, Republic of); Kim, Woo Nyon [Department of Chemical and Biological Engineering, Korea University, Seoul 136-701 (Korea, Republic of); Hong, Soon Man, E-mail: smhong@kist.re.kr [Center for Materials Architecturing, Institute for Multi-Disciplinary Convergence of Materials, Korea Institute of Science and Technology (KIST), Seoul 136-791 (Korea, Republic of); Nanomaterials Science and Engineering, University of Science and Technology, Daejeon 305-350 (Korea, Republic of); Koo, Chong Min, E-mail: koo@kist.re.kr [Center for Materials Architecturing, Institute for Multi-Disciplinary Convergence of Materials, Korea Institute of Science and Technology (KIST), Seoul 136-791 (Korea, Republic of); Nanomaterials Science and Engineering, University of Science and Technology, Daejeon 305-350 (Korea, Republic of)

    2014-06-01

    Graphical abstract: - Highlights: • HDPE/silane-treated boron nitride (mBN) composites were fabricated. • The HDPE/mBN composites revealed a strong adhesion behavior at the interface of matrix/filler. • The HDPE/mBN composites show superior radiation shielding, thermoconductive and mechanical properties to the composites containing pristine BN and B{sub 4}C fillers. - Abstract: High-density polyethylene (HDPE) composites with modified boron nitride (mBN) fillers, functionalized with an organosilane, were fabricated through conventional melt-extrusion processing techniques. The properties and performances of these composites were compared with those of the composites containing pristine BN and boron carbide (B{sub 4}C) fillers. The silane functionalization of the BN fillers strongly improved the interfacial adhesion between the polymer matrix and the filler. As a result, the HDPE/mBN composites showed a better dispersion state of the filler particles, larger tensile modulus, greater effective thermal conductivity, and better neutron shielding property compared with the HDPE/BN and HDPE/B{sub 4}C composites.

  6. Polyethylene/boron-containing composites for radiation shielding

    International Nuclear Information System (INIS)

    Shin, Ji Wook; Lee, Jang-Woo; Yu, Seunggun; Baek, Bum Ki; Hong, Jun Pyo; Seo, Yongsok; Kim, Woo Nyon; Hong, Soon Man; Koo, Chong Min

    2014-01-01

    Graphical abstract: - Highlights: • HDPE/silane-treated boron nitride (mBN) composites were fabricated. • The HDPE/mBN composites revealed a strong adhesion behavior at the interface of matrix/filler. • The HDPE/mBN composites show superior radiation shielding, thermoconductive and mechanical properties to the composites containing pristine BN and B 4 C fillers. - Abstract: High-density polyethylene (HDPE) composites with modified boron nitride (mBN) fillers, functionalized with an organosilane, were fabricated through conventional melt-extrusion processing techniques. The properties and performances of these composites were compared with those of the composites containing pristine BN and boron carbide (B 4 C) fillers. The silane functionalization of the BN fillers strongly improved the interfacial adhesion between the polymer matrix and the filler. As a result, the HDPE/mBN composites showed a better dispersion state of the filler particles, larger tensile modulus, greater effective thermal conductivity, and better neutron shielding property compared with the HDPE/BN and HDPE/B 4 C composites

  7. 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

  8. 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.)

  9. 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 ).

  10. Layer-splitting technique for testing the recursive scheme for multilayer shields gamma ray buildup factors

    International Nuclear Information System (INIS)

    Alkhatib, Sari F.; Park, Chang Je; Jeong, Hae Yong; Lee, Yongdeok

    2016-01-01

    Highlights: • A simple formalism is suggested for the recursive approach and then it is used to produce buildup factors for certain multilayer shields. • The newly layer-splitting technique is implemented on the studied cases for testing the suggested formalism performance. • The buildup factors are generated using cubic polynomial fitting functions that are produced based on previous well-acknowledge data. - Abstract: This study illustrates the implementation of the newly suggested layer-splitting testing technique. This technique is introduced in order to be implemented in examining suggested formalisms for the recursive scheme (or iterative scheme). The recursive scheme is a concept used in treating and producing the gamma ray buildup factors in the case of multilayer shields. The layer-splitting technique simply enforces the scheme to treat a single layer of one material as two separated layers with similar characteristics. Thus it subjects the scheme to an abnormal definition of the multilayer shield that will test its performance in treating the successive layers. Thus, it will act as a method of verification for the approximations and assumptions taken in consideration. A simple formalism was suggested for the recursive scheme then the splitting technique was implemented on it. The results of implementing both the suggested formalism and the splitting technique are then illustrated and discussed. Throughout this study, cubic polynomial fitting functions were used to generate the data of buildup factors for the basic single-media that constitute the multilayer shields understudy. This study is limited to the cases of multiple shields consisting of repeated consecutive thin layers of lead–water and iron–water shields for 1 MeV gamma rays. The produced results of the buildup factor values through the implementation of the suggested formalism showed good consistency with the Monte Carlo simulation results of Lin and Jiang work. In the implementation of

  11. 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)

  12. Evaluation of nuclear data for radiation shielding by model calculations and international co-operation aspects

    International Nuclear Information System (INIS)

    Canetta, E.; Maino, G.; Menapace, E.

    2001-01-01

    The matter is reviewed, also following previous discussions at ICRS-9, concerning evaluation and related theoretical activities on nuclear data for radiation shielding within the framework of international co-operation initiatives, according to recognised needs and priorities. Both cross-section data.- for reactions induced by neutrons and photons - and nuclear structure data have been considered. In this context, main contributions and typical results are presented from theoretical and evaluation activities at the ENEA Applied Physics Division, especially concerning neutron induced reaction data up to 20 MeV and photonuclear reaction data such as photon absorption and (gamma,n) cross-sections. Relevant aspects of algebraic nuclear models and of evaporation and pre-equilibrium models are discussed. (authors)

  13. 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...

  14. 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...

  15. 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...

  16. Radiation shielding material and method of fabricating the same

    International Nuclear Information System (INIS)

    Nagai, Haruo; Uehara, Hiroshi; Imamura, Katsuji.

