Elastomeric neutron shielding material and process of production

International Nuclear Information System (INIS)

John, G.; Knorr, W.

1987-01-01

Elastomeric neutron shielding material made of plastic with high hydrogen content, characterized in that the shielding material is a polymeric reaction product of a reaction between (a) polyol on the base of polybutadiene which compares with polyethylene with regard to hydrogen content, and (b) aliphatic diisocyanate, and in that the hydrogen content is higher than 8 weight per cent. (orig.) [de

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)

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

International Nuclear Information System (INIS)

Dhang, Prosenjit; Verma, Rishi; Shyam, Anurag

2015-01-01

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

Shielding experiments in different materials with 252Cf neutron spectra

International Nuclear Information System (INIS)

Sathian, Deepa; Marathe, P.K.; Pal, Rupali; Jayalakshmi, V.; Chourasiya, G.; Mayya, Y.S.

2008-01-01

Adequate shielding for neutron sources can be determined using analytical method or by actually measuring the attenuation for the target configuration. This paper describes the measurement of Half Value Thickness (HVT), Tenth Value Thickness (TVT), Σ values for four different shielding materials, using a standard 252 Cf neutron source and comparing with the values calculated using an empirical relationship. BF 3 based REM-counter is used for measurement of neutron dose equivalent, against different thickness of the shielding material. The experimental HVT and S values are in good agreement with the calculated values. From this study, it is concluded that, among the four materials studied, high density polyethylene (HDPE) is best suitable for the shielding of a 252 Cf neutron source. (author)

Cosmic Ray Interactions in Shielding Materials

International Nuclear Information System (INIS)

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

2011-01-01

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

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.

Development of neutron shielding concrete containing iron content materials

Science.gov (United States)

Sariyer, Demet; Küçer, Rahmi

2018-02-01

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

Development of neutron shielding material using metathesis-polymer matrix

Energy Technology Data Exchange (ETDEWEB)

Sakurai, Yoshinori E-mail: ysakurai@rri.kyoto-u.ac.jp; Sasaki, Akira; Kobayashi, Tooru

2004-04-21

A neutron shielding material using a metathesis-polymer matrix, which is a thermosetting resin, was developed. This shielding material has characteristics that can be controlled for different mixing ratios of neutron absorbers and for formation in the laboratory. Additionally, the elastic modulus can be changed at the hardening process, from a flexible elastoma to a mechanically tough solid. Experiments were performed at the Kyoto University Research Reactor in order to determine the important characteristics of this metathesis-polymer shielding material, such as neutron shielding performance, secondary gamma-ray generation and activation. The metathesis-polymer shielding material was shown to be practical and as effective as the other available shielding materials, which mainly consist of thermoplastic resin.

Thermal testing of solid neutron shielding materials

International Nuclear Information System (INIS)

Boonstra, R.H.

1990-03-01

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

Thermal testing of solid neutron shielding materials

International Nuclear Information System (INIS)

Boonstra, R.N.

1990-01-01

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

Using glass as a shielding material

International Nuclear Information System (INIS)

Yousef, S.

2002-04-01

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

Using glass as a shielding material

International Nuclear Information System (INIS)

Yousef, S.

2003-01-01

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

Development of highly effective neutron shielding material made of phenol-novolac type epoxy resin

Energy Technology Data Exchange (ETDEWEB)

Cho, Soo Haeng; Jeong, Myeong Soo; Hong, Sun Seok; Lee, Won Kyoung; Kim, Ik Soo; Shin, Young Joon; Do, Jae Bum; Ro, Seung Gy; Oh, Seok Jin

1998-06-01

Because the exposure to radiation in the nuclear facilities can be fatal to human, it is important to reduce the radiation dose level to a tolerable level. The purpose of this study is to develop highly effective neutron shielding materials for the shipping and storage cask of radioactive materials or in the nuclear/radiation facilities. On this study, we developed epoxy resin based neutron shielding materials and their various material properties, including neutron shielding ability, fire resistance, combustion characteristics, radiation resistance, thermal and mechanical properties were evaluated experimentally. Especially we developed phenol-novolac type epoxy resin based neutron shielding materials and their characteristics were also evaluated. (author). 22 refs., 11 tabs., 21 figs

Development of Neutron Shielding Material for Cask and Accelerator

International Nuclear Information System (INIS)

Kang, Hee Young; Seo, Ki Seog; Lee, Byung Chul; Park, Chang Jae; Kim, Ho Dong

2008-01-01

The neutron shielding materials are used as a neutron shield for spent fuel shipping cask, beam accelerators and neutron generators. At early stage, the neutron attenuations of materials were evaluated with the cross sections. After that, benchmark or mock-up experiments on the multi-layer problem to confirm the shielding characteristics or to evaluate analysis accuracy were reported. Recently, the need to transport spent nuclear fuels is increasing due to the current limited storage capacity. The on-site storage capacity at some of nuclear power plants is expected to be full in near future. With a growing inventory of spent fuels at power plants, these spent fuels need to be transported to other storage facilities. Shipping casks have been developed to safely transport spent fuels that emit high neutrons and gamma-ray radiation. The external radiation level of the shipping cask from the spent fuel must be limited to meet the standards specified by the IAEA radioactive material package regulation, so it is important to develop a proper neutron shielding material for a shipping cask. Neutron shielding experiments and analyses on the shielding effects of materials have been conducted, and some experiments have been performed to examine the shielding effects of selected materials. The shielding experiments consist of evaluating not only the shielding effects of a material alone but also the effects of the material thickness. The experimental results were compared with those obtained by using the MCNP-5c code

Development and application of high performance liquid shielding materials

International Nuclear Information System (INIS)

Miura, Toshimasa; Omata, Sadao; Otano, Naoteru; Hirao, Yoshihiro; Kanai, Yasuji

1998-01-01

Development of liquid shielding material with good performance for neutron and γ-ray was investigated. Lead, hydrogen and boron were selected as the elements of shielding materials which were made by the ultraviolet curing method. Good performance shielding materials with about 1 mm width to neutron and gamma ray were produced by mixing lead, boron compound and ultraviolet curing monomer with many hydrogens. The shielding performance was the same as a concrete with two times width. The activation was very small such as 1/10 6 -1/10 8 of the standard concrete. The weight and the external appearance did not charged from room temperature to 100degC. Polyfunctional monomer had good thermal resistance. This shielding material was applied to double bending cylindrical duct and annulus ring duct. The results proved the shielding materials developed had good performance. (S.Y.)

Uranium-lead shielding for nuclear material transportation systems

International Nuclear Information System (INIS)

Lusk, E.C.; Miller, N.E.; Basham, S.J. Jr.

1978-01-01

The basis for the selection of shielding materials for spent fuel shipping containers is described with comments concerning the favorable and unfavorable aspects of steel, lead, and depleted uranium. A concept for a new type of material made of depleted uranium and lead is described which capitalizes on the best cask shielding characteristics of both materials. This cask shielding is made by filling the shielding cavity with pieces of depleted uranium and then backfilling the interstitial voids with lead. The lead would be bonded to the uranium and also to the cask shells if desired. Shielding density approaching 80 percent of that of solid uranium could be achieved, while a density of 65 percent is readily obtainable. This material should overcome the problems of the effect of lead melting in the fire accident, high thermal gradients at uranium-stainless steel interfaces and at a major reduction in cost over that of a solid uranium shielded cask. A development program is described to obtain information on the properties of the composite material to aid in design analysis and licensing and to define the fabrication techniques

Novel shielding materials for space and air travel

International Nuclear Information System (INIS)

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

2006-01-01

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

Ground-based simulations of cosmic ray heavy ion interactions in spacecraft and planetary habitat shielding materials

Science.gov (United States)

Miller, J.; Zeitlin, C.; Heilbronn, L.; Borak, T.; Carter, T.; Frankel, K. A.; Fukumura, A.; Murakami, T.; Rademacher, S. E.; Schimmerling, W.;

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.)

Dynamic Test Method Based on Strong Electromagnetic Pulse for Electromagnetic Shielding Materials with Field-Induced Insulator-Conductor Phase Transition

Science.gov (United States)

Wang, Yun; Zhao, Min; Wang, Qingguo

2018-01-01

In order to measure the pulse shielding performance of materials with the characteristic of field-induced insulator-conductor phase transition when materials are used for electromagnetic shielding, a dynamic test method was proposed based on a coaxial fixture. Experiment system was built by square pulse source, coaxial cable, coaxial fixture, attenuator, and oscilloscope and insulating components. S11 parameter of the test system was obtained, which suggested that the working frequency ranges from 300 KHz to 7.36 GHz. Insulating performance is good enough to avoid discharge between conductors when material samples is exposed in the strong electromagnetic pulse field up to 831 kV/m. This method is suitable for materials with annular shape, certain thickness and the characteristic of field-induced insulator-conductor phase transition to get their shielding performances of strong electromagnetic pulse.

