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)

[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.

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}.

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)

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)

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)

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.)

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.

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.

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.

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

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

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)

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 .

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

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...

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

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.

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)

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.)

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.

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)

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)

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

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.

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.

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)

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.

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)

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

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.

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.)

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.

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)

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.

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

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.

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

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)

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

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)

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.

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.

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.

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.

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

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.

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.)

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)

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

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.

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.

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.)

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.)

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

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.)

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)

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 %

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

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)

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.

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.

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

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.)

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.

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.

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.

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)

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

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

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.)

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

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.

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.

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.

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''

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)

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

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)

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)

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

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)

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)

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

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

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

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.)

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.)

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)

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)

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)

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

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)

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

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.

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.

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

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.

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.

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

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.

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.

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

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.

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).

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.

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.)

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.

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.

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.

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)

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.

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.

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)

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.

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

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

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 ...

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

Hot Cell Window Shielding Analysis Using MCNP

International Nuclear Information System (INIS)

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

2009-01-01

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

Scintillation counter, maximum gamma aspect

International Nuclear Information System (INIS)

Thumim, A.D.

1975-01-01

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

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

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)

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.

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

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.

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

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.

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

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.

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.

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

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.)

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.

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)

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.

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.)

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)

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

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...

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)

The use of portable shields in industrial radiography

International Nuclear Information System (INIS)

Oliveira e Silva, J.A. de.

1988-01-01

This paper shows techniques actually used to reduce radiations exposure taxes during examinations execution by gamma radiography in regions of high population density. A portable equipment of radiation shield for using in exams by gamma radiography in pipelines, that is an adjustable device in an object body that will be examined, joining a measured collimator and a shield geometrically arranged so that the radiation restrict to impress the radiographic film used in examination without reaching people, objects or self-movings injurious that are nearness. (C.M.) [pt

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.

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

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)

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.

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.

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.

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

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

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)

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

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

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

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.

ANS shielding standards for light-water reactors

International Nuclear Information System (INIS)

Trubey, D.K.

1982-01-01

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

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.)

Designing shields for KeV photons with genetic algorithms

International Nuclear Information System (INIS)

Asbury, Stephen; Holloway, James P.

2011-01-01

Shielding of x-ray sources and low energy gamma rays is often accomplished with lead aprons, comprising a thin layer (0.5 mm to 1 mm) of lead or similar high-Z material. In previous work the authors used Genetic Algorithms to explore the design of a shadow shield for space applications. Now those techniques have been applied to the problem of shielding humans from low energy gamma radiation. This paper uses a simple geometry to explore layering various materials as a method to reduce mass and dose for thin gamma shields. The genetic algorithms discover layers of materials with various Z is in fact more effective than an equivalent mass of Pb alone for lower energy gammas, but as the incident radiation energy increases the efficacy of such layering diminishes. The utility of varying Z for lower energy gammas is in part due to their complementary K-edges, where one material compensates for the transmission that would occur just below the K-edge in another material. (author)

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)

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)

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.

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.

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)

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)

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

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.).

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

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

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

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)

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

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)

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.

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)

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

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

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.

Several problems in accelerator shielding study

International Nuclear Information System (INIS)

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

1980-01-01

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

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.)

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...

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

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...

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.

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)

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)

Shielding proposal to reduce cross-talk from ITER lower port to equatorial port

Energy Technology Data Exchange (ETDEWEB)

Juarez, Rafael, E-mail: rjuarez@ind.uned.es [Departamento de Ingeniería Energética, ETSII-UNED, Calle Juan del Rosal 12, Madrid 28040 (Spain); Pampin, Raul [F4E, Torres Diagonal Litoral B3, Josep Pla 2, Barcelona 08019 (Spain); Levesy, Bruno [ITER Organization, 13115 Route de Vinon sur Verdon, St Paul Lez Durance (France); Moro, Fabio [ENEA, Via Enrico Fermi 45, Frascati, Rome (Italy); Suarez, Alejandro [ITER Organization, 13115 Route de Vinon sur Verdon, St Paul Lez Durance (France); Catalan, J.P.; Sanz, Javier [Departamento de Ingeniería Energética, ETSII-UNED, Calle Juan del Rosal 12, Madrid 28040 (Spain)