    1979-01-01

    Purpose: To provide a radiation shielding material containing lead acrylates, which material is provided with an excellent optical transparency and mechanical strength. Constitution: The material comprises a polymer consisting of a substate monomer selected from the group of (hydroxy) alkyl metacrylate, hydroxyalkyl acrylate and styrene and lead (meta) acrylate, and an/organic acid lead represented by a general formula, (RCOO)sub(a) Pb where a: an integer equivalent to the valency of lead, and R: an unsaturated hydrocarbon group. Furthermore, both substances are caused to be copresent so that the ratio x (weight percentage) of metacrylic acid lead or acrylic acid lead to the entire monomer and the blending ratio y (weight part) of organic acid lead to 100% by weight of the entire monomer satisfy specific conditions. (Aizawa, K.)

  17. 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

  18. Low background gamma ray spectrometer using the anticoincidence shield technique at KAERI

    International Nuclear Information System (INIS)

    Byun, Jong In; Choi, Yun Ho; Kwak, Seung Im; Hwang, Han Yull; Chung, Kun Ho; Choi, Geun Sik; Park, Doo Won; Lee, Chang Woo

    2002-01-01

    We develop a ultra-low background gamma ray spectrometer, using active and passive shielding technique at the same time. Cosmic ray induced background is suppressed by means of active shield devices consisting of plastic scintillating plates of 50 mm thick and anti-coincidence electronic system. The shield is made of 150 mm thick walls of very low activity lead, especially 20 mm with activity of -1 and 0.36 s -1 with and without active shield, respectively, on the regions from 50 keV to 3 MeV. The detection efficiency curve has been precisely measured for regions from 80 keV to 2 MeV with a 10 3 ml marinelli beaker sample, made with calibrated mixed-sources consists of 109 Cd, 57 Co, 139 Ce, 203 Hg, 113 Sn, 85 Sr, 137 Cs, 60 Co and 88 Y. The virtues of the method are demonstrated by applying on experiment that requires the lowest detection limit

  19. 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

  20. 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)

  1. 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)

  2. 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)

  3. 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)

  4. 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)

  5. 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)

  6. 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)

  7. On the omnipresent background gamma radiation of the continuous spectrum

    Energy Technology Data Exchange (ETDEWEB)

    Banjanac, R.; Maletić, D.; Joković, D., E-mail: yokovic@ipb.ac.rs; Veselinović, N.; Dragić, A.; Udovičić, V.; Aničin, I.

    2014-05-01

    The background spectrum of a germanium detector, shielded from the radiations arriving from the lower and open for the radiations arriving from the upper hemisphere, is studied by means of absorption measurements, both in a ground level and in an underground laboratory. The low-energy continuous portion of this background spectrum that peaks at around 100 keV, which is its most intense component, is found to be of very similar shape at the two locations. It is established that it is mostly due to the radiations of the real continuous spectrum, which is quite similar to the instrumental one. The intensity of this radiation is in our cases estimated to about 8000 photons/(m{sup 2}s·2π·srad) in the ground level laboratory, and to about 5000 photons/(m{sup 2}s·2π·srad) in the underground laboratory, at the depth of 25 m.w.e. Simulations by GEANT4 and CORSIKA demonstrate that this radiation is predominantly of terrestrial origin, due to environmental gamma radiations scattered off the materials that surround the detector (the “skyshine radiation”), and to a far less extent to cosmic rays of degraded energy. - Highlights: • We studied the low-energy part of continuous background spectra of germanium detectors. • The study was performed at the ground level and at the shallow underground sites. • The instrumental spectrum is due to radiations of the similar continuous spectrum. • The low-energy radiation is of both terrestrial and cosmic-ray origin. • In our study, we find that this radiation is of predominantly terrestrial origin.

  8. Measurements of TFTR D-T radiation shielding efficiency

    International Nuclear Information System (INIS)

    Kugel, H.W.; Ascione, G.; Elwood, S.; Gilbert, J.; Ku, L.P.; Levine, J.; Rule, K.; Azziz, N.; Goldhagen, P.; Hajnal, F.

    1994-11-01

    Measurements of neutron and gamma dose-equivalents were performed in the Test Cell, at the outer Test Cell wall, in nearby work areas, and out to the nearest property lines at a distance of 180 m. Argon ionization chambers, moderated 3 He proportional counters, and fission chamber detectors were used to obtain measurements of neutron and gamma dose-equivalents per D-T neutron during individual TFTR discharges. These measured neutron and gamma D-T dose-equivalents per TFTR neutron characterize the effects of local variations in material density resulting from the complex asymmetric site geometry. The measured dose-equivalents per TFTR D-T neutron and the cumulative neutron production were used to determine that the planned annual TFTR neutron production of 1 x 10 21 D-T neutrons is consistent with the design objective of limiting the total dose-equivalent at the property line, from all radiation sources and pathways, to less than 10 mrem per year

  9. Set of programs for determining exposure and dose rates from selected sources of gamma radiation

    International Nuclear Information System (INIS)

    Hep, J.; Kralovcova, E.; Smutny, V.; Valenta, V.

    1982-01-01

    The programs are described for the determination of exposure and dose rate of gamma radiation from point, surface, linear and volume sources with and without shielding. The computation is conducted using the classical method taking into consideration the buildup factor. For the computation of the buildup factor in heterogeneous shielding the Broder and Kitazuma formulas are used. Kitazuma's alpha coefficients were calculated recurrently using a new semi-empirical method. Taylor's approximation was used for the calculation of the buildup factor in a single layer

  10. Locating gamma radiation source by self collimating BGO detector system

    Energy Technology Data Exchange (ETDEWEB)

    Orion, I; Pernick, A; Ilzycer, D; Zafrir, H [Israel Atomic Energy Commission, Yavne (Israel). Soreq Nuclear Research Center; Shani, G [Ben-Gurion Univ. of the Negev, Beersheba (Israel)

    1996-12-01

    The need for airborne collimated gamma detector system to estimate the radiation released from a nuclear accident has been established. A BGO detector system has been developed as an array of separate seven cylindrical Bismuth Germanate scintillators, one central detector symmetrically surrounded by six detectors. In such an arrangement, each of the detectors reduced the exposure of other detectors in the array to a radiation incident from a possible specific spatial angle, around file array. This shielding property defined as `self-collimation`, differs the point source response function for each of the detectors. The BGO detector system has a high density and atomic number, and therefore provides efficient self-collimation. Using the response functions of the separate detectors enables locating point sources as well as the direction of a nuclear radioactive plume with satisfactory angular resolution, of about 10 degrees. The detector`s point source response, as function of the source direction, in a horizontal plane, has been predicted by analytical calculation, and was verified by Monte-Carlo simulation using the code EGS4. The detector`s response was tested in a laboratory-scale experiment for several gamma ray energies, and the experimental results validated the theoretical (analytical and Monte-Carlo) results. (authors).