Neutron shielding material

International Nuclear Information System (INIS)

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

1980-01-01

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

Method of measurement on materials shielding effectiveness test in time domain

International Nuclear Information System (INIS)

Liu Shunkun; Han Jun; Chen Xiangyue

2009-01-01

Windows method is a measurement of slot coupling effect in nature when it is used to measure material's shielding effectiveness. The error of measurement will become serious when it is used to measure material's shielding effectiveness in low frequency band. It is difficult to measure material's shielding effectiveness of electromagnetic pulse with Windows method. Device under test method (DUT method) was presented in this paper to overcome the limitations of Windows method in material's shielding effectiveness test. The method can be used to measure any material's shielding Effectiveness effectively through the design and the test of the DUT.The method was used to measure shielding effectiveness of special cement .Compared with theoretical analysis,the measurement result prove the DUT method to be very efficient in material's shielding effectiveness test. (authors)

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

Evaluation of Shielding Performance for Newly Developed Composite Materials

Science.gov (United States)

Evans, Beren Richard

This work details an investigation into the contributing factors behind the success of newly developed composite neutron shield materials. Monte Carlo simulation methods were utilized to assess the neutron shielding capabilities and secondary radiation production characteristics of aluminum boron carbide, tungsten boron carbide, bismuth borosilicate glass, and Metathene within various neutron energy spectra. Shielding performance and secondary radiation data suggested that tungsten boron carbide was the most effective composite material. An analysis of the macroscopic cross-section contributions from constituent materials and interaction mechanisms was then performed in an attempt to determine the reasons for tungsten boron carbide's success over the other investigated materials. This analysis determined that there was a positive correlation between a non-elastic interaction contribution towards a material's total cross-section and shielding performance within the thermal and epi-thermal energy regimes. This finding was assumed to be a result of the boron-10 absorption reaction. The analysis also determined that within the faster energy regions, materials featuring higher non-elastic interaction contributions were comparable to those exhibiting primarily elastic scattering via low Z elements. This allowed for the conclusion that composite shield success within higher energy neutron spectra does not necessitate the use elastic scattering via low Z elements. These findings suggest that the inclusion of materials featuring high thermal absorption properties is more critical to composite neutron shield performance than the presence of constituent materials more inclined to maximize elastic scattering energy loss.

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)

SPADA: a project to study the effectiveness of shielding materials in space

International Nuclear Information System (INIS)

Pugliese, M.; Casolino, M.; Cerciello, V.

2008-01-01

The SPADA (SPAce Dosimetry for Astronauts) project is a part of an extensive teamwork that aims to optimize shielding solutions against space radiation. Shielding is indeed all irreplaceable tool to reduce, exposure of crews of future Moon and Mars missions. We concentrated our studies on two flexible materials, Kevlar (R) and Nextel (R), because of their ability to protect space infrastructure from micro meteoroids measured radiation hardness of these shielding materials and compared to polyethylene, generally acknowledged as the most effective space radiation shield with practical applications in spacecraft. Both flight test (on the International Space Station and on the Russian FOTON M3 rocket), with passive dosimeters and accelerator-based experiments have been performed. Accelerator tests using high-energy Fe ions have demonstrated that Kevlar is almost as effective as polyethylene in shielding heavy ions, while Nextel is a poor shield against, high-charge and -energy particles. Preliminary results from spaceflight, however, show that for the radiation environment ill low-Earth orbit. dominated by trapped protons, thin shields of Kevlar and Nextel provide limited reduction.

A practical neutron shielding design based on data-base interpolation

International Nuclear Information System (INIS)

Jiang, S.H.; Sheu, R.J.

1993-01-01

Neutron shielding design is an important part of the construction of nuclear reactors and high-energy accelerators. Neutron shielding design is also indispensable in the packaging and storage of isotopic neutron sources. Most efforts in the development of neutron shielding design have been concentrated on nuclear reactor shielding because of its huge mass and strict requirement of accuracy. Sophisticated computational tools, such as transport and Monte Carlo codes and detailed data libraries have been developed. In principle, now, neutron shielding, in spite of its complexity, can be designed in any detail and with fine accuracy. However, in most practical cases, neutron shielding design is accomplished with simplified methods. Unlike practical gamma-ray shielding design, where exponential attenuation coupled with buildup factors has been applied effectively and accurately, simplified neutron shielding design, either by using removal cross sections or by applying charts or tables of transmission factors such as the National Council on Radiation Protection and Measurements (NCRP) 38 (Ref. 1) for general neutron protection or to NCRP 51 (Ref. 2) for accelerator neutron shielding, is still very primitive and not well established. The available data are limited in energy range, materials, and thicknesses, and the estimated results are only roughly accurate. It is the purpose of this work to establish a simple, convenient, and user-friendly general-purpose computational tool for practical preliminary neutron shielding design that is reasonably accurate. A wide-range (energy, material, and thickness) data base of dose transmission factors has been generated by applying one-dimensional transport calculations in slab geometry

Radiation shielding and effective atomic number studies in different types of shielding concretes, lead base and non-lead base glass systems for total electron interaction: A comparative study

International Nuclear Information System (INIS)

Kurudirek, Murat

2014-01-01

Highlights: • Radiation shielding calculations for concretes and glass systems. • Assigning effective atomic number for the given materials for total electron interaction. • Glass systems generally have better shielding ability than concretes. - Abstract: Concrete has been widely used as a radiation shielding material due to its extremely low cost. On the other hand, glass systems, which make everything inside visible to observers, are considered as promising shielding materials as well. In the present work, the effective atomic numbers, Z eff of some concretes and glass systems (industrial waste containing glass, Pb base glass and non-Pb base glass) have been calculated for total electron interaction in the energy region of 10 keV–1 GeV. Also, the continuous slowing down approximation (CSDA) ranges for the given materials have been calculated in the wide energy region to show the shielding effectiveness of the given materials. The glass systems are not only compared to different types of concretes but also compared to the lead base glass systems in terms of shielding. Moreover, the obtained results for total electron interaction have been compared to the results for total photon interaction wherever possible. In general, it has been observed that the glass systems have superior properties than most of the concretes over the high-energy region with respect to the electron interaction. Also, glass systems without lead show better electron stopping than lead base glasses at some energy regions as well. Along with the photon attenuation capability, it is seen that Fly Ash base glass systems have not only greater electron stopping capability but also have greater photon attenuation especially in high energy region when compared with standard shielding concretes

Radiation shielding and effective atomic number studies in different types of shielding concretes, lead base and non-lead base glass systems for total electron interaction: A comparative study

Energy Technology Data Exchange (ETDEWEB)

Kurudirek, Murat, E-mail: mkurudirek@gmail.com

2014-12-15

Highlights: • Radiation shielding calculations for concretes and glass systems. • Assigning effective atomic number for the given materials for total electron interaction. • Glass systems generally have better shielding ability than concretes. - Abstract: Concrete has been widely used as a radiation shielding material due to its extremely low cost. On the other hand, glass systems, which make everything inside visible to observers, are considered as promising shielding materials as well. In the present work, the effective atomic numbers, Z{sub eff} of some concretes and glass systems (industrial waste containing glass, Pb base glass and non-Pb base glass) have been calculated for total electron interaction in the energy region of 10 keV–1 GeV. Also, the continuous slowing down approximation (CSDA) ranges for the given materials have been calculated in the wide energy region to show the shielding effectiveness of the given materials. The glass systems are not only compared to different types of concretes but also compared to the lead base glass systems in terms of shielding. Moreover, the obtained results for total electron interaction have been compared to the results for total photon interaction wherever possible. In general, it has been observed that the glass systems have superior properties than most of the concretes over the high-energy region with respect to the electron interaction. Also, glass systems without lead show better electron stopping than lead base glasses at some energy regions as well. Along with the photon attenuation capability, it is seen that Fly Ash base glass systems have not only greater electron stopping capability but also have greater photon attenuation especially in high energy region when compared with standard shielding concretes.

New shielding material development for compact accelerator-driven neutron source

Directory of Open Access Journals (Sweden)

Guang Hu

2017-04-01

Full Text Available The Compact Accelerator-driven Neutron Source (CANS, especially the transportable neutron source is longing for high effectiveness shielding material. For this reason, new shielding material is researched in this investigation. The component of shielding material is designed and many samples are manufactured. Then the attenuation detection experiments were carried out. In the detections, the dead time of the detector appeases when the proton beam is too strong. To grasp the linear range and nonlinear range of the detector, two currents of proton are employed in Pb attenuation detections. The transmission ratio of new shielding material, polyethylene (PE, PE + Pb, BPE + Pb is detected under suitable current of proton. Since the results of experimental neutrons and γ-rays appear as together, the MCNP and PHITS simulations are applied to assisting the analysis. The new shielding material could reduce of the weight and volume compared with BPE + Pb and PE + Pb.

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)

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)

Time-domain simulation and waveform reconstruction for shielding effectiveness of materials against electromagnetic pulse

International Nuclear Information System (INIS)

Hu, Xiao-feng; Chen, Xiang; Wei, Ming

2013-01-01

Shielding effectiveness (SE) of materials of current testing standards is often carried out by using continuous-wave measurement and amplitude-frequency characteristics curve is used to characterize the results. However, with in-depth study of high-power electromagnetic pulse (EMP) interference, it was discovered that only by frequency-domain SE of materials cannot be completely characterized by shielding performance of time-domain pulsed-field. And there is no uniform testing methods and standards of SE of materials against EMP. In this paper, the method of minimum phase transfer function is used to reconstruct shielded time-domain waveform based on the analysis of the waveform reconstruction method. Pulse of plane waves through an infinite planar material is simulated by using CST simulation software. The reconstructed waveform and simulation waveform is compared. The results show that the waveform reconstruction method based on the minimum phase can be well estimated EMP waveform through the infinite planar materials.

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

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)

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.

Preparation and properties of the fast-curing γ-ray-shielding materials based on polyurethane

Energy Technology Data Exchange (ETDEWEB)

Ni, Minxuan; Tang, Xiao Bin; Chai, Hao; Zhang, Yun; Chen, Tuo; Chen, Da [Dept. of Nuclear Science and Engineering, Nanjing University of Aeronautics and Astronautics, Nanjing (China)

2016-12-15

In this study, fast-curing shielding materials were prepared with a two-component polyurethane matrix and a filler material of PbO through a one-step, laboratory-scale method. With an increase in the filler content, viscosity increased. However, the two components showed a small difference. Curing time decreased as the filler content increased. The minimum tack-free time of 27 s was obtained at a filler content of 70 wt%. Tensile strength and compressive strength initially increased and then decreased as the filler content increased. Even when the filler content reached 60 wt%, mechanical properties were still greater than those of the matrix. Cohesional strength decreased as the filler content increased. However, cohesional strength was still greater than 100 kPa at a filler content of 60 wt%. The γ-ray-shielding properties increased with the increase in the filler content, and composite thickness could be increased to improve the shielding performance when the energy of γ-rays was high. When the filler content was 60 wt%, the composite showed excellent comprehensive properties.