2015-12-15

Radiation cross-talk from Torus Cryopump LP to EP was found to be a phenomenon driving Shutdown Dose Rates at EP Port Interspace after 12 days of cooling time, as relevant as neutron permeation through EP itself. In this work three different shields are proposed to mitigate the radiation cross-talk: two neutron shields placed inside LP, and a temporary gamma shield placed at EP PI during maintenance activities. Contributions from different reactor regions to Shutdown Dose Rates are computed, for the unshielded design, as long as the different shielded cases. The Rigorous-Two-Steps (R2S) method was used. The neutron shields inside TCP-LP are found to reduce SDR at EP PI 43 μSv/h and 99 μSv/h, while the gamma shield inside EP PI offers a reduction of 157 μSv/h in its heaviest configuration. Among these relevant reductions, the gamma shield inside the EP PI offers the best shielding option, as it reduces gamma cross-talk from TCP-LP and also protects EP PI from Port Duct and EP bellows activation, while it does not interfere with TCP performance.

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.

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.

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)

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.)

Alternative methodology for irradiation reactor experimental shielding calculation

International Nuclear Information System (INIS)

Vellozo, Sergio de Oliveira; Vital, Helio de Carvalho

1996-01-01

Due to a change in the project of the Experimental Irradiation Reactor, its shielding design had to be recalculated according to an alternative simplified analytical approach, since the standard transport calculations were temporarily unavailable. In the calculation of the new width for the shielding made up of steel and high-density concrete layers, the following radiation components were considered: fast neutrons and primary gammas (produced by fission and beta decay), from the core; and secondary gammas, produced by thermal neutron capture in the shielding. (author)

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

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.)

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...

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)

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...

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...

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)

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.)

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.

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.

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

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.

Optimal selection for shielding materials by fuzzy linear programming

International Nuclear Information System (INIS)

Kanai, Y.; Miura, N.; Sugasawa, S.

1996-01-01

An application of fuzzy linear programming methods to optimization of a radiation shield is presented. The main purpose of the present study is the choice of materials and the search of the ratio of mixture-component as the first stage of the methodology on optimum shielding design according to individual requirements of nuclear reactor, reprocessing facility, shipping cask installing spent fuel, ect. The characteristic values for the shield optimization may be considered their cost, spatial space, weight and some shielding qualities such as activation rate and total dose rate for neutron and gamma ray (includes secondary gamma ray). This new approach can reduce huge combination calculations for conventional two-valued logic approaches to representative single shielding calculation by group-wised optimization parameters determined in advance. Using the fuzzy linear programming method, possibilities for reducing radiation effects attainable in optimal compositions hydrated, lead- and boron-contained materials are investigated

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.

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

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

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)

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.

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.)

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

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

The status of shielding research at Tajoura research center

International Nuclear Information System (INIS)

El-Bakkoush, F.A.

2005-01-01

This paper gives a description to the shielding research activities which have been carried-out at the radiation shielding group ,Tajoura Research Center. This includes the design of different types of concrete shields made from local aggregates which have suitable radiation attenuation properties. These include, Ordinary Concrete(with density p = 2.3 ton/m3) heavy weight concrete (with density p =3.6 ton/m3) and heat resistant concrete with aggregates having bound- in water. Investigation have been carried -out by measuring the neutron and gamma-rays spectra which have been transmitted through barriers having different thickness. These were performed using a collimated beam of reactor neutrons and gamma-ray transmitted from the horizontal channel no 1 of Tajoura-Research reactor with 10 MW Max ape rating power. The transmitted fast neutron and gamma spectra were measured by neutron-gamma spectrometer employing NE-213 liquid organic scintillater. Discrimination of against undesired pulses of neutrons or gamma-ray was achieved by a pulse shape discrimination method based on differences in the shape of the decay part of the emitted pulses. The obtained results are presented in the form of displayed neutron and gamma spectra measured behind different thickness of the investigated concrete shield. These spectra were used to derive the macroscopic cross section for at different energy for material under investigation

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...

Methodology of shielding calculation for nuclear reactors

International Nuclear Information System (INIS)

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

1982-01-01

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

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)

Calculation and mapping of gamma radiation field in the pool of Apsara reactor

International Nuclear Information System (INIS)

Singh, Tej; Singh, Kanchhi; Sharma, ARchana; Somakumar, K.; Raina, V.K.; Srinivasan, P.; Prasad, S.K.; Babu, D.A.R.; Sharma, D.N.