  11. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-11

    Gamma ray shielding behaviour of any material can be studied by various interaction parameters such as total mass attenuation coefficient (μ{sub m}); half value layer (HVL); tenth value layer (TVL); effective atomic number (Z{sub eff}), electron density (N{sub el}), effective atomic weight (A{sub eff}) and buildup factor. For gamma rays, the accurate measurements of μ{sub m} (cm{sup 2} g{sup −1}) theoretically require perfect narrow beam irradiation geometry. However, the practical geometries used for the experimental investigations deviate from perfect-narrowness thereby the multiple scattered photons cause systematic errors in the measured values of μ{sub m}. Present investigation is an attempt to find the optimum value of absorber thickness (low-Z) for which these errors are insignificant and acceptable. Both experimental and theoretical calculations have been performed to investigate the effect of absorber's thickness on μ{sub m} of six low-Z (10gamma-ray energies 661.66 keV, 1173.24 keV and 1332.50 keV. A computer program (GRIC2-toolkit) was designed for theoretical evaluation of shielding parameters of any material. Good agreement of theoretical and measured values of μ{sub m} was observed for all absorbers with thickness ≤0.5 mean free paths, thus considered it as optimum thickness for low-Z materials in the selected energy range. White cement was found to possess maximum shielding effectiveness for the selected gamma rays. - Highlights: • Optimum thickness value is 0.5 mfp for low-Z absorbers in energy range 662–1332 keV. • For accurate measurement of μ{sub m} absorber's thickness should be ≤optimum thickness. • GRIC2-toolkit is useful for γ-ray shielding analysis of composite materials.

  13. Effect of molybdenum on gamma ray shielding and structural properties of PbO-B2O3 glasses

    Science.gov (United States)

    Dogra, Mridula; Singh, K. J.; Kaur, Kulwinder

    2018-04-01

    The present study is aimed at developing new shielding materials for gamma ray shielding applications. Transparent glasses of the composition xMoO3-0.7PbO-(0.3-x)B2O3 where x= 0.03 to 0. 06 (mole fraction) have been prepared by using melt-quenchingtechnique. Gamma ray shielding properties have been evaluated in terms of mass attenuation coefficient and half value layer parameter at photon energies 662 and 1173 keV. These shielding parameters are also compared with standard shielding material`concretes'. It has been found that prepared glass system shows better shielding properties than barite and ordinary concretes proving the possibility of its usage as an alternate to conventional concrete for gamma ray shielding applications. The density, molar volume, X-Ray Diffraction, Fourier Transform InfraRed and Raman studies have been performed to study the structural properties of the glass system. It has been analyzed from FTIR and Raman studies that bridging oxygens increase with the decrease of MoO3 content in the glass composition.

  14. Application of the personnel photographic monitoring method to determine equivalent radiation dose beyond proton accelerator shielding

    International Nuclear Information System (INIS)

    Gel'fand, E.K.; Komochkov, M.M.; Man'ko, B.V.; Salatskaya, M.I.; Sychev, B.S.

    1980-01-01

    Calculations of regularities to form radiation dose beyond proton accelerator shielding are carried out. Numerical data on photographic monitoring dosemeter in radiation fields investigated are obtained. It was shown how to determine the total equivalent dose of radiation fields beyond proton accelerator shielding by means of the photographic monitoring method by introduction into the procedure of considering nuclear emulsions of division of particle tracks into the black and grey ones. A comparison of experimental and calculational data has shown the applicability of the used calculation method for modelling dose radiation characteristics beyond proton accelerator shielding [ru

  15. Measuring element for the detection and determination of radiation doses of gamma radiation and neutrons

    International Nuclear Information System (INIS)

    Jahn, W.; Piesch, E.

    1975-01-01

    A measuring element detects and proves both gamma and neutron radiation. The element includes a photoluminescent material which stores gamma radiation and particles of arsenic and phosphorus embedded in the photoluminescent material for detecting neutron radiation. (U.S.)

  16. 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)

  17. 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

  18. 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.

  19. 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.

  20. 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.

  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. 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)

  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. Effect of gamma radiation on polyvinylpyrrolidone hydrogels

    International Nuclear Information System (INIS)

    Oliveira, M.J.A.; Vásquez, P.A.S.; Alcântara, M.T.S.; Munhoz, M.M.L.; Lugão, A.B.

    2017-01-01

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Zakaria, M.; Dekhtyar, Y.; Bogucharska, T.; Noskov, V. [Riga Technical Univ., Biomedical Engineering and Nanotechnology Institute (Latvia)

    2006-07-01

    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)

  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. Gamma radiation sterilises bacteria-contaminated waste

    International Nuclear Information System (INIS)

    Anon.

    1978-01-01

    The use of gamma radiation for the sterilisation of sewage and hospital waste etc., is briefly described. A sterilisation plant delivered by Sulzer is illustrated diagrammatically and its functioning explained, while a photograph illustrates a similar plant delivered by Geodel Systems, of Canada. The latter firm has adapted this type of plant for the radiolytical destruction of phenols, cyanides, alkyl benzenesulphonates and similar wastes. (JIW)

  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. Modification of genetic effects of gamma radiation by laser radiation

    International Nuclear Information System (INIS)

    Khotyljova, L.V.; Khokhlova, S.A.; Khokhlov, I.V.