Conservative method for determination of material thickness used in shielding of veterinary facilities

International Nuclear Information System (INIS)

Lava, Deise D.; Borges, Diogo da S.; Affonso, Renato R.W.; Moreira, Maria de L.; Guimaraes, Antonio C.F.

2014-01-01

For determination of an effective method for shielding of veterinary rooms, was provided shielding methods generally used in rooms which works with X-ray production and radiotherapy. Every calculation procedure is based in traditional variables used to transmission calculation. The thickness of the materials used for primary and secondary shieldings are obtained to respect the limits set by the Brazilian National Nuclear Energy Commission (CNEN). This work presents the development of a computer code in order to serve as a practical tool for determining rapid and effective materials and their thicknesses to shield veterinary facilities. The code determines transmission values of the shieldings and compares them with data from transmission 'maps' provided by NCRP-148 report. These 'maps' were added to the algorithm through interpolation techniques of curves of materials used for shielding. Each interpolation generates about 1,000,000 points that are used to generate a new curve. The new curve is subjected to regression techniques, which makes possible to obtain nine degree polynomial, and exponential equations. These equations whose variables consist of transmission of values, enable trace all the points of this curve with high precision. The data obtained from the algorithm were satisfactory with official data presented by the National Council of Radiation Protection and Measurements (NCRP) and can contribute as a practical tool for verification of shielding of veterinary facilities that require using Radiotherapy techniques and X-ray production

Gamma radiation shielding materials improved with burning resistance

International Nuclear Information System (INIS)

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

1985-01-01

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

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.

Benchmark analysis and evaluations of materials for shielding

International Nuclear Information System (INIS)

Benton, E.R.; Gersey, B.B.; Uchihori, Y.; Yasuda, N.; Kitamura, H.; Shavers, M.R.

2005-01-01

The goal of this project is to provide a benchmark set of heavy ion beam measurements behind ''standard'' targets made using radiation detectors routinely used for astronaut dosimetry and to test the radiation shielding properties of candidate multifunctional spacecraft materials. These measurements are used in testing and validating space radiation transport codes currently being developed by NASA and in selecting promising materials for further development. The radiation dosimetry instruments being used include CR-39 plastic nuclear track detector (PNTD), Tissue-Equivalent Proportional Counter (TEPC), the Liulin Mobile Dosimetry Unit (MDU) and thermoluminescent detector (TLD). Each set of measurements include LET/y spectra, and dose and dose equivalent as functions of shield thickness. Measurements are being conducted at the NIRS HIMAC, using heavy-ion beams of energy commonly encountered in the galactic cosmic ray (GCR) environment and that have been identified as being of particular concern to the radiation protection of space crews. Measurements are being made behind a set of standard'' targets including Al, Cu, polyethylene (HDPE) and graphite that vary in thickness from 0.5 to > 30 g/cm 2 . In addition, we are measuring the shielding properties of novel shielding materials being developed by and for NASA, including carbon and polymer composites. (author)

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

RADIO SHIELDING PROPERTIES OF CONCRETE BASED ON SHUNGITE NANOMATERIALS

Directory of Open Access Journals (Sweden)

BELOUSOVA Elena Sergeevna

2013-04-01

Full Text Available Modifications of shielding construction materials based on Portland cement with the addition of powder nanomaterial shungite were developed. Attenuation and re­flection of electromagnetic radiation for obtained materials were studied. Recommen­dations for using are given.

Time-domain modeling for shielding effectiveness of materials against electromagnetic pulse based on system identification

International Nuclear Information System (INIS)

Chen, Xiang; Chen, Yong Guang; Wei, Ming; Hu, Xiao Feng

2013-01-01

Shielding effectiveness (SE) of materials against electromagnetic pulse (EMP) cannot be well estimated by traditional test method of SE of materials which only consider the amplitude-frequency characteristic of materials, but ignore the phase-frequency ones. In order to solve this problem, the model of SE of materials against EMP was established based on system identification (SI) method with time-domain linear cosine frequency sweep signal. The feasibility of the method in this paper was examined depending on infinite planar material and the simulation research of coaxial test method and windowed semi-anechoic box of materials. The results show that the amplitude-frequency and phase-frequency information of each frequency can be fully extracted with this method. SE of materials against strong EMP can be evaluated with time-domain low field strength (voltage) of cosine frequency sweep signal. And SE of materials against a variety EMP will be predicted by the model.

Bulk-shield design for the Fusion Materials Irradiation Test facility

International Nuclear Information System (INIS)

Carter, L.L.; Mann, F.M.; Morford, R.J.; Johnson, D.L.; Huang, S.T.

1982-07-01

The accelerator-based Fusion Materials Irradiation Test (FMIT) facility will provide a high-fluence, fusion-like radiation environment for the testing of materials. While the neutron spectrum produced in the forward direction by the 35 MeV deuterons incident upon a flowing lithium target is characterized by a broad peak around 14 MeV, a high energy tail extends up to about 50 MeV. Some shield design considerations are reviewed

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

Science.gov (United States)

Barthel, Joseph; Sarigul-Klijn, Nesrin

2018-03-01

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

Discussion on the Standardization of Shielding Materials — Sensitivity Analysis of Material Compositions

Directory of Open Access Journals (Sweden)

Ogata Tomohiro

2017-01-01

Full Text Available The overview of standardization activities for shielding materials is described. We propose a basic approach for standardizing material composition used in radiation shielding design for nuclear and accelerator facilities. We have collected concrete composition data from actual concrete samples to organize a representative composition and its variance data. Then the sensitivity analysis of the composition variance has been performed through a simple 1-D dose calculation. Recent findings from the analysis are summarized.

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)

Gamma ray absorption of cylindrical fissile material with dual shields

International Nuclear Information System (INIS)

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

2005-01-01

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

Decision-making methodology of optimal shielding materials by using fuzzy linear programming

International Nuclear Information System (INIS)

Kanai, Y.; Miura, T.; Hirao, Y.

2000-01-01

The main purpose of our studies are to select materials and determine the ratio of constituent materials as the first stage of optimum shielding design to suit the individual requirements of nuclear reactors, reprocessing facilities, casks for shipping spent fuel, etc. The parameters of the shield optimization are cost, space, weight and some shielding properties such as activation rates or individual irradiation and cooling time, and total dose rate for neutrons (including secondary gamma ray) and for primary gamma ray. Using conventional two-valued logic (i.e. crisp) approaches, huge combination calculations are needed to identify suitable materials for optimum shielding design. Also, re-computation is required for minor changes, as the approach does not react sensitively to the computation result. Present approach using a fuzzy linear programming method is much of the decision-making toward the satisfying solution might take place in fuzzy environment. And it can quickly and easily provide a guiding principle of optimal selection of shielding materials under the above-mentioned conditions. The possibility or reducing radiation effects by optimizing the ratio of constituent materials is investigated. (author)

Material shielding of power frequency magnetic fields: Research and testing results from the EPRI Power Delivery Center-Lenox. Final report

International Nuclear Information System (INIS)

Anderson, C.B.

1998-06-01

Extensive investigations of a variety of material shielding methods have been performed at the EPRI Power Delivery Center--Lenox, Massachusetts. This work is part of a larger shielding investigation being done for EPRI by Electric Research and Management, Inc. (ERM) as part of the Magnetic Field Management Target in the EPRI Environment Group. Part of this work, involving cylinders of material, is to be included in a shielding handbook being prepared by ERM. Material shielding tests, not included in the handbook, as well as additional material shielding research, including testing, analyses, and computer simulations performed at the EPRI Power Delivery Center--Lenox are documented here. One of the major complications of using materials to shield magnetic fields is the mathematical complexity of the phenomenon involved. The result is that analytical solutions exist only for a very small number of simple geometries such as spheres, infinitely long cylinders, and infinite sheets. In practice, the materials typically come in the form of sheets. At present, there are no analytical methods for directly determining the shielding effectiveness of finite sheets of material, however, EPRI is sponsoring work in this area. There are some methods based on conformal mapping which can provide a solution for simple two-dimensional sheets. While such methods are useful in gaining insight into the mechanisms of shielding, they are not realistic enough to provide accurate shielding estimates. Empirical techniques are still required to determine the shielding effectiveness of material sheets. The material shielding tests and computer simulations are described in the report. The results of these tests and simulations have been used to develop a number of material shielding design rules for use in practical applications

Enhancement of thermal neutron self-shielding in materials surrounded by reflectors

International Nuclear Information System (INIS)

Cornelia Chilian; Gregory Kennedy

2012-01-01

Materials containing from 41 to 1124 mg chlorine and surrounded by polyethylene containers of various thicknesses, from 0.01 to 5.6 mm, were irradiated in a research reactor neutron spectrum and the 38 Cl activity produced was measured as a function of polyethylene reflector thickness. For the material containing the higher amount of chlorine, the 38 Cl specific activity decreased with increasing reflector thickness, indicating increased neutron self-shielding. It was found that the amount of neutron self-shielding increased by as much as 52% with increasing reflector thickness. This is explained by neutrons which have exited the material subsequently reflecting back into it and thus increasing the total mean path length in the material. All physical and empirical models currently used to predict neutron self-shielding have ignored this effect and need to be modified. A method is given for measuring the adjustable parameter of a self-shielding model for a particular sample size and combination of neutron reflectors. (author)

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.)

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)

JSD1000: multi-group cross section sets for shielding materials

International Nuclear Information System (INIS)

Yamano, Naoki

1984-03-01

A multi-group cross section library for shielding safety analysis has been produced by using ENDF/B-IV. The library consists of ultra-fine group cross sections, fine-group cross sections, secondary gamma-ray production cross sections and effective macroscopic cross sections for typical shielding materials. Temperature dependent data at 300, 560 and 900 K have been also provided. Angular distributions of the group to group transfer cross section are defined by a new method of ''Direct Angular Representation'' (DAR) instead of the method of finite Legendre expansion. The library designated JSD1000 are stored in a direct access data base named DATA-POOL and data manipulations are available by using the DATA-POOL access package. The 3824 neutron group data of the ultra-fine group cross sections and the 100 neutron, 20 photon group cross sections are applicable to shielding safety analyses of nuclear facilities. This report provides detailed specifications and the access method for the JSD1000 library. (author)

Hydramite II screening tests of potential bremsstrahlung converter debris shield materials

International Nuclear Information System (INIS)

Reedy, E.D. Jr.; Hedemann, M.A.; Stark, M.A.