2007-12-01

Theoretical simulation of the radiation transport occurring in the Apsara core and bulk shield was carried out using two different radiation transport codes, MCNP and QADCG. The MCNP is a Monte Carlo based statistical method solving Boltzmann transport equation, where as the latter code QADCG is a point kernel based deterministic method with build-up factor correction. The aim of the simulation was to do a dose mapping and estimate the expected value of gamma dose rates at various locations where experimental measurements were conducted. Details regarding the simulation techniques employed by both the MCNP and QADCG software with reference to the Apsara core and shield geometry and source gamma energy distribution in the fuel plates are presented in this report. Different types of particle tallies requested in MCNP and QADCG are discussed. Details of variance reduction methods employed in reducing the statistical uncertainty of Monte Carlo simulation are also mentioned in the report. The statistical errors associated with Monte Carlo based simulation varied between 3% - 6% in most of the energy bins that contribute to the total fluence and hence to the dose rates. It was observed that the experimental values and the theoretically simulated values match each other closely following a similar trend except for certain experimental locations which had photon flux contributions from extraneous sources like the N-16 activity present in water, beam tubes and pool liner towards shielding corner. It is seen that the theoretical values are found to be larger than experimental values by factors ranging from 1.1 to 3 depending on the water shield thickness. This study served in validation of the experimental measurements conducted by GM counter based teletector and dipole based detectors. In addition, the comparison provided a confirmation of the accuracy of the radiation transport simulation techniques used for dose rate evaluation in case of complex source geometries and

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

Energy Technology Data Exchange (ETDEWEB)

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

1993-04-15

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

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...

Numerical simulation of a reinforced concrete shield around a nuclear reactor

International Nuclear Information System (INIS)

Mahama, Mumuni Salifu

1996-02-01

Ghana currently operates a Research Reactor and other nuclear facilities including a Gamma Irradiation Facility, a Radiographic Non-Destructive Testing laboratory and would be operating in the nearest future a Radiotherapy Centre. Each of these has a concrete radiation shield as a major safety device. In carrying out its functions, a concrete radiation shield may be subjected to thermal and mechanical stresses. A facility for analysing these stresses is desirable. Two computer codes have been developed under this programme for radiation shielding computation and stress analysis of cylindrical reactor shields. (au)

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

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

Shielding study of a fusion machine. Elaboration of a global shielding calculation scheme for the Tokamak tore Supra

International Nuclear Information System (INIS)

Diop, C.M'B.

1984-01-01

This thesis presents a global shielding calculation scheme for neutron and gamma rays arising from the Tokamak TORE SUPRA fusion device, in which a deuterium plasma is used. To study the shield parameters we have elabored a important chaining of neutron and gamma transport codes, TRIPOLI, ANISN, MERCURE 4, allowing to evaluate the radial and skyshine components of the dose rate behind the concrete shield. The study of thermonuclear neutron activation is fundamental to define a tokamak exploitation strategy. For this, two formalisme have been developed. They are based on a modelization of the activation reaction rates according to TRIPOLI, ANISN, and MERCURE 4 codes capabilities. The first one calculates, in one dimensional geometry, the desactivation gamma dose rate inside the vacuum chamber. The second one is a tridimensional model which determines the spatial variation of the gamma dose rate in the machine room. The problem of the existence of runaway electrons and associated secondaries radiations, bremsstrahlung gamma rays particularly, is approched. The results which are presented have contributed to define the parameters of the concrete shield and a strategy for TORE SUPRA Tokamak exploitation [fr

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)

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)

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

Detailed mechanical design of the LIPAc beam dump radiological shielding

Energy Technology Data Exchange (ETDEWEB)

Nomen, Oriol, E-mail: onomen@irec.cat [IREC, Barcelona, Catalonia (Spain); CDEI-UPC, Barcelona, Catalonia (Spain); Martínez, José I.; Arranz, Fernando; Iglesias, Daniel; Barrera, Germán; Brañas, Beatriz [CIEMAT, Madrid (Spain); Ogando, Francisco [UNED, Madrid (Spain); Molla, Joaquín [CIEMAT, Madrid (Spain); Sanmartí, Manel [IREC, Barcelona, Catalonia (Spain)