    1988-01-01

    Full text: Mutants obtained by means of ionizing radiation and chemical mutagens often show low viability and productivity that makes their use in plant breeding difficult. Methods reducing the destructive mutagen action on important functions of plant organism and increasing quality and practical value of induced mutants would be interesting. We believe that one method for increasing efficiency of experimental mutagenesis in plants is the application of laser radiation as a modificator of genetic effects of ionizing radiation and chemical mutagens. Combined exposure of wheat seedlings to a gamma radiation dose of 2 kR and to laser radiation with the wave length of 632.8 nm (power density - 20 mVt/cm 2 , exposure - 30 min.) resulted in reducing the chromosomal aberration percentage from 30.5% in the gamma version to 16.3% in the combined treatment version. A radiosensibilizing effect was observed at additional exposure of gamma irradiated wheat seeds to laser light with the wave length of 441.6 nm where chromosomal aberration percentage increased from 22% in the gamma-irradiation version to 31% in the combined treatment version. By laser radiation it is also possible to normalize mitotic cell activity suppressed by gamma irradiation. Additional seedling irradiation with the light of helium-neon laser (632.8 nm) resulted in recovery of mitotic cell activity from 21% to 62% and increasing the average content of DNA per nucleus by 10%. The influence of only laser radiation on plant variability was also studied and it was shown that irradiation of wheat seeds and seedlings with pulsed and continuous laser light of visible spectrum resulted in phenotypically altered forms in M 2 . Their frequencies was dependent upon power density, dose and radiation wave length. Number of altered forms increased in going from long-wave to short-wave spectrum region. In comparing efficiency of different laser types of pulsed and continuous exposure (dose - 180 J/cm 2 ) 2% of altered

  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

    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

  13. Bibliography, subject index, and author index of the literature examined by the Radiation Shielding Information Center (Reactor and Weapons Radiation Shielding)

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  15. 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

  16. The Study of Radiation of Gamma-Ray Background at Sedimentology Laboratorium, P3TIR, BATAN, Using Gamma Spectrometry

    International Nuclear Information System (INIS)

    Lubis, Ali Arman; Aliyanta, Barokah; Darman

    2002-01-01

    The measurement of background radiation of gamma-ray has been done at Sedimentology Laboratory, SDAL building, P3TIR, BATAN using gamma spectrometer. The measurement was done without shielding with the range of energy between 50 keV and 1500 keV. The identified radiations are coming from environmental radionuclide and man-made radionuclide as well with 32 energy peaks. The environmental radionuclides are from Uranium series, Thorium series, and 4 0 K having dose rate of 12.510 ± O.980, 36.408 ± 3.243, 9.455 ±O.016 n Sv/day, respectively, whilst man-made radionuclide is 6 O C o having dose rate of O.136 ±O.078 n Sv/day

  17. Contribution of 210Pb bremsstrahlung to the background of lead shielded gamma spectrometers

    International Nuclear Information System (INIS)

    Mrda, D.; Bikit, I.; Veskovic, M.; Forkapic, S.

    2007-01-01

    Lead, which is often used as a shielding material, contains 210 Pb (T 1/2 =22.3 y). The 46.54 keV γ-intensity of 210 Pb can be easily reduced by an inner lining, but the bremsstrahlung caused by the β-decay of its daughter, 210 Bi, with a maximal electron energy of 1.16 MeV, will contribute to the gamma detector background. The spectrum of this bremsstrahlung is calculated by numerically fitting the β-spectrum and integrating the Koch-Motz formula. The absorption of the bremsstrahlung in the lead and detection efficiencies for the HPGe detector are calculated by the effective solid angle algorithm, using corrections for the photopeak/Compton ratio of cross-sections in Ge. By comparison with the measured background spectrum, it is shown that, for the lead with 25 Bq/kg of 210 Pb up to 500 keV of gamma spectrum, the bremsstrahlung contribution to the background is about 20% for our surface-based detector system. Also, we compared our calculations with a Monte Carlo simulation of another detector system with a shield containing 1 Bq/kg of 210 Pb and found that our analytical method gives a value of roughly two times higher than the Monte Carlo one for the total bremsstrahlung contribution. The quality of the analytical semi-empirical method is proved by the reasonable agreement with the experimental results published

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

    Energy Technology Data Exchange (ETDEWEB)

    Mariano H, E

    1991-05-15

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

  19. Investigation of novel composite material based on extra-heavy concrete and basalt fiber for gamma radiation protection properties

    International Nuclear Information System (INIS)

    Romanenko, Yi.M.; Nosovs'kij, A.V.; Gulyik, V.Yi.; Golyuk, M.Yi.

    2018-01-01

    The paper presents a new composite material for radiation protection based on extra-heavy concrete reinforced by basalt fiber. Basalt fiber is a new material for concrete reinforcement, which provides improved mechanical characteristics of concrete, reduces the level of microcracks and increases the durability of concrete. Within the scope of present work, the gamma-ray radiation protection properties of concrete reinforced with basalt fiber was modeled. Two types of extra-heavy concrete were used for this paper. The main gamma-ray attenuation coefficients such as mean atomic number, mean atomic mass, mean electron density, effective atomic number, effective electron density, Murty effective atomic number were analyzed with help of WinXCom software. It has been shown that the addition of basalt fiber to concrete does not impair its gamma-ray radiation shielding properties. With increasing the basalt fiber dosage in concrete, the radiation properties against gamma radiation are improved.