1986-03-01

Results of a brief test series aimed at screening a number of potential bremsstrahlung converter debris shield materials are reported. These tests were run on Sandia National Laboratories' Hydramite II accelerator using a diode configuration which produces a pinched electron beam. The materials tested include: (1) laminated Kevlar 49/polyester and E-glass/polyester composites, (2) a low density laminated Kevlar 49 composite, and (3) two types of through-the-thickness reinforced Kevlar 49 composites. As expected, tests using laminated Kevlar 49/polyester shields showed that shield permanent set (i.e., permanent deflection) increased with increasing tantalum conversion foil thickness and decreased with increasing shield thickness. The through-the-thickness reinforced composites developed localized, but severe, back surface damage. The laminated composites displayed little back surface damage, although extensive internal matrix cracking and ply delaminations were generated. Roughly the same degree of permanent set was produced in shields made from the low density Kevlar 49 composite and the Kevlar 49/polyester. The E-glass reinforced shields exhibited relatively low levels of permanent set

Thermophysical Properties of Heat Resistant Shielding Material

International Nuclear Information System (INIS)

Porter, W.D.

2004-01-01

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

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.

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

International Nuclear Information System (INIS)

Nurhayati Ramli; Ahmad Nabil Abdul Rahim; Ariff Shah Ismail

2011-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-09-15

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

Studies of ionizing radiation shielding effectiveness of silica-based commercial glasses used in Bangladeshi dwellings

Directory of Open Access Journals (Sweden)

Sabina Yasmin

2018-06-01

Full Text Available Following the rapid growing economy, the Bangladeshi dwellers are replacing their traditional (mud-, bamboo-, and wood-based houses to modern multistoried buildings, where different types of glasses are being used as decorative as well as structural materials due to their various advantageous properties. In this study, we inquire the protective and dosimetric capability of commercial glasses for ionizing radiation. Four branded glass samples (PHP-Bangladesh, Osmania-Bangladesh, Nasir-Bangladesh, and Rider-China of same thickness and color but different elemental weight fractions were analyzed for shielding and dosimetric properties. The chemical composition of the studied material was evaluated by EDX technique. A well-shielded HPGe γ-ray spectrometer combined with associated electronics was used to evaluate the attenuation coefficients of the studied materials for 59 keV, 661 keV, 1173 keV and 1332 keV photon energies. A number of shielding parameters- half value layer (HVL, radiation protection efficiency (RPE and effective atomic number (Zeff were also evaluated. The data were compared with the available literature (where applicable to understand its shielding capability relative to the standard materials such as lead. Among the studied brands, Rider (China shows relatively better indices to be used as ionizing radiation shielding material. The obtained, Zeff of the studied glass samples showed comparable values to the TLD-200 dosimeter, thus considered suitable for environmental radiation monitoring purposes. Keywords: Silica-based commercial glass, HPGe γ-ray spectrometry, EDX analyses, Shielding effectiveness, Dosimetric properties

Research of glass fibre used in the electromagnetic wave shielding and absorption composite material

Science.gov (United States)

Xu, M.; Jia, F.; Bao, H. Q.; Cui, K.; Zhang, F.

2016-07-01

Electromagnetic shielding and absorption composite material plays an important role in the defence and economic field. Comparing with other filler, Glass fibre and its processed product—metal-coated glass fibre can greatly reduce the material's weight and costs, while it still remains the high strength and the electromagnetic shielding effectiveness. In this paper, the electromagnetic absorption mechanism and the reflection mechanism have been investigated as a whole, and the shielding effectiveness of the double-layer glass fibre composite material is mainly focused. The relationship between the shielding effectiveness and the filled glass fibre as well as its metal-coated product's parameters has also been studied. From the subsequent coaxial flange and anechoic chamber analysis, it can be confirmed that the peak electromagnetic shielding effectiveness of this double-layer material can reach -78dB while the bandwidth is from 2GHz to 18GHz.

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

Dosimetric characteristics of Thermo-Shield material for orthovoltage photon beams

International Nuclear Information System (INIS)

Bahmaid, Mohammad; Kim, Siyong; Liu, Chihray R.; Palta, Jatinder R.

2003-01-01

Conventionally, lead has been used for field shaping in orthovoltage radiation therapy. Recently, a compensator material named Thermo-Shield was presented for field shaping in electron beams. Thermo-Shield is composed of nontoxic, high atomic weight metal particles dispersed in a thermoplastic matrix. It is manually moldable and conforms to human anatomy or any shape at temperatures of 108-132 degree sign F. It is reusable and can be continuously reshaped to better fit the treatment field. Dosimetric characteristics of thermoplastic material were studied for Philips RT250 orthovoltage photon beams ranging from 75 to 250 kVp. It was found that Thermo-Shield should be four to five times thicker than lead to achieve the same transmission (less than 5%). However, it did not cause significant degradation in penumbra. Clinical procedures for use are discussed

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.

Double-layer neutron shield design as neutron shielding application

Science.gov (United States)

Sariyer, Demet; Küçer, Rahmi

2018-02-01

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

Development of high-performance shielding material by heat curing method

Energy Technology Data Exchange (ETDEWEB)

Miura, Toshimasa; Hirao, Yoshihiro; Hayashi, Takayuki; Okuno, Koichi; Sato, Osamu [National Maritime Research Institute, Ibaraki (Japan)

2002-07-01

A high-performance shielding material is developed by a heat curing method. It is mainly made of a thermosetting resin, lead powder, and a boron compound. To make the resin, a single functional monomer stearyl methacrylate (SMA) is used. To get good dispersion of lead and the boron compound in the resin, the viscosity of the SMA is increased by adding a small amount of a peroxide into the liquid monomer and heating up to the temperature of 100 .deg. C. Next, a peroxide, lead powder, a boron compound, a three functional monomer, and a curing accelerator are mixed into the viscous SMA. The mixture is cured in an atmosphere of nitrogen after removing bubbles using a vacuum pump. Measured properties of the cured material are as follows. The curing rate of SMA is 97 %. The density is kept 2.35 g/cm{sub 3} in the range from room temperature to 150 .deg. C. The weight-change measured by a thermogravimetry is 0.16 % in the range from room temperature to 200 .deg. C. Details of fragments in the gas released from the material is analyzed by a gas chromatography and a mass spectrometry. The hydrogen content of the material is 6.04x10 {sub 22} /cm{sub 3} . The shielding effect is calculated for a fission source by an Sn code ANISN. The shielding effect of the curing material is excellent. For example, concrete shield of a certain thickness can be replaced by the material having a thickness less than a half of concrete. Several samples of the material are irradiated at an irradiation equipment of the research reactor JRR-4 installed at Japan Atomic Energy Research Institute. At the 14{sub th} day after irradiating with the thermal neutron fluence of 6.6x10{sub 15} /cm{sub 2} , the radioactivity is less than one tenth of 75 Bq/g above which materials are regulated as the radioactive substance in Japan.

Elrotherm shielding systems. New pioneering material composites; Elrotherm-Abschirmsysteme. Neue Zukunftsweisende Materialkompositionen

Energy Technology Data Exchange (ETDEWEB)

Zika-Beyerlein, B [ElringKlinger (Germany). Geschaeftsbereich Abschirmtechnik

2004-09-01

Tightly packed engine compartments put special demands on thermal and acoustic shielding systems. With new material composites allowing for particularly thin-walled and light shielding parts, ElringKlinger is well equipped for the future. (orig.)

Shield materials recommended for space power nuclear reactors

Science.gov (United States)

Kaszubinski, L. J.

1973-01-01

Lithium hydride is recommended for neutron attenuation and depleted uranium is recommended for gamma ray attenuation. For minimum shield weights these materials must be arranged in alternate layers to attenuate the secondary gamma rays efficiently. In the regions of the shield near the reactor, where excessive fissioning occurs in the uranium, a tungsten alloy is used instead. Alloys of uranium such as either the U-0.5Ti or U-8Mo are available to accommodate structural requirements. The zone-cooled casting process is recommended for lithium hydride fabrication. Internal honeycomb reinforcement to control cracks in the lithium hydride is recommended.

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)

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

International Nuclear Information System (INIS)

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

2013-01-01

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

Neutron shielding material and a process for producing the same

International Nuclear Information System (INIS)

Tadokoro, S.; Segawa, H.

1982-01-01

A neutron shielding material comprises a polymerization product of a monomer mixture of an alkyl methacrylate or styrene and a boric acid containing a polyol constituent. Such a material may be formed into transparent sheets with high mechanical strength

About the Scythian Shields

Directory of Open Access Journals (Sweden)

About the Scythian Shields

2017-10-01

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

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)

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

Data.gov (United States)

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

Identification of Shielding Material Configurations Using NMIS Imaging

International Nuclear Information System (INIS)

Grogan, Brandon R.; Mihalczo, John T.; McConchie, Seth M.; Mullens, James Allen

2011-01-01

The Nuclear Materials Identification System (NMIS) uses fast neutron tomographic imaging to nonintrusively examine the interior structure of shielded objects. The pixel values in such images represent the attenuation coefficients of the time- and directionally-tagged fast neutrons from a deuterium-tritium (D T) neutron generator. The reconstruction techniques use either a filtered back projection or a maximum likelihood expectation maximization algorithm. As a first test of the capabilities of these reconstruction techniques to correctly identify individual parts inside of an object, fast neutron imaging was used to identify the regions of shielding surrounding a depleted uranium casting from a library of possible parts. The shielding consisted of multiple regions of common materials such as steel, lead, aluminum, and polyethylene. First, the full object was imaged, and then each of the individual parts was imaged. Several additional parts that were not present in the original object were also imaged to form a library. The individual parts were compared to the full object, and the correct ones were identified using three different methods. These methods included a visual match, an iterative fit of each part, and a mathematical test comparing the sum of squared errors. The successful results demonstrate an initial application of matching. This suggests that it should be possible to implement more sophisticated matching techniques using automated pixel-by-pixel comparison methods in the future.