2013-10-15

Highlights: ► Mechanical design of the IFMIF LIPAc beam dump shielding has been performed. ► Lead shutter design performed to shield radiation from beam dump when LIPAc is off. ► External loads, working and dismantling conditions, included as design constraints. -- Abstract: The LIPAc is a 9 MeV, D{sup +} linear prototype accelerator for the validation of the IFMIF accelerator design. The high intensity, 125 mA CW beam is stopped in a copper cone involving a high production of neutrons and gamma radiation and activation of its surface. The beam stopper is surrounded by a shielding to attenuate the resulting radiation so that dose rate values comply with the limits at the different zones of the installation. The shielding includes for that purpose polyethylene rings, water tanks and gray cast iron rings. A lead shutter has also been designed to shield the gamma radiation that comes through the beam tube when the linear accelerator is not in operation, in order to allow access inside the building for maintenance tasks. The present work summarizes the detailed mechanical design of the beam dump shielding and the lead shutter taking into account the design constraints, such as working conditions and other external loads, as well as including provisions for dismantling.

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

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.

Shielding structure analysis for LSDS facility

Energy Technology Data Exchange (ETDEWEB)

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

2014-05-15

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

Shielding structure analysis for LSDS facility

International Nuclear Information System (INIS)

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

2014-01-01

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

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

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)

Shielding for a tandem accelerator coupled to linac booster

International Nuclear Information System (INIS)

Bhattacharyya, S.; Bisht, J.S.; Venkataraman, G.

1996-01-01

Shielding calculation for the Beam-Hall-II of pelletron facility, augmented with linac booster in its phase-II at Nuclear Science Centre, New Delhi, has been done. An estimate is obtained by reduction factor method considering source radiation of monoenergetic neutrons, which is then compared with the detail computation using computer code ALICE considering total energy and angular distribution of neutrons. Another code ASFIT is used to take into account the build up of gamma dose from (n, gamma) reactions within the concrete shield incorporating new radiation weighting factors as recommended by ICRP-60. (author). 8 refs., 2 figs

Radiation anomaly detection algorithms for field-acquired gamma energy spectra

Science.gov (United States)

Mukhopadhyay, Sanjoy; Maurer, Richard; Wolff, Ron; Guss, Paul; Mitchell, Stephen

2015-08-01

The Remote Sensing Laboratory (RSL) is developing a tactical, networked radiation detection system that will be agile, reconfigurable, and capable of rapid threat assessment with high degree of fidelity and certainty. Our design is driven by the needs of users such as law enforcement personnel who must make decisions by evaluating threat signatures in urban settings. The most efficient tool available to identify the nature of the threat object is real-time gamma spectroscopic analysis, as it is fast and has a very low probability of producing false positive alarm conditions. Urban radiological searches are inherently challenged by the rapid and large spatial variation of background gamma radiation, the presence of benign radioactive materials in terms of the normally occurring radioactive materials (NORM), and shielded and/or masked threat sources. Multiple spectral anomaly detection algorithms have been developed by national laboratories and commercial vendors. For example, the Gamma Detector Response and Analysis Software (GADRAS) a one-dimensional deterministic radiation transport software capable of calculating gamma ray spectra using physics-based detector response functions was developed at Sandia National Laboratories. The nuisance-rejection spectral comparison ratio anomaly detection algorithm (or NSCRAD), developed at Pacific Northwest National Laboratory, uses spectral comparison ratios to detect deviation from benign medical and NORM radiation source and can work in spite of strong presence of NORM and or medical sources. RSL has developed its own wavelet-based gamma energy spectral anomaly detection algorithm called WAVRAD. Test results and relative merits of these different algorithms will be discussed and demonstrated.

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)

Shielding of medically used proton accelerators; Abschirmung von medizinisch genutzten Protonenbeschleunigern

Energy Technology Data Exchange (ETDEWEB)

Ewen, Klaus

2014-10-01

In several standards of the standards committee radiology (NRA) the shielding of proton accelerators (cyclotrons) for medical utilization is described. Proton beams can be used in nuclear medicine for PET (proton emission tomography) isotope production or for radiotherapeutic use. The dominating radiation from proton induced nuclear reactions is fast neutron radiation. The calculation procedure for appropriate shielding measures according to the NAR standards is described step-by-step. AN adequate shielding of fast neutrons is also sufficient for the generated gamma radiation.