  20. A study on radiation shielding and safety analysis for a synchrotron radiation beamline

    International Nuclear Information System (INIS)

    Asano, Yoshihiro

    2001-03-01

    Methods of shielding design and safety analysis are presented for a beam-line of synchrotron radiation. This paper consists of the shielding and safety study of synchrotron radiation with extremely intense and low energy photon below several hundreds keV, and the study for the behavior of remarkable high-energy photons up to 8 GeV, which can creep into beam-lines. A new shielding design code, STAC8 was developed to estimate the leakage dose outside the beam line hutch (an enclosure of the beam, optical elements or experimental instruments) easily and quickly with satisfactory accuracy. The code can calculate consistently from sources of synchrotron radiation to dose equivalent outside hutches with considering the build up effect and polarization effect. Validity of the code was verified by comparing its calculations with those of Monte Carlo simulations and measurement results of the doses inside the hutch of the BL14C of Photon Factory in the High Energy Accelerator Research Organization (KEK), showing good agreements. The shielding design calculations using STAC8 were carried out to apply to the practical beam-lines with the considering polarization effect and clarified the characteristics of the typical beam-line of the third generation synchrotron radiation facility, SPring-8. In addition, the shielding calculations were compared with the measurement outside the shield wall of the bending magnet beam-line of SPring-8, and showed fairly good agreement. The new shielding problems, which have usually been neglected in shielding designs for existing synchrotron radiation facilities, are clarified through the analysis of the beam-line shielding of SPring-8. The synchrotron radiation from the SPring-8 has such extremely high-intensity involving high energy photons that the scattered synchrotron radiation from the concrete floor of the hutch, the ground shine, causes a seriously high dose. The method of effective shielding is presented. For the estimation of the gas

  1. A study on radiation shielding and safety analysis for a synchrotron radiation beamline

    Energy Technology Data Exchange (ETDEWEB)

    Asano, Yoshihiro [Japan Atomic Energy Research Inst., Kansai Research Establishment, Synchrotron Radiation Research Center, Mikazuhi, Hyogo (Japan)

    2001-03-01

    Methods of shielding design and safety analysis are presented for a beam-line of synchrotron radiation. This paper consists of the shielding and safety study of synchrotron radiation with extremely intense and low energy photon below several hundreds keV, and the study for the behavior of remarkable high-energy photons up to 8 GeV, which can creep into beam-lines. A new shielding design code, STAC8 was developed to estimate the leakage dose outside the beam line hutch (an enclosure of the beam, optical elements or experimental instruments) easily and quickly with satisfactory accuracy. The code can calculate consistently from sources of synchrotron radiation to dose equivalent outside hutches with considering the build up effect and polarization effect. Validity of the code was verified by comparing its calculations with those of Monte Carlo simulations and measurement results of the doses inside the hutch of the BL14C of Photon Factory in the High Energy Accelerator Research Organization (KEK), showing good agreements. The shielding design calculations using STAC8 were carried out to apply to the practical beam-lines with the considering polarization effect and clarified the characteristics of the typical beam-line of the third generation synchrotron radiation facility, SPring-8. In addition, the shielding calculations were compared with the measurement outside the shield wall of the bending magnet beam-line of SPring-8, and showed fairly good agreement. The new shielding problems, which have usually been neglected in shielding designs for existing synchrotron radiation facilities, are clarified through the analysis of the beam-line shielding of SPring-8. The synchrotron radiation from the SPring-8 has such extremely high-intensity involving high energy photons that the scattered synchrotron radiation from the concrete floor of the hutch, the ground shine, causes a seriously high dose. The method of effective shielding is presented. For the estimation of the gas

  2. A comparison of dose savings of lead and lightweight aprons for shielding of 99m-Technetium radiation

    International Nuclear Information System (INIS)

    Warren-Forward, H.; Cardew, P.; Smith, B.; Clack, L.; McWhirter, K.; Johnson, S.; Wessel, K.

    2007-01-01

    Nuclear medicine technologists (NMTs) have the highest effective doses of radiation among medical workers. With increase in the use of lightweight materials in diagnostic radiography, the aim was to compare the effectiveness of lead and lightweight aprons in shielding from 99m-Technetium ( 99m Tc) gamma rays. The doses received from a scattering phantom to the entrance, 9 cm depth and exit of a phantom were measured with LiF:Mg, Cu, P thermoluminescent dosemeters (TLDs). Doses and spectra were assessed without no shielding, with 0.5-mm lead and lightweight aprons. The lead and lightweight aprons decreased entrance surface doses by 76 and 59%, respectively. The spectral analysis showed that the lightweight apron provided better dose reduction at energies 99m Tc labelled radiopharmaceutical. (authors)

  3. Application of Interval Predictor Models to Space Radiation Shielding

    Science.gov (United States)

    Crespo, Luis G.; Kenny, Sean P.; Giesy,Daniel P.; Norman, Ryan B.; Blattnig, Steve R.

    2016-01-01

    This paper develops techniques for predicting the uncertainty range of an output variable given input-output data. These models are called Interval Predictor Models (IPM) because they yield an interval valued function of the input. This paper develops IPMs having a radial basis structure. This structure enables the formal description of (i) the uncertainty in the models parameters, (ii) the predicted output interval, and (iii) the probability that a future observation would fall in such an interval. In contrast to other metamodeling techniques, this probabilistic certi cate of correctness does not require making any assumptions on the structure of the mechanism from which data are drawn. Optimization-based strategies for calculating IPMs having minimal spread while containing all the data are developed. Constraints for bounding the minimum interval spread over the continuum of inputs, regulating the IPMs variation/oscillation, and centering its spread about a target point, are used to prevent data over tting. Furthermore, we develop an approach for using expert opinion during extrapolation. This metamodeling technique is illustrated using a radiation shielding application for space exploration. In this application, we use IPMs to describe the error incurred in predicting the ux of particles resulting from the interaction between a high-energy incident beam and a target.

  4. 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.

  5. Performance study and influence of radiation emission energy and soil contamination level on γ-radiation shielding of stabilised/solidified radionuclide-polluted soils

    International Nuclear Information System (INIS)

    Falciglia, Pietro P.; Puccio, Valentina; Romano, Stefano; Vagliasindi, Federico G.A.

    2015-01-01

    This work focuses on the stabilisation/solidification (S/S) of radionuclide-polluted soils at different 232 Th levels using Portland cement alone and with barite aggregates. The potential of S/S was assessed applying a full testing protocol and calculating γ-radiation shielding (γRS) index, that included the measurement of soil radioactivity before and after the S/S as a function of the emission energy and soil contamination level. The results indicate that setting processes are strongly dependent on the contaminant concentration, and for contamination level higher than 5%, setting time values longer than 72 h. The addition of barite aggregates to the cement gout leads to a slight improvement of the S/S performance in terms of durability and contaminant leaching but reduces the mechanical resistance of the treated soils samples. Barite addition also causes an increase in the γ-rays shielding properties of the S/S treatment up to about 20%. Gamma-ray measurements show that γRS strongly depends on the energy, and that the radioactivity with the contamination level was governed by a linear trend, while, γRS index does not depend on the radionuclide concentration. Results allow the calculated γRS values and those available from other experiments to be applied to hazard radioactive soil contaminations. - Highlights: • We assess the effects of 232 Th contamination on performance of S/S treated soil. • We assess the γ-radiation shielding of the S/S materials as a function of energy. • We report a full testing protocol for assessing S/S resistance performance. • Emission energy influences the γ radiation shielding of the S/S. • Barite gives high γ-radiation shielding and low contaminant leaching

  6. 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.