{sup 3}He detector analysis of some special shielding materials

Energy Technology Data Exchange (ETDEWEB)

Avdic, S; Pesic, M [Boris Kidric, Institute of Nuclear Sciences, Beograd (Yugoslavia); Marinkovic, P [ETF Belgrade Univ. (Yugoslavia)

1990-07-01

The shielding properties of commercial materials of reactor Experiments, Inc. (R/X) were analyzed at the facility which includes bare heavy water experimental reactor RB with external neutron converter ENC, The fast neutron spectrum measurements in energy range from 1 MeV to 10 MeV was performed using ORTEC semiconductor neutron detector with He{sup 3} in diode coincidence arrangement. The neutron spectra have been evaluated from measured pulse-height distribution using numerical code HE3 for computation of detector efficiency in a collimated neutron beam. The neutron dose rates behind ENC with and without sample R/X material were determined using cubic spline interpolation routine for calculating the corresponding flux-dose rate conversion factors. Satisfactory shielding properties of the examined material in a fast neutron field in measurements and calculations are demonstrated. (author)

Determination of boron in Jabroc wood used as a shielding material in nuclear reactors

International Nuclear Information System (INIS)

Kamble, Granthali S.; Manisha, V.; Venkatesh, K.

2015-01-01

Jabroc are non-impregnated, densified wood laminates developed commercially for a wide range of industrial applications. Jabroc can be used with other neutron shielding materials such as Lead to form complex shielding structures. Its relative light weight and cleanliness in handling are additional features that make it a suitable candidate for the standard design of neutron shielding equipment. Jabroc can also be impregnated with Boron up to a maximum of 4% to be used in areas where Gamma radiation produced on Neutron capture reaches unacceptable dose rates. Boron impregnated Jabroc wood finds application in TAPS 3 and 4 as a shielding material for the Ion Chambers and the Horizontal Flux Units (HFU). The shielding property of this material is optimized by incorporating requisite amount of boron in wood. Boron content in this material has to be determined accurately prior to its use in the nuclear reactors. In this work a method was standardized to determine boron in Jabroc wood samples to check for conformance to specifications. The wood sample flakes were wetted with saturated barium hydroxide solution and dries under IR. The sample was ashed in a muffle furnace at 600℃ for 2 h

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

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

Science.gov (United States)

Thibeault, Sheila A.; Fay, Catharine C.; Lowther, Sharon E.; Earle, Kevin D.; Sauti, Godfrey; Kang, Jin Ho; Park, Cheol; McMullen, Amelia M.

2012-01-01

than possible with hydrogen storage; however, a systematic experimental hydrogenation study has not been reported. A combination of the two approaches may be explored to provide yet higher hydrogen content. The hydrogen containing BNNT produced in our study will be characterized for hydrogen content and thermal stability in simulated space service environments. These new materials systems will be tested for their radiation shielding effectiveness against high energy protons and high energy heavy ions at the HIMAC facility in Japan, or a comparable facility. These high energy particles simulate exposure to SEP and GCR environments. They will also be tested in the LaRC Neutron Exposure Laboratory for their neutron shielding effectiveness, an attribute that determines their capability to shield against the secondary neutrons found inside structures and on lunar and planetary surfaces. The potential significance is to produce a radiation protection enabling technology for future exploration missions. Crew on deep space human exploration missions greater than approximately 90 days cannot remain below current crew Permissible Exposure Limits without shielding and/or biological countermeasures. The intent of this research is to bring the Agency closer to extending space missions beyond the 90-day limit, with 1 year as a long-term goal. We are advocating a systems solution with a structural materials component. Our intent is to develop the best materials system for that materials component. In this Phase I study, we have shown, computationally, that hydrogen containing BNNT is effective for shielding against GCR, SEP, and neutrons over a wide range of energies. This is why we are focusing on hydrogen containing BNNT as an innovative advanced concept. In our future work, we plan to demonstrate, experimentally, that hydrogen, boron, and nitrogen based materials can provide mechanically strong, thermally stable, structural materials with effective radiation shielding against GCR

Assessment of radiation shielding materials for protection of space crews using CR-39 plastic nuclear track detector

International Nuclear Information System (INIS)

DeWitt, J.M.; Benton, E.R.; Uchihori, Y.; Yasuda, N.; Benton, E.V.; Frank, A.L.

2009-01-01

A significant obstacle to long duration human space exploration such as the establishment of a permanent base on the surface of the Moon or a human mission to Mars is the risk posed by prolonged exposure to space radiation. In order to keep mission costs at acceptable levels while simultaneously minimizing the risk from radiation to space crew health and safety, a judicious use of optimized shielding materials will be required. We have undertaken a comprehensive study using CR-39 plastic nuclear track detector (PNTD) to characterize the radiation shielding properties of a range of materials-both common baseline materials such as Al and polyethylene, and novel multifunctional materials such as carbon composites-at heavy ion accelerators. The study consists of analyzing CR-39 PNTD exposed in front of and behind shielding targets of varying composition and at a number of depths (target thicknesses) relevant to the development and testing of materials for space radiation shielding. Most targets consist of 10 cm x 10 cm slabs of solid materials ranging in thickness from 1 to >30 g/cm 2 . Exposures have been made to beams of C, O, Ne, Si, Ar, and Fe at energies ranging from 290 MeV/amu to 1 GeV/amu at the National Institute of Radiological Sciences HIMAC and the NASA Space Radiation Laboratory (NSRL) at Brookhaven National Laboratory. Analysis of the exposed detectors yields LET spectrum, dose, and dose equivalent as functions of target depth and composition, and incident heavy ion charge, energy, and fluence. Efforts are currently underway to properly weigh and combine these results into a single quantitative estimate of a material's ability to shield space crews from the interplanetary galactic cosmic ray flux.

Shield design for the Fusion Materials Irradiation Test facility

International Nuclear Information System (INIS)

Carter, L.L.; Mann, F.M.; Morford, R.J.; Wilcox, A.D.; Johnson, D.L.; Huang, S.T.

1983-03-01

The shield design for the Fusion Materials Irradiation Test facility is based upon one-, two- and three-dimensional transport calculations with experimental measurements utilized to refine the nuclear data including the neutron cross sections from 20 to 50 MeV and the gamma ray and neutron source terms. The high energy neutrons and deuterons produce activation products from the numerous reactions that are kinematically allowed. The analyses for both beam-on and beam-off (from the activation products) conditions have required extensive nuclear data libraries and the utilization of Monte Carlo, discrete ordinates, point kernel and auxiliary computer codes

Depleted uranium hexafluoride: The source material for advanced shielding systems

Energy Technology Data Exchange (ETDEWEB)

Quapp, W.J.; Lessing, P.A. [Idaho National Engineering Lab., Idaho Falls, ID (United States); Cooley, C.R. [Department of Technology, Germantown, MD (United States)

1997-02-01

The U.S. Department of Energy (DOE) has a management challenge and financial liability problem in the form of 50,000 cylinders containing 555,000 metric tons of depleted uranium hexafluoride (UF{sub 6}) that are stored at the gaseous diffusion plants. DOE is evaluating several options for the disposition of this UF{sub 6}, including continued storage, disposal, and recycle into a product. Based on studies conducted to date, the most feasible recycle option for the depleted uranium is shielding in low-level waste, spent nuclear fuel, or vitrified high-level waste containers. Estimates for the cost of disposal, using existing technologies, range between $3.8 and $11.3 billion depending on factors such as the disposal site and the applicability of the Resource Conservation and Recovery Act (RCRA). Advanced technologies can reduce these costs, but UF{sub 6} disposal still represents large future costs. This paper describes an application for depleted uranium in which depleted uranium hexafluoride is converted into an oxide and then into a heavy aggregate. The heavy uranium aggregate is combined with conventional concrete materials to form an ultra high density concrete, DUCRETE, weighing more than 400 lb/ft{sup 3}. DUCRETE can be used as shielding in spent nuclear fuel/high-level waste casks at a cost comparable to the lower of the disposal cost estimates. Consequently, the case can be made that DUCRETE shielded casks are an alternative to disposal. In this case, a beneficial long term solution is attained for much less than the combined cost of independently providing shielded casks and disposing of the depleted uranium. Furthermore, if disposal is avoided, the political problems associated with selection of a disposal location are also avoided. Other studies have also shown cost benefits for low level waste shielded disposal containers.

On stress/strain shielding and the material stiffness paradigm for dental implants.

Science.gov (United States)

Korabi, Raoof; Shemtov-Yona, Keren; Rittel, Daniel

2017-10-01

Stress shielding considerations suggest that the dental implant material's compliance should be matched to that of the host bone. However, this belief has not been confirmed from a general perspective, either clinically or numerically. To characterize the influence of the implant stiffness on its functionality using the failure envelope concept that examines all possible combinations of mechanical load and application angle for selected stress, strain and displacement-based bone failure criteria. Those criteria represent bone yielding, remodeling, and implant primary stability, respectively MATERIALS AND METHODS: We performed numerical simulations to generate failure envelopes for all possible loading configurations of dental implants, with stiffness ranging from very low (polymer) to extremely high, through that of bone, titanium, and ceramics. Irrespective of the failure criterion, stiffer implants allow for improved implant functionality. The latter reduces with increasing compliance, while the trabecular bone experiences higher strains, albeit of an overall small level. Micromotions remain quite small irrespective of the implant's stiffness. The current paradigm favoring reduced implant material's stiffness out of concern for stress or strain shielding, or even excessive micromotions, is not supported by the present calculations, that point exactly to the opposite. © 2017 Wiley Periodicals, Inc.