RADSHI: shielding calculation program for different geometries sources

International Nuclear Information System (INIS)

Gelen, A.; Alvarez, I.; Lopez, H.; Manso, M.

1996-01-01

A computer code written in pascal language for IBM/Pc is described. The program calculates the optimum thickness of slab shield for different geometries sources. The Point Kernel Method is employed, which enables the obtention of the ionizing radiation flux density. The calculation takes into account the possibility of self-absorption in the source. The air kerma rate for gamma radiation is determined, and with the concept of attenuation length through the equivalent attenuation length the shield is obtained. The scattering and the exponential attenuation inside the shield material is considered in the program. The shield materials can be: concrete, water, iron or lead. It also calculates the shield for point isotropic neutron source, using as shield materials paraffin, concrete or water. (authors). 13 refs

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.

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.

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

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

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

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

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

International Nuclear Information System (INIS)

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

1996-01-01

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

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

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)

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

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.)

Estimation of Shielding Thickness for a Prototype Department of Energy National Spent Nuclear Fuel Program Transport Cask

Energy Technology Data Exchange (ETDEWEB)

SANCHEZ,LAWRENCE C.; MCCONNELL,PAUL E.

2000-07-01

Preliminary shielding calculations were performed for a prototype National Spent Nuclear Fuel Program (NSNFP) transport cask. This analysis is intended for use in the selection of cask shield material type and preliminary estimate of shielding thickness. The radiation source term was modeled as cobalt-60 with radiation exposure strength of 100,000 R/hr. Cobalt-60 was chosen as a surrogate source because it simultaneous emits two high-energy gammas, 1.17 MeV and 1.33 MeV. This gamma spectrum is considered to be large enough that it will upper bound the spectra of all the various spent nuclear fuels types currently expected to be shipped within the prototype cask. Point-kernel shielding calculations were performed for a wide range of shielding thickness of lead and depleted uranium material. The computational results were compared to three shielding limits: 200 mrem/hr dose rate limit at the cask surface, 50 mR/hr exposure rate limit at one meter from the cask surface, and 10 mrem/hr limit dose rate at two meters from the cask surface. The results obtained in this study indicated that a shielding thickness of 13 cm is required for depleted uranium and 21 cm for lead in order to satisfy all three shielding requirements without taking credit for stainless steel liners. The system analysis also indicated that required shielding thicknesses are strongly dependent upon the gamma energy spectrum from the radiation source term. This later finding means that shielding material thickness, and hence cask weight, can be significantly reduced if the radiation source term can be shown to have a softer, lower energy, gamma energy spectrum than that due to cobalt-60.

Performance of neutron and gamma personnel dosimetry in mixed radiation fields

International Nuclear Information System (INIS)

Swaja, R.E.; Sims, C.S.

1981-01-01

From 1974 to 1980, six personnel dosimetry intercomparison studies (PDIS) were conducted at the Oak Ridge National Laboratory (ORNL) to evaluate the performance of personnel dosimeters in a variety of neutron and gamma fields produced by operating the Health Physics Research Reactor (HPRR) in the steady state mode with and without spectral modifying shields. A total of 58 different organizations participated in these studies which produced approximately 2000 measurements of neutron and gamma dose equivalents on anthropomorphic phantoms for five different reactor spectra. Based on these data, the relative performance of three basic types of neutron dosimeters [nuclear emulsion film, thermoluminescent (TLD), and track-etch] and two basic types of gamma dosimeters (film and TLD) in mixed radiation fields was assessed

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

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...

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

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

The influence of Shelter's FCM on the shield efficiency at there of containing

International Nuclear Information System (INIS)

Gorbachev, B.I.

2000-01-01

The reasonable detailed quantitative estimations of the influence of the γ-radiation capture and scattering processes in the Shelter's FCM material on the shield precautions efficiency at there of containing for the further shelf purpose. The Monte-Carlo calculations was carry out by the software Micro Shield 4.00 serial 4.00-00283, which make it possible correctly to account for the radiation shielding geometry, the radiation sources geometry, the radiation sources spectrums and the processes of the gamma-rays multi scattering in 'thick' shielding. Results presented in the tables, which is convenient to use. 3 refs., 18 tab

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.

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)

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.

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.

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.

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

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

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)

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.