  7. Analysis of the ITER computational shielding benchmark with the Monte Carlo TRIPOLI-4{sup ®} neutron gamma coupled calculations

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yi-Kang, E-mail: yi-kang.lee@cea.fr

    2016-11-01

    Highlights: • Verification and validation of TRIPOLI-4 radiation transport calculations for ITER shielding benchmark. • Evaluation of CEA-V5.1.1 and FENDL-3.0 nuclear data libraries on D–T fusion neutron continuous energy transport calculations. • Advances in nuclear analyses for nuclear heating and radiation damage in iron. • This work also demonstrates that the “safety factors” concept is necessary in the nuclear analyses of ITER. - Abstract: With the growing interest in using the continuous-energy TRIPOLI-4{sup ®} Monte Carlo radiation transport code for ITER applications, a key issue that arises is whether or not the released TRIPOLI-4 code and its associated nuclear data libraries are verified and validated for the D–T fusion neutronics calculations. Previous published benchmark results of TRIPOLI-4 code on the ITER related activities have concentrated on the first wall loading, the reactor dosimetry, the nuclear heating, and the tritium breeding ratio. To enhance the TRIPOLI-4 verification and validation on neutron-gamma coupled calculations for fusion device application, the computational ITER shielding benchmark of M. E. Sawan was performed in this work by using the 2013 released TRIPOLI-4.9S code and the associated CEA-V5.1.1 data library. First wall, blanket, vacuum vessel and toroidal field magnet of the inboard and outboard components were fully modelled in this 1-D toroidal cylindrical benchmark. The 14.1 MeV source neutrons were sampled from a uniform isotropic distribution in the plasma zone. Nuclear responses including neutron and gamma fluxes, nuclear heating, and material damage indicator were benchmarked against previous published results. The capabilities of the TRIPOLI-4 code on the evaluation of above physics parameters were presented. The nuclear data library from the new FENDL-3.0 evaluation was also benchmarked against the CEA-V5.1.1 results for the neutron transport calculations. The results show that both data libraries

  8. Analysis of the ITER computational shielding benchmark with the Monte Carlo TRIPOLI-4® neutron gamma coupled calculations

    International Nuclear Information System (INIS)

    Lee, Yi-Kang

    2016-01-01

    Highlights: • Verification and validation of TRIPOLI-4 radiation transport calculations for ITER shielding benchmark. • Evaluation of CEA-V5.1.1 and FENDL-3.0 nuclear data libraries on D–T fusion neutron continuous energy transport calculations. • Advances in nuclear analyses for nuclear heating and radiation damage in iron. • This work also demonstrates that the “safety factors” concept is necessary in the nuclear analyses of ITER. - Abstract: With the growing interest in using the continuous-energy TRIPOLI-4 ® Monte Carlo radiation transport code for ITER applications, a key issue that arises is whether or not the released TRIPOLI-4 code and its associated nuclear data libraries are verified and validated for the D–T fusion neutronics calculations. Previous published benchmark results of TRIPOLI-4 code on the ITER related activities have concentrated on the first wall loading, the reactor dosimetry, the nuclear heating, and the tritium breeding ratio. To enhance the TRIPOLI-4 verification and validation on neutron-gamma coupled calculations for fusion device application, the computational ITER shielding benchmark of M. E. Sawan was performed in this work by using the 2013 released TRIPOLI-4.9S code and the associated CEA-V5.1.1 data library. First wall, blanket, vacuum vessel and toroidal field magnet of the inboard and outboard components were fully modelled in this 1-D toroidal cylindrical benchmark. The 14.1 MeV source neutrons were sampled from a uniform isotropic distribution in the plasma zone. Nuclear responses including neutron and gamma fluxes, nuclear heating, and material damage indicator were benchmarked against previous published results. The capabilities of the TRIPOLI-4 code on the evaluation of above physics parameters were presented. The nuclear data library from the new FENDL-3.0 evaluation was also benchmarked against the CEA-V5.1.1 results for the neutron transport calculations. The results show that both data libraries can be

  9. Improvement of BaO:B2O3:Fly ash glasses: Radiation shielding, physical and optical properties

    International Nuclear Information System (INIS)

    Tuscharoen, S.; Kaewkhao, J.; Limkitjaroenporn, P.; Limsuwan, P.; Chewpraditkul, W.

    2012-01-01

    Highlights: ► BaO:B 2 O 3 :Fly ash glasses have been improved in radiation Shielding, physical and optical properties. ► The visible light transmission of RHA glass was better than SiO 2 . ► At all BaO concentrations, exhibited the better half values layer in comparison window and ordinary concrete. -- Abstract: Rice husk ash glass (RHA-glass) of composition xBaO:(80 − x)B 2 O 3 :20RHA where x = 45, 50, 55, 60, 65 and 70 wt.% have been prepared using melt-quenching method and investigated on their optical, physical and gamma-rays shielding properties. The densities of these glass samples were increased with increasing of BaO content, due to higher molecular weight of BaO comparing with B 2 O 3 . The molar volume of these glasses was increased with increasing content of BaO; BaO acts as modifier to increase the loose packing. The visible light transmission of RHA glass was better than SiO 2 glass prepared in same formula and preparing condition. The experimental values of gamma ray shielding properties such as; mass attenuation coefficients, atomic cross sections and effective atomic numbers, were found in good agreement with the theoretical values as calculated from WinXCom. Moreover the glass system at all BaO concentrations, exhibited the better half values layer in comparison window and ordinary concrete.

  10. 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.