Studies of ionizing radiation shielding effectiveness of silica-based commercial glasses used in Bangladeshi dwellings

Science.gov (United States)

Yasmin, Sabina; Barua, Bijoy Sonker; Khandaker, Mayeen Uddin; Chowdhury, Faruque-Uz-Zaman; Rashid, Md. Abdur; Bradley, David A.; Olatunji, Michael Adekunle; Kamal, Masud

2018-06-01

Following the rapid growing economy, the Bangladeshi dwellers are replacing their traditional (mud-, bamboo-, and wood-based) houses to modern multistoried buildings, where different types of glasses are being used as decorative as well as structural materials due to their various advantageous properties. In this study, we inquire the protective and dosimetric capability of commercial glasses for ionizing radiation. Four branded glass samples (PHP-Bangladesh, Osmania-Bangladesh, Nasir-Bangladesh, and Rider-China) of same thickness and color but different elemental weight fractions were analyzed for shielding and dosimetric properties. The chemical composition of the studied material was evaluated by EDX technique. A well-shielded HPGe γ-ray spectrometer combined with associated electronics was used to evaluate the attenuation coefficients of the studied materials for 59 keV, 661 keV, 1173 keV and 1332 keV photon energies. A number of shielding parameters- half value layer (HVL), radiation protection efficiency (RPE) and effective atomic number (Zeff) were also evaluated. The data were compared with the available literature (where applicable) to understand its shielding capability relative to the standard materials such as lead. Among the studied brands, Rider (China) shows relatively better indices to be used as ionizing radiation shielding material. The obtained, Zeff of the studied glass samples showed comparable values to the TLD-200 dosimeter, thus considered suitable for environmental radiation monitoring purposes.

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.

Evaluation of neutron shielding made of cement type material

International Nuclear Information System (INIS)

Seshimo, Takuya; Nagai, Takayuki; Onose, Atsushi; Takuma, Yasuhisa; Tanuma, Hiroyuki; Otagawa, Masaaki

1998-01-01

We prepared boron-containing cement and evaluated the characteristics of this new cement. This is the material of neutron shielding which is lighter than existing one. The quality we aimed is: H ≥ 0.025 g/cm 3 , B ≥ 0.065 g/cm 3 , density ≤ 1.70 g/cm 3 . We made test pieces changing water powder ratio (W/P), adding amount of air entraining agent, adding amount of water reducing agent, and time of vibration, and then, evaluated the characteristics. The measured parameters are the air content, mortar flow and homogeneity for cement mortar, homogeneity and compressive strength for hardened one. From the results of these tests, we confirmed the possibility of making neutron shielding that can satisfy the aimed quality using this boron-containing cement. After all, we established the method of making the neutron shielding, and this method was used in the construction of RETF. (author)

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.)

Electromagnetic interference shielding with Portland cement paste containing carbon materials and processed fly ash

Directory of Open Access Journals (Sweden)

Zornoza, E.

2010-12-01

Full Text Available The study described in this article explored the effect of adding different types of carbon materials (graphite powder and three types of carbon fibre, fly ash (with 5.6%, 15.9% and 24.3% Fe₂O₃, and a mix of both on electromagnetic interference (EMI shielding in Portland cement pastes. The parameters studied included the type and aspect ratio of the carbonic material, composite material thickness, the frequency of the incident electromagnetic radiation and the percentage of the magnetic fraction in the fly ash. The findings showed that the polyacrylonitrile-based carbon fibres, which had the highest aspect ratio, provided more effective shielding than any of the other carbon materials studied. Shielding was more effective in thicker specimens and at higher radiation frequencies. Raising the magnetic fraction of the fly ash, in turn, also enhanced paste shielding performance. Finally, adding both carbon fibre and fly ash to the paste resulted in the most effective EMI shielding as a result of the synergies generated.

En el presente trabajo se investiga la influencia de la adición de diferentes tipos de materiales carbonosos (polvo de grafito y 3 tipos de fibra de carbono, de una ceniza volante con diferentes contenidos de fase magnética (5,6%, 15,9% y 24,3% de Fe₂O₃ y de una mezcla de ambos, sobre la capacidad de apantallar interferencias electromagnéticas de pastas de cemento Pórtland. Entre los parámetros estudiados se encuentra: el tipo de material carbonoso, la relación de aspecto del material carbonoso, el espesor del material compuesto, la frecuencia de la radiación electromagnética incidente y el porcentaje de fracción magnética en la ceniza volante. Los resultados obtenidos indican que entre los materiales carbonosos estudiados son las fibras de carbono basadas en poliacrilonitrilo con una mayor relación de aspecto las que dan mejores resultados de apantallamiento. Al aumentar

SHIELD 1.0: development of a shielding calculator program in diagnostic radiology; SHIELD 1.0: desenvolvimento de um programa de calculo de blindagem em radiodiagnostico

Energy Technology Data Exchange (ETDEWEB)

Santos, Romulo R.; Real, Jessica V.; Luz, Renata M. da [Hospital Sao Lucas (PUCRS), Porto Alegre, RS (Brazil); Friedrich, Barbara Q.; Silva, Ana Maria Marques da, E-mail: ana.marques@pucrs.br [Pontificia Universidade Catolica do Rio Grande do Sul (PUCRS), Porto Alegre, RS (Brazil)

2013-08-15

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

Investigation of Turkish marbles as shielding materials

International Nuclear Information System (INIS)

Atasoy, H.; Tarcan, G.; Doekmen, S.

1992-01-01

The natural Turkish marbles, especially Usak Green (UG), Aegean Purple (AP), and Marmara White (MW) were tested as shielding materials using standard gamma sources such as Co-60, Cs-137 and Eu-152. The experiment showed that UG, AP and MW are very effective shields against gamma-rays. The result for this experiment is that the gamma-ray attenuation coefficients of UG, AP and MW are almost equal for the energy range from 0.1 MeV to 1.4 MeV. Also, the elemental compositions of the natural UG, AP and MW marbles have been determined by fast and thermal neutron activation analysis and fourteen elements including Na, Mg, Al, Si, Cl, K, Ca, V, Ti, Mn, Fe, La, Ba and Sc have been found using the gamma spectroscopic method. The range of element contents of all Turkish marbles are remarkably different, but most of the elements are common such as Ca, Fe, Na, Cl, Mg, Si. (orig.)

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)

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.

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.

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

Data.gov (United States)

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

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.

Practical high-density shielding materials for medical linear accelerator rooms

International Nuclear Information System (INIS)

Barish, R.J.

1990-01-01

High-energy linear accelerators are replacing lower energy units in radiation therapy centers. Radiation protection requirements necessitate expensive reconstruction of existing treatment rooms to accommodate these new machines. We describe two shielding materials: one made by embedding small pieces of scrap steel in cement, and the other made with cast iron in cement. Both materials produce high-density barriers at low cost using standard construction methods

Secondary gamma-ray data for shielding calculation

International Nuclear Information System (INIS)

Miyasaka, Sunichi

1979-01-01

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

Potential Use of In Situ Material Composites such as Regolith/Polyethylene for Shielding Space Radiation

Science.gov (United States)

Theriot, Corey A.; Gersey, Buddy; Bacon, Eugene; Johnson, Quincy; Zhang, Ye; Norman, Jullian; Foley, Ijette; Wilkins, Rick; Zhou, Jianren; Wu, Honglu

2010-01-01

NASA has an extensive program for studying materials and methods for the shielding of astronauts to reduce the effects of space radiation when on the surfaces of the Moon and Mars, especially in the use of in situ materials native to the destination reducing the expense of materials transport. The most studied material from the Moon is Lunar regolith and has been shown to be as efficient as aluminum for shielding purposes (1). The addition of hydrogenous materials such as polyethylene should increase shielding effectiveness and provide mechanical properties necessary of structural materials (2). The neutron radiation shielding effectiveness of polyethylene/regolith stimulant (JSC-1A) composites were studied using confluent human fibroblast cell cultures exposed to a beam of high-energy spallation neutrons at the 30deg-left beam line (ICE house) at the Los Alamos Neutron Science Center. At this angle, the radiation spectrum mimics the energy spectrum of secondary neutrons generated in the upper atmosphere and encountered when aboard spacecraft and high-altitude aircraft. Cell samples were exposed in series either directly to the neutron beam, within a habitat created using regolith composite blocks, or behind 25 g/sq cm of loose regolith bulk material. In another experiment, cells were also exposed in series directly to the neutron beam in T-25 flasks completely filled with either media or water up to a depth of 20 cm to test shielding effectiveness versus depth and investigate the possible influence of secondary particle generation. All samples were sent directly back to JSC for sub-culturing and micronucleus analysis. This presentation is of work performed in collaboration with the NASA sponsored Center for Radiation Engineering and Science for Space Exploration (CRESSE) at Prairie View A&M.

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

Data.gov (United States)

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

Research on shielding neutron efficiency of some boron-bearing fabric and transparent resin materials

International Nuclear Information System (INIS)

Chen Changmao; Liu Jinhua; Su Jingling; Wang Zheng

1995-01-01

The shielding neutron efficiency of boron-bearing materials developed recently is introduced. The thermal neutron shield ratios for two kinds of non-woven cloth with thickness of 58 mg/cm 2 and 153 mg/cm 2 are 51% and 79% respectively. Their mass attenuation coefficient for 0.186, 24.4 and 144 keV neutron are 1.56, 1.29 and 0.9 cm 2 /g respectively. The thermal neutron shield ratio is 85% for the natural boron-bearing transparent resin plate with the thickness of 0.59 g/cm 2 , and 97% for enriched boron or gadolinium bearing resin plate. The shield ratios of all three materials for 24.4 keV neutrons are 38%. The transparence of natural light for enriched boron-bearing resin plates shows no considerable change after they were exposed to thermal neutrons up to 6 Sv. After they were exposed up to 20 Sv, the transparence decreases to 50% but thermal neutron shield ratio does not change. The gadolinium-bearing plate has a very strong thermal neutron-capture gamma radiation and its dose-equivalent is greater than that of incident thermal neutrons

The distribution property of shielding materials used in radiological protection

International Nuclear Information System (INIS)

Costa, Paulo R.