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.

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

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.

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.

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)

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

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

International Nuclear Information System (INIS)

Sato, Satoshi

2003-09-01

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

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

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

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.)

Concrete shielding for nuclear ship 'Mutsu'

International Nuclear Information System (INIS)

Nagase, Tetsuo; Saito, Tetsuo

1983-01-01

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

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

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

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)

Parameters calculation of shielding experiment

International Nuclear Information System (INIS)

Gavazza, S.

1986-02-01

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

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

International Nuclear Information System (INIS)

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

2013-01-01

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

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)

Monte Carlo based radial shield design of typical PWR reactor

Energy Technology Data Exchange (ETDEWEB)

Gul, Anas; Khan, Rustam; Qureshi, M. Ayub; Azeem, Muhammad Waqar; Raza, S.A. [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan). Dept. of Nuclear Engineering; Stummer, Thomas [Technische Univ. Wien (Austria). Atominst.

2016-11-15

Neutron and gamma flux and dose equivalent rate distribution are analysed in radial and shields of a typical PWR type reactor based on the Monte Carlo radiation transport computer code MCNP5. The ENDF/B-VI continuous energy cross-section library has been employed for the criticality and shielding analysis. The computed results are in good agreement with the reference results (maximum difference is less than 56 %). It implies that MCNP5 a good tool for accurate prediction of neutron and gamma flux and dose rates in radial shield around the core of PWR type reactors.

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

International Nuclear Information System (INIS)

Katzung, W.; Bottcher, J.

2009-01-01

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

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

Evaluation of dose attenuation factor of armored car against radiation accidents

International Nuclear Information System (INIS)

Sato, Tatsuhiko; Fujii, Katsutoshi; Murayama, Takashi

2002-03-01

The Tokyo Fire Department developed an armored car against radiation accidents. The car is covered by lead shields for attenuating dose from gamma rays. Dose from neutrons also can be attenuated by pouring water into tanks attached to the surface of the car. However, dose attenuation factors of the radiation shields had been determined by an estimation of single-layer shield, and more precise evaluation of multi-layer shield was required. By request from the Tokyo Fire Department, a precise evaluation of the dose attenuation in multi-layer shield was carried out. The evaluation was made by a Monte Carlo radiation transport simulation code MCNP4B for the shields used in the front, side and back of the car. Three types of the radiation sources ( 252 Cf as a neutron source, 60 Co as a gamma ray source, and radiation source corresponding to the JCO criticality accident) were considered in the calculation. Benchmark experiments using neutron and gamma ray sources were also performed for ensuring the evaluation method. As a result, it was found out that doses of neutron and gamma ray were attenuated to approximately 10% and 25% by the thickest shield, respectively. These values were close to the ones which had already obtained by the estimation of single-layer shield. (author)

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...

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

Attenuation of Neutron and Gamma Radiation by a Composite Material Based on Modified Titanium Hydride with a Varied Boron Content

Science.gov (United States)

Yastrebinskii, R. N.

2018-04-01

The investigations on estimating the attenuation of capture gamma radiation by a composite neutron-shielding material based on modified titanium hydride and Portland cement with a varied amount of boron carbide are performed. The results of calculations demonstrate that an introduction of boron into this material enables significantly decreasing the thermal neutron flux density and hence the levels of capture gamma radiation. In particular, after introducing 1- 5 wt.% boron carbide into the material, the thermal neutron flux density on a 10 cm-thick layer is reduced by 11 to 176 factors, and the capture gamma dose rate - from 4 to 9 times, respectively. The difference in the degree of reduction in these functionals is attributed to the presence of capture gamma radiation in the epithermal region of the neutron spectrum.

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.