  11. Portable neutron and gamma-radiation instruments

    International Nuclear Information System (INIS)

    Murray, W.S.; Butterfield, K.B.

    1990-01-01

    This paper reports on the design and building of a smart neutron and gamma-radiation detection systems with embedded microprocessors programmed in the FORTH language. These portable instruments can be battery-powered and can provide many analysis functions not available in most radiation detectors. Local operation of the instruments is menu-driven through a graphics liquid crystal display and hex keypad; remote operation is through a serial communications link. While some instruments simply count particles, others determine the energy of the radiation as well as the intensity. The functions the authors have provided include absolute source-strength determination. Feynmann variance analysis, sequential-probability ratio test, and time-history recording

  12. Origin of the diffuse background gamma radiation

    International Nuclear Information System (INIS)

    Stecker, F.W.; Puget, J.L.

    1974-05-01

    Recent observations have now provided evidence for diffuse background gamma radiation extending to energies beyond 100 MeV. There is some evidence of isotropy and implied cosmological origin. Significant features in the spectrum of this background radiation were observed which provide evidence for its origin in nuclear processes in the early stages of the big-band cosmology and tie in these processes with galaxy formation theory. A crucial test of the theory may lie in future observations of the background radiation in the 100 MeV to 100 GeV energy range which may be made with large orbiting spark-chamber satellite detectors. A discussion of the theoretical interpretations of present data, their connection with baryon symmetric cosmology and galaxy formation theory, and the need for future observations are given. (U.S.)

  13. 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

  14. Dose control for external radiation; Kawalan dos untuk sinaran luar

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1988-12-31

    The chapter briefly discussed the following subjects: time and distance limitation, shielding, effects of radiations i.e. alpha particles, beta particles, x-ray, gamma ray, neutron. half value layer, design of shields and operation of shielding.

  15. Radiation dose reduction at a price: the effectiveness of a thyroid shield during head CT scanning

    International Nuclear Information System (INIS)

    Fu Qiang; Lu Tao; Zhang Ling

    2008-01-01

    Objective: To assess radiation dose to the thyroid in patients undergoing head CT scanning and to evaluate dose reduction to the thyroid by load shielding. Methods: A post-morterm was scanned by different model and study was undertaken to evaluate the dose reduction by thyroid lead shields and assess their practicality in a clinical setting. (a)No thyroid shields and (b) thyroid shield. One thermoluminescent dosimeters (TLDs)were placed over the thyroid gland center, A thyroid lead shield (Pb eq 0.5mm)was placed around the neck of post-morterm. Scan parameter, CTDIw and DLP were recorded. Results: (a) 0.207mSv; (b) 0.085mSv. A mean effective radiation dose reduction of 58% was seen in the shielded versus the unshielded. Conclusion: Thyroid exposure to scattered radiation from head CT scanning only once is associated with a low but not negligible risk of cancer, but accumulatived doses to the thyroid are serious, highlighting the need for increased awareness of patient radiation protection. Thyroid lead shielding yields significant radiation protection, which should be used routinely during head CT scan. (authors)

  16. Measurement of concentrations of {gamma}-ray emitters induced in the concrete shield of the JAERI electron linac facility

    Energy Technology Data Exchange (ETDEWEB)

    Endo, Akira; Kawasaki, Katsuya; Kikuchi, Masamitsu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Harada, Yasunori

    1997-07-01

    Measurement has been made to study distributions of {gamma}-ray emitters induced in the concrete shield of the JAERI electron linac facility. Core boring was carried out at seven positions to take samples from the concrete shield, and {gamma}-ray counting rates and {gamma}-ray spectra of these samples were measured with a NaI(Tl) detector and a Ge semiconductor detector, respectively. The following radionuclides were detected in the concrete samples: {sup 60}Co, {sup 134}Cs, {sup 152}Eu and {sup 154}Eu generated through thermal neutron capture reaction, and {sup 22}Na and {sup 54}Mn generated through nuclear reactions by bremsstrahlung and fast neutrons. The relation between the distributions of {gamma}-ray emitters, as a function of the depth of concrete, and the positions of core boring is discussed. (author)

  17. Degradation of polycarbonate induced by gamma radiation

    International Nuclear Information System (INIS)

    Araujo, E.S. de; Guedes, S.M.L.

    1992-01-01

    Polycarbonate (PC) DUROLON amorphous, of molecular weight 22000 g/mol is used in medical supplies and may be sterilized by gamma radiation. The main chain scission and polymer degradation occur when this polymer is irradiated. The value G = 1.54 to DUROLON was obtained by equation: 10 6 /M v = 10 6 /M v' + 0.054 G R. The degradation without crosslinking it is not general rule to all types of polycarbonates; an comparison was realized. The infrared (FT-IR) spectra of irradiated PC by gamma rays with different doses showed the main chain scissions in carbonyl groups. The mechanism of polymeric degradation to DUROLON, observed by NMR spectra, is a recombination of phenoxy and phenyl radicals. (author)

  18. 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

  19. 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...

  20. 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...

  1. 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...

  2. 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...

  3. 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

  4. New applications and developments in the neutron shielding

    Directory of Open Access Journals (Sweden)

    Uğur Fatma Aysun

    2017-01-01

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

  5. New applications and developments in the neutron shielding

    Science.gov (United States)

    Uğur, Fatma Aysun

    2017-09-01

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

  6. 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)

  7. Shielding practice

    International Nuclear Information System (INIS)

    Sauermann, P.F.

    1985-08-01

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

  8. Progress Toward Electrostatic Radiation Shielding of Interplanetary Spacecraft: Strategies, Concepts and Technical Challenges of Human Exploration Beyond Low Earth Orbit

    Science.gov (United States)

    Metzger, Philip T.; Lane, John E.; Youngquist, Robert C.