1996-01-01

A parametric model for the distribution of the radiation scattered by shielding materials is proposed. The advantages regarding to chemical composition or to linear attenuation coefficient of the media as a function of energy are presented. Comparative results between the model and experimental data are reported

Magnetic Materials Characterization and Modeling for the Enhanced Design of Magnetic Shielding of Cryomodules in Particle Accelerators

Energy Technology Data Exchange (ETDEWEB)

Sah, Sanjay [Virginia Commonwealth Univ., Richmond, VA (United States)

2016-05-31

Particle accelerators produce beams of high-energy particles, which are used for both fundamental and applied scientific research and are critical to the development of accelerator driven sub-critical reactor systems. An effective magnetic shield is very important to achieve higher quality factor (Qo) of the cryomodule of a particle accelerator. The allowed value of field inside the cavity due to all external fields (particularly the Earth’s magnetic field) is ~15 mG or less. The goal of this PhD dissertation is to comprehensively study the magnetic properties of commonly used magnetic shielding materials at both cryogenic and room temperatures. This knowledge can be used for the enhanced design of magnetic shields of cryomodes (CM) in particle accelerators. To this end, we first studied the temperature dependent magnetization behavior (M-H curves) of Amumetal and A4K under different annealing and deformation conditions. This characterized the effect of stress or deformation induced during the manufacturing processes and subsequent restoration of high permeability with appropriate heat treatment. Next, an energy based stochastic model for temperature dependent anhysteretic magnetization behavior of ferromagnetic materials was proposed and benchmarked against experimental data. We show that this model is able to simulate and explain the magnetic behavior of as rolled, deformed and annealed amumetal and A4K over a large range of temperatures. The experimental results for permeability are then used in a finite element model (FEM) in COMSOL to evaluate the shielding effectiveness of multiple shield designs at room temperature as well as cryogenic temperature. This work could serve as a guideline for future design, development and fabrication of magnetic shields of CMs.

γ-ray shielding behaviors of some nuclear engineering materials

International Nuclear Information System (INIS)

Mann, Kulwinder Singh

2017-01-01

The essential requirement of a material to be used for engineering purposes at nuclear establishments is its ability to attenuate the most penetrating ionizing radiations, gamma (γ)-rays. Mostly, high-Z materials such as heavy concrete, lead, mercury, and their mixtures or alloys have been used in the construction of nuclear establishments and thus termed as nuclear engineering materials (NEM). The NEM are classified into two categories, namely opaque and transparent, depending on their behavior towards the visible spectrum of EM waves. The majority of NEM are opaque. By contrast, various types of glass, which are transparent to visible light, are necessary at certain places in the nuclear establishments. In the present study, γ-ray shielding behaviors (GSB) of six glass samples (transparent NEM) were evaluated and compared with some opaque NEM in a wide range of energy (15 keV–15 MeV) and optical thickness (OT). The study was performed by computing various γ-ray shielding parameters (GSP) such as the mass attenuation coefficient, equivalent atomic number, and buildup factor. A self-designed and validated computer-program, the buildup factor-tool, was used for various computations. It has been established that some glass samples show good GSB, thus can safely be used in the construction of nuclear establishments in conjunction with the opaque NEM as well

γ-ray shielding behaviors of some nuclear engineering materials

Energy Technology Data Exchange (ETDEWEB)

Mann, Kulwinder Singh [Dept. of Physics, D.A.V. College, Punjab (India)

2017-06-15

The essential requirement of a material to be used for engineering purposes at nuclear establishments is its ability to attenuate the most penetrating ionizing radiations, gamma (γ)-rays. Mostly, high-Z materials such as heavy concrete, lead, mercury, and their mixtures or alloys have been used in the construction of nuclear establishments and thus termed as nuclear engineering materials (NEM). The NEM are classified into two categories, namely opaque and transparent, depending on their behavior towards the visible spectrum of EM waves. The majority of NEM are opaque. By contrast, various types of glass, which are transparent to visible light, are necessary at certain places in the nuclear establishments. In the present study, γ-ray shielding behaviors (GSB) of six glass samples (transparent NEM) were evaluated and compared with some opaque NEM in a wide range of energy (15 keV–15 MeV) and optical thickness (OT). The study was performed by computing various γ-ray shielding parameters (GSP) such as the mass attenuation coefficient, equivalent atomic number, and buildup factor. A self-designed and validated computer-program, the buildup factor-tool, was used for various computations. It has been established that some glass samples show good GSB, thus can safely be used in the construction of nuclear establishments in conjunction with the opaque NEM as well.

γ-Ray Shielding Behaviors of Some Nuclear Engineering Materials

Directory of Open Access Journals (Sweden)

Kulwinder Singh Mann

2017-06-01

Full Text Available The essential requirement of a material to be used for engineering purposes at nuclear establishments is its ability to attenuate the most penetrating ionizing radiations, gamma (γ-rays. Mostly, high-Z materials such as heavy concrete, lead, mercury, and their mixtures or alloys have been used in the construction of nuclear establishments and thus termed as nuclear engineering materials (NEM. The NEM are classified into two categories, namely opaque and transparent, depending on their behavior towards the visible spectrum of EM waves. The majority of NEM are opaque. By contrast, various types of glass, which are transparent to visible light, are necessary at certain places in the nuclear establishments. In the present study, γ-ray shielding behaviors (GSB of six glass samples (transparent NEM were evaluated and compared with some opaque NEM in a wide range of energy (15 keV–15 MeV and optical thickness (OT. The study was performed by computing various γ-ray shielding parameters (GSP such as the mass attenuation coefficient, equivalent atomic number, and buildup factor. A self-designed and validated computer-program, the buildup factor-tool, was used for various computations. It has been established that some glass samples show good GSB, thus can safely be used in the construction of nuclear establishments in conjunction with the opaque NEM as well.

Simulation of photon attenuation coefficients for high effective shielding material Lead-Boron Polyethyene

Science.gov (United States)

Zhang, L.; Jia, M. C.; Gong, J. J.; Xia, W. M.

2017-12-01

The mass attenuation coefficient of various Lead-Boron Polyethylene samples which can be used as the photon shielding materials in marine reactor, have been simulated using the MCNP-5 code, and compared with the theoretical values at the photon energy range 0.001MeV—20MeV. A good agreement has been observed. The variations of mass attenuation coefficient, linear attenuation coefficient and mean free path with photon energy between 0.001MeV to 100MeV have been plotted. The result shows that all the coefficients strongly depends on the photon energy, material atomic composition and density. The dose transmission factors for source Cesium-137 and Cobalt-60 have been worked out and their variations with the thickness of various sample materials have also been plotted. The variations show that with the increase of materials thickness the dose transmission factors decrease continuously. The results of this paper can provide some reference for the use of the high effective shielding material Lead-Boron Polyethyene.

A comparative study of two digestion methods employed for the determination boron in ferroboron used as an advanced shielding material

International Nuclear Information System (INIS)

Kamble, Granthali S.; Manisha, V.; Venkatesh, K.

2015-01-01

Shielding of nuclear reactor core is an important requirement of fast reactors. An important objective of future Fast Breeder Reactors (FBRs) is to reduce the volume of shields. A large number of materials have been considered for use to reduce the neutron flux to acceptable levels. A shield material which brings down the energy of neutrons by elastic and inelastic scattering along with absorption will be more effective. Ferro boron is identified as one of the advanced shielding materials considered for use in future FBRs, planned to be constructed in India. Ferroboron is an economical and indigenously available material which qualifies as a promising shield material through literature survey and scoping calculations. Experiments have been conducted in KAMINI reactor to understand the effectiveness of prospective shield material Ferro-boron as an in-core shield material for future FBRs. The Ferro boron used in these experiments contained 11.8% and 15% of boron. Precise determination of boron content in these ferro boron samples is very important to determine its effectiveness as a shield material. In this work a comparative study was carried out to determine the boron content in ferro boron samples. In the first method the sample was treated with incremental amounts of nitric acid under reflux (to prevent rigorous reaction and volatalisation of boron). The solution was gradually heated and the solution was filtered through a Whatman Filter paper no. 41. The undissolved ferro boron residue collected in the filter paper after filtration, is transferred to a platinum crucible; mixed with sodium carbonate and is ashed. The crucible is placed over a burner for 1 h to fuse the contents. The fused mass is leached in dilute hydrochloric acid, added to the nitric acid filtrate and made up to pre-determined volume

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

Electromagnetic shielding

International Nuclear Information System (INIS)

Tzeng, Wen-Shian V.

1991-01-01

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

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.)

Nuclear reactions and self-shielding effects of gamma-ray database for nuclear materials

Energy Technology Data Exchange (ETDEWEB)

Fujita, Mitsutane; Noda, Tetsuji [National Research Institute for Metals, Tsukuba, Ibaraki (Japan)

2001-03-01

A database for transmutation and radioactivity of nuclear materials is required for selection and design of materials used in various nuclear reactors. The database based on the FENDL/A-2.0 on the Internet and the additional data collected from several references has been developed in NRIM site of 'Data-Free-Way' on the Internet. Recently, the function predicted self-shielding effect of materials for {gamma}-ray was added to this database. The user interface for this database has been constructed for retrieval of necessary data and for graphical presentation of the relation between the energy spectrum of neutron and neutron capture cross section. It is demonstrated that the possibility of chemical compositional change and radioactivity in a material caused by nuclear reactions can be easily retrieved using a browser such as Netscape or Explorer. (author)

Nuclear reactions and self-shielding effects of gamma-ray database for nuclear materials

International Nuclear Information System (INIS)

Fujita, Mitsutane; Noda, Tetsuji

2001-01-01

A database for transmutation and radioactivity of nuclear materials is required for selection and design of materials used in various nuclear reactors. The database based on the FENDL/A-2.0 on the Internet and the additional data collected from several references has been developed in NRIM site of 'Data-Free-Way' on the Internet. Recently, the function predicted self-shielding effect of materials for γ-ray was added to this database. The user interface for this database has been constructed for retrieval of necessary data and for graphical presentation of the relation between the energy spectrum of neutron and neutron capture cross section. It is demonstrated that the possibility of chemical compositional change and radioactivity in a material caused by nuclear reactions can be easily retrieved using a browser such as Netscape or Explorer. (author)

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...