Improvement of top shield analysis technology for CANDU 6 reactor

International Nuclear Information System (INIS)

Kim, Kyo Yoon; Jin, Young Kwon; Lee, Sung Hee; Moon, Bok Ja; Kim, Yong Il

1996-07-01

As for Wolsung NPP unit 1, radiation shielding analysis was performed by using neutron diffusion codes, one-dimensional discrete ordinates code ANISN, and analytical methods. But for Wolsung NPP unit 2, 3, and 4, two-dimensional discrete ordinates code DOT substituted for neutron diffusion codes. In other words, the method of analysis and computer codes used for radiation shielding of CANDU 6 type reactor have been improved. Recently Monte Carlo MCNP code has been widely utilized in the field of radiation physics and other radiation related areas because it can describe an object sophisticately by use of three-dimensional modelling and can adopt continuous energy cross-section library. Nowadays Monte Carlo method has been reported to be competitive to discrete ordinate method in the field of radiation shielding and the former has been known to be superior to the latter for complex geometry problem. However, Monte Carlo method had not been used for radiation streaming calculation in the shielding design of CANDU type reactor. Neutron and gamma radiations are expected to be streamed from calandria through the penetrations to reactivity mechanism deck (R/M deck) because many reactivity control units which are established on R/M deck extend from R/M deck to calandria within penetrations, which are provided by guide tube extensions. More precise estimation of radiation streaming is required because R/M deck is classified as an accessible area where atomic worker can access when necessary. Therefore neutron and gamma dose rates were estimated using MCNP code on the R/M deck in the top shield system of CANDU 6 reactor. 9 tabs., 17 figs., 21 refs. (Author)

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

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)

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.

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)

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 ).

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.)

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.)

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.

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.

Study characteristics of new concrete mixes and their mechanical, physical, and gamma radiation attenuation features

Energy Technology Data Exchange (ETDEWEB)

El-Samrah, Moamen G.; Abdel-Rahman, Mohamed A.E. [Nuclear Engineering Department, Military Technical College Kobry El-kobbah, Cairo (Egypt); Kany, Amr M.I. [Physics Department, Faculty of Science, Al-Azhar University, Cairo (Egypt)

2018-02-01

Ordinary concrete and those of different compositions are regarded as suitable material in many applications concerning with gamma and neutron radiation shielding purposes. They are widely used in nuclear power plant, medical facilities, nuclear shelters, and for radioactive materials transportation as well as storage of radioactive wastes. In this study four different concrete mixes were prepared with the following different types of coarse aggregates: dolomite, barite, goethite, and steel slag. The effect of changes in the fine aggregates, selected to be 50 % local sand and 50 % limonite with addition of 10 % silica fume (SF) and 10 % fly ash (FA) by replacement of the total cement weight, on the performance of the samples was also investigated. To examine the performance of such samples for radiation shielding applications, a set of physical, mechanical, and radiation attenuation properties was studied and compared with those of ordinary concrete. This investigation includes compressive strength, slump test, bulk density, ultrasonic pulse velocity test, and gamma rays attenuation measurements for the different samples. A verification of the experimental results concerning the radiation attenuation measurements was performed using WinXcom program (Version 3.1). The experimental results revealed that all concrete mixes; goethite-limonite concrete (G.L), barite-limonite concrete (B.L), steel slag-limonite concrete (S.L) and dolomite concrete (D.C) have good physical and mechanical properties that successfully satisfying them as high performance concretes. In addition the barite-limonite and the steel slag-limonite have the higher γ-ray attenuation coefficients at low and high energy range and hence have a better radiation shielding. The obtained results from WinXcom program calculations showed a good agreement with the experimental results concerning γ-ray attenuation measurements for the studied concrete mixes. (copyright 2018 WILEY-VCH Verlag GmbH and Co. KGa

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

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.

Parameters calculation of a shielding experiment and evaluation of calculation methodology

International Nuclear Information System (INIS)

Gavazza, S.; Otto, A.C.; Gomes, I.C.; Maiorino, J.R.

1986-01-01

In this text is carried out the evaluation of radiation transport methodology, comparying the calculated reactions and dose rates, for neutrons and gamma-rays, with the experimental measurements obtained on iron shield, irradiated in YAYOI reactor. Were employed the ENDF/B-IV and VITAMIN-C libraries and the AMPX-II modular system for generation of cross sections, collapsed by the ANISN code. The transport calculation were made by using the DOT 3.5 code, adjusting the spectrum of the iron shield boundary source to the reactions and dose rates, measured at the beginning of shield. The distributions calculated for neutrons and gamma-rays, on iron shield, presented coherence with the experimental measurements. (Author) [pt

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

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

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...

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.)

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...

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

International Nuclear Information System (INIS)

Tian Yingnan; Chen Yixue; Yang Shouhai

2013-01-01

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

Special concrete shield selection using the analytic hierarchy process

International Nuclear Information System (INIS)

Abulfaraj, W.H.