    2004-01-01

    The radiation problem is a serious obstacle to solar system exploration. Electrostatic shielding was previously dismissed as unworkable. This was based on the false assumption that radial symmetry is needed to provide isotropic protection. KSC recently demonstrated the feasibility of asymmetric, multipole electrostatic shielding. Combined with passive shielding it might solve the radiation problem

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    Science.gov (United States)

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

    2005-01-01

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

  11. Evaluation of the performance of peridotite aggregates for radiation shielding concrete

    International Nuclear Information System (INIS)

    Wang, Jinjun; Li, Guofeng; Meng, Dechuan

    2014-01-01

    Highlights: • Using peridotite rich in crystal water as aggregates of radiation-shielding concrete. • Performance of peridotite concrete is simulated and compared with ordinary concrete. • Performance of concrete samples is tested. • Neutron shielding performance can be significantly enhanced by peridotite aggregates. - Abstract: Peridotite is a kind of material that is rich in crystal water. In this paper, peridotite is used as fine and coarse aggregates for radiation shielding concrete. The transmission data of different concrete thickness and different energy neutron are calculated using Monte-Carlo method. The neutron shielding performance of the peridotite concrete samples are tested using 241 Am-Be neutron source. The results show that the peridotite is an excellent neutron shielding material

  12. Gamma Radiation Effect on Titan Yellow Solutions

    International Nuclear Information System (INIS)

    El-Banna, M.

    2012-01-01

    In the present work, the radiation induced color bleaching of Titan yellow dye (TY) in different solvents has been studied. The color bleaching of the dye solutions upon irradiation was followed spectrophotometrically. The % color bleaching of the dyes in different solvent systems was plotted against different gamma irradiation doses used and was determined and the obtained relationships were found to be linear in most cases. These relationships were used as calibration curves to determine the unknown irradiation dose. The results obtained were reproducible and showed differences from calculated values ranging from 10 % to 15 %

  13. Study and application of Dot 3.5 computer code in radiation shielding problems

    International Nuclear Information System (INIS)

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

    1983-01-01

    The application of nuclear transportation code S sub(N), Dot 3.5, to radiation shielding problems is revised. Aiming to study the better available option (convergence scheme, calculation mode), of DOT 3.5 computer code to be applied in radiation shielding problems, a standard model from 'Argonne Code Center' was selected and a combination of several calculation options to evaluate the accuracy of the results and the computational time was used, for then to select the more efficient option. To illustrate the versatility and efficacy in the application of the code for tipical shielding problems, the streaming neutrons calculation along a sodium coolant channel is ilustrated. (E.G.) [pt

  14. 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

  15. 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)

  16. Experimental shielding evaluation of the radiation protection provided by the structurally significant components of residential structures.

    Science.gov (United States)

    Dickson, E D; Hamby, D M

    2014-03-01

    The human health and environmental effects following a postulated accidental release of radioactive material to the environment have been a public and regulatory concern since the early development of nuclear technology. These postulated releases have been researched extensively to better understand the potential risks for accident mitigation and emergency planning purposes. The objective of this investigation is to provide an updated technical basis for contemporary building shielding factors for the US housing stock. Building shielding factors quantify the protection from ionising radiation provided by a certain building type. Much of the current data used to determine the quality of shielding around nuclear facilities and urban environments is based on simplistic point-kernel calculations for 1950s era suburbia and is no longer applicable to the densely populated urban environments realised today. To analyse a building's radiation shielding properties, the ideal approach would be to subject a variety of building types to various radioactive sources and measure the radiation levels in and around the building. While this is not entirely practicable, this research analyses the shielding effectiveness of ten structurally significant US housing-stock models (walls and roofs) important for shielding against ionising radiation. The experimental data are used to benchmark computational models to calculate the shielding effectiveness of various building configurations under investigation from two types of realistic environmental source terms. Various combinations of these ten shielding models can be used to develop full-scale computational housing-unit models for building shielding factor calculations representing 69.6 million housing units (61.3%) in the United States. Results produced in this investigation provide a comparison between theory and experiment behind building shielding factor methodology.

  17. Experimental shielding evaluation of the radiation protection provided by the structurally significant components of residential structures

    International Nuclear Information System (INIS)

    Dickson, E D; Hamby, D M

    2014-01-01

    The human health and environmental effects following a postulated accidental release of radioactive material to the environment have been a public and regulatory concern since the early development of nuclear technology. These postulated releases have been researched extensively to better understand the potential risks for accident mitigation and emergency planning purposes. The objective of this investigation is to provide an updated technical basis for contemporary building shielding factors for the US housing stock. Building shielding factors quantify the protection from ionising radiation provided by a certain building type. Much of the current data used to determine the quality of shielding around nuclear facilities and urban environments is based on simplistic point-kernel calculations for 1950s era suburbia and is no longer applicable to the densely populated urban environments realised today. To analyse a building’s radiation shielding properties, the ideal approach would be to subject a variety of building types to various radioactive sources and measure the radiation levels in and around the building. While this is not entirely practicable, this research analyses the shielding effectiveness of ten structurally significant US housing-stock models (walls and roofs) important for shielding against ionising radiation. The experimental data are used to benchmark computational models to calculate the shielding effectiveness of various building configurations under investigation from two types of realistic environmental source terms. Various combinations of these ten shielding models can be used to develop full-scale computational housing-unit models for building shielding factor calculations representing 69.6 million housing units (61.3%) in the United States. Results produced in this investigation provide a comparison between theory and experiment behind building shielding factor methodology. (paper)

  18. Cosmic gamma-ray background radiation. Current understandings and problems

    International Nuclear Information System (INIS)

    Inoue, Yoshiyuki

    2015-01-01

    The cosmic gamma-ray background radiation is one of the most fundamental observables in the gamma-ray band. Although the origin of the cosmic gamma-ray background radiation has been a mystery for a long time, the Fermi gamma-ray space telescope has recently measured it at 0.1-820 GeV and revealed that the cosmic GeV gamma-ray background is composed of blazars, radio galaxies, and star-forming galaxies. However, Fermi still leaves the following questions. Those are dark matter contribution, origins of the cosmic MeV gamma-ray background, and the connection to the IceCube TeV-PeV neutrino events. In this proceeding, I will review the current understandings of the cosmic gamma-ray background and discuss future prospects of cosmic gamma-ray background radiation studies. (author)

  19. 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.)

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

    DEFF Research Database (Denmark)

    Singh, K; Singh, H; Sharma, G

    2005-01-01

    , 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.......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...