1994-01-01

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

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.

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)

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

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

International Nuclear Information System (INIS)

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

2000-01-01

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

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.

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.

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.

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

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

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

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)

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.

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.

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

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

Radiological shielding of low power compact reactor: calculation and design

International Nuclear Information System (INIS)

Marino, Raul

2004-01-01

The development of compact reactors becoming a technology that offers great projection and innumerable use possibilities, both in electricity generation and in propulsion.One of the requirements for the operation of this type of reactor is that it must include a radiological shield that will allow for different types of configurations and that, may be moved with the reactor if it needs to be transported.The nucleus of a reactor emits radiation, mainly neutrons and gamma rays in the heat of power, and gamma radiation during the radioactive decay of fission products.This radiation must be restrained in both conditions of operation to avoid it affecting workers or the public.The combination of different materials and properties in layers results in better performance in the form of a decrease in radiation, hence causing the dosage outside the reactor, whether in operation or shut down, to fall within the allowed limits.The calculations and design of radiological shields is therefore of paramount importance in reactor design.The choice of material and the design of the shield have a strong impact on the cost and the load capacity, the latter being one of the characteristics to optimize.The imposed condition of design is that the reactor can be transported together with the decay shield in a standard container of 40 foot [es

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

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

Structural and optical properties of Bi2O3-B2O3-CdO-Na2O glass system for gamma ray shielding applications

Science.gov (United States)

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

2018-05-01

Quaternary system of the composition (0.15+x) Bi2O3-(0.55-x) B2O3-0.15CdO-0.15Na2O (where x=0, 0.1, 0.3 and 0.5 mole fraction) has been synthesized using melt-quenching technique. Gamma ray shielding properties are measured in terms of mass attenuation coefficient and half value layer at photon energies 662, 1173 and 1332 keV. These parameters are compared with standard nuclear radiation shielding `barite and ferrite' concretes. The results reflect better radiation shielding properties as compared to barite and ferrite concretes. Effective atomic number is calculated at photon energies 662 and 1173 keV. Density, molar volume and XRD studies are analyzed to know physical and structural properties of the glass system. Optical band gap, refractive index and molar refraction are calculated from UV-Visible measurements. Decrease in optical band gap and increase in molar refraction have been observed indicating the increase of non-bridging oxygens in the structure.

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.

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)

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)

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.)

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...

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

International Nuclear Information System (INIS)

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

2017-01-01

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

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

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...

Gamma-ray irradiation of a boreal forest ecosystem

International Nuclear Information System (INIS)

Guthrie, J.E.; Dugle, J.R.

1983-01-01

A long-term radiation ecology research project called Field Irradiator - Gamma (FIG) began at the Whiteshell Nuclear Research Establishment in 1968. The experimental area is in southeastern Manitoba and is located on the western edge of the Precambrian shield. The project studies the ecological effects continuous exposure to a gradient of gamma radiation has on a mixed boreal forest ecosystem. The gradient ranges from 1 to 460,000 times the natural background radiation level. This paper describes the forest, the gamma irradiator and its radiation field, and the research program

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)

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.

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)

Penetration shielding applications of CYLSEC

International Nuclear Information System (INIS)

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

1985-01-01

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

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.

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.)

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

Science.gov (United States)

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

2011-01-01

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

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

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

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.)

Monteray Mark-I: Computer program (PC-version) for shielding calculation with Monte Carlo method

International Nuclear Information System (INIS)

Pudjijanto, M.S.; Akhmad, Y.R.

1998-01-01

A computer program for gamma ray shielding calculation using Monte Carlo method has been developed. The program is written in WATFOR77 language. The MONTERAY MARH-1 is originally developed by James Wood. The program was modified by the authors that the modified version is easily executed. Applying Monte Carlo method the program observe photon gamma transport in an infinity planar shielding with various thick. A photon gamma is observed till escape from the shielding or when its energy less than the cut off energy. Pair production process is treated as pure absorption process that annihilation photons generated in the process are neglected in the calculation. The out put data calculated by the program are total albedo, build-up factor, and photon spectra. The calculation result for build-up factor of a slab lead and water media with 6 MeV parallel beam gamma source shows that they are in agreement with published data. Hence the program is adequate as a shielding design tool for observing gamma radiation transport in various media

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)

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)

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.

[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