Evaluation of Neutron shielding efficiency of Metal hydrides

Energy Technology Data Exchange (ETDEWEB)

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

2012-05-15

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

EBT-P gamma-ray-shielding analysis

International Nuclear Information System (INIS)

Gohar, Y.

1983-01-01

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

Magnetic shielding for superconducting RF cavities

Science.gov (United States)

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

2017-03-01

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

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.

Monte Carlo simulation of photon buildup factors for shielding materials in diagnostic x-ray facilities

International Nuclear Information System (INIS)

Kharrati, Hedi; Agrebi, Amel; Karoui, Mohamed Karim

2012-01-01

Purpose: A simulation of buildup factors for ordinary concrete, steel, lead, plate glass, lead glass, and gypsum wallboard in broad beam geometry for photons energies from 10 keV to 150 keV at 5 keV intervals is presented. Methods: Monte Carlo N-particle radiation transport computer code has been used to determine the buildup factors for the studied shielding materials. Results: An example concretizing the use of the obtained buildup factors data in computing the broad beam transmission for tube potentials at 70, 100, 120, and 140 kVp is given. The half value layer, the tenth value layer, and the equilibrium tenth value layer are calculated from the broad beam transmission for these tube potentials. Conclusions: The obtained values compared with those calculated from the published data show the ability of these data to predict shielding transmission curves. Therefore, the buildup factors data can be combined with primary, scatter, and leakage x-ray spectra to provide a computationally based solution to broad beam transmission for barriers in shielding x-ray facilities.

Monte Carlo simulation of photon buildup factors for shielding materials in diagnostic x-ray facilities.

Science.gov (United States)

Kharrati, Hedi; Agrebi, Amel; Karoui, Mohamed Karim

2012-10-01

A simulation of buildup factors for ordinary concrete, steel, lead, plate glass, lead glass, and gypsum wallboard in broad beam geometry for photons energies from 10 keV to 150 keV at 5 keV intervals is presented. Monte Carlo N-particle radiation transport computer code has been used to determine the buildup factors for the studied shielding materials. An example concretizing the use of the obtained buildup factors data in computing the broad beam transmission for tube potentials at 70, 100, 120, and 140 kVp is given. The half value layer, the tenth value layer, and the equilibrium tenth value layer are calculated from the broad beam transmission for these tube potentials. The obtained values compared with those calculated from the published data show the ability of these data to predict shielding transmission curves. Therefore, the buildup factors data can be combined with primary, scatter, and leakage x-ray spectra to provide a computationally based solution to broad beam transmission for barriers in shielding x-ray facilities.

Detail AR2-PR1 geochronological scale of the Baltic Shield

International Nuclear Information System (INIS)

Balashov, Yu.A.

1995-01-01

The version of the detail scale of the Precambrian Baltic Shield, based on selection of new crust increment through mantel genesis material is presented. The AR 2 -PR 1 -scale of the Baltic Shield is based through U-Pb, Rb-Sr and Sm-Nd-methods for the Cord Peninsula on the Precambrian period of the other Baltic Shield regions

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-03-15

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

Radiation shielding material and method of fabricating the same

International Nuclear Information System (INIS)

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

1979-01-01

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

Development and qualification of materials and processes for radiation shielding of Galileo spacecraft electronic components

International Nuclear Information System (INIS)

Hribar, F.; Bauer, J.L.; O'Donnell, T.P.

1990-01-01

Several materials and processing methods were evaluated for use on the JPL Galileo spacecraft in the area of radiation shielding for electronics. Development and qualification activities involving an aluminum structural laminate are described. These activities included requirements assessment, design tradeoffs, materials selection, adhesive bonding development, mechanical properties measurements, thermal stability assessment, and nondestructive evaluation. This paper presents evaluation of three adhesives for bonding tantalum to aluminum. The concept of combining a thin sheet of tantalum with two outer aluminum face sheets using adhesive bonding was developed successfully. This radiation shield laminate also provides a structural shear plate for mounting electronic assemblies

Shielding property of bismuth glass based on MCNP 5 and WINXCOM simulated calculation

International Nuclear Information System (INIS)

Zhang Zhicheng; Zhang Jinzhao; Liu Ze; Lu Chunhai; Chen Min

2013-01-01

Background: Currently, lead glass is widely used as observation window, while lead is toxic heavy metal. Purpose: Non-toxic materials and their shielding effects are researched in order to find a new material to replace lead containing material. Methods: The mass attenuation coefficients of bismuth silicate glass were investigated with gamma-ray's energy at 0.662 MeV, 1.17 MeV and 1.33 MeV, respectively, by MCNP 5 (Monte Carlo) and WINXCOM program, and compared with those of the lead glass. Results: With attenuation factor K, shielding and mechanical properties taken into consideration bismuth glass containing 50% bismuth oxide might be selected as the right material. Dose rate distributions of water phantom were calculated with 2-cm and 10-cm thick glass, respectively, irradiated by 137 Cs and 60 Co in turn. Conclusion: Results show that the bismuth glass may replace lead glass for radiation shielding with appropriate energy. (authors)

Photonic Bandgap (PBG) Shielding Technology

Science.gov (United States)

Bastin, Gary L.

2007-01-01

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

Modeling of ferrite-based materials for shielding enclosures

International Nuclear Information System (INIS)

Koledintseva, Marina; Drewniak, James; Zhang Yaojiang; Lenn, James; Thoms, Melanie

2009-01-01

An analytical model for a magneto-dielectric composite material is presented based on the Maxwell Garnett rule for a dielectric mixture, and on Bruggeman's effective medium theory for permeability of a ferrite powder embedded in a dielectric. In order to simultaneously treat frequency-dispersive permittivity and permeability of a composite in a full-wave FDTD code, a new algorithm based on discretized auxiliary differential equations has been implemented. In this paper, numerical examples of modeling structures containing different magneto-dielectric mixtures are presented

Shielded regenerative neutron detector

International Nuclear Information System (INIS)

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

1978-01-01

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

Long term testing of materials for tube shielding, stage 2; Laangtidsprovning av tubskyddsmaterial, etapp 2

Energy Technology Data Exchange (ETDEWEB)

Norling, Rikard; Hjoernhede, Anders; Mattsson, Mattias

2012-02-15

Circulating Fluidized Bed (CFB) boilers are commonly used for combustion of biomass and are used to some extent for Waste-to-Energy (WtE) plants as well. The superheaters of the latter are for obvious reasons more prone to suffer from high temperature corrosion caused by the corrosive species in the fuel, mainly chlorides. Frequently the final (hottest) superheater is positioned in the loop seal, where the circulating bed material is returned to the furnace after being separated from the flue gas by a cyclone. The environment in the loop seal is relatively free of chlorides, since these primarily follow the flue gas into the convection pass. Hence, higher steam temperature can be allowed without excessive damage to the final superheater. On the other hand the superheaters, which are located in the convection pass, are more exposed to the corrosive species of the flue gas. Further, they are eroded by particles entrained in the gas flow. Particles and condensing gaseous species are to a large extent deposited on the superheaters, which limits the heat transfer and promotes corrosion. The deposits are regularly removed e.g. by soot blowers. The pressurized steam from soot blowers causes additional erosion damage to that caused by entrained particles. It shall be noted that the actual damage is caused by a combined mechanism of erosion and corrosion denoted erosion-corrosion, which usually results in dramatically accelerated wear. To avoid excessive erosion damage on the superheater tubes the first tube row of each bundle is protected by tube shielding. In its simplest form the shields are made from a steel sheet that has been bent into a semi-circular half-cylinder shell. These shields are attached onto the wind-side of the tubes by hangers. A typical material for tube shielding is the austenitic high temperature resistant stainless steel 253MA. Life of tube shielding depends on numerous factors such as boiler design, superheater location, fuel and operating

FIRE-RESISTANT SHIELDING COATING BASED ON SHUNGITE-CONTAINING PAINT

Directory of Open Access Journals (Sweden)

BELOUSOVA Elena Sergeevna

2013-08-01

Full Text Available Today when specific shielded facilities are designed the construction materials and shields should meet a range of fire safety requirements. A composite coating on the basis of a water-based fire-resistant paint filled with shungite nanopowder can be applied onto walls, floors, ceilings and other surfaces in the shielded areas to reduce electromagnetic radiation and simultaneously to ensure fire safety. Shungit is a mineral with multilayer carbon fullerene globules which diameter is 10–30 nm. Due to the high conductivity shungite is able to weaken electromagnetic radiation. A coating made of schungite-containing paint on a cellulose substrate was subjected to the open flame under the temperature of 1700° C for 3 minutes and 40 seconds. That resulted in the formation of insulating foam layer without mechanical damage of the substrate. The XRD diffraction analysis of the powder obtained in the process of flame influence on the coating showed the formation of the such substances as orthoclase, barite, rutile, etc. Carbon contained in shungit and used as a filler for the fireproof paint wasn’t detected. This fact indicates carbon oxidation as the result of its burning out. The shielding efficiency of the composite coating after open flame exposure was measured for the frequency range 8…12 GHz with the use of the panoramic attenuation meter and voltage standing wave ratio meter YA2R-67-61 with a sweep generator and waveguides. After that the reflection and transmission coefficients were calculated. The results of measurements and calculations showed decrease of the reflection and transmission coefficients due to conductivity decrease and dielectric losses changes of the composite coating provided by silica content increase and carbon percentage decrease.

Optimization of multi-layered metallic shield

International Nuclear Information System (INIS)

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

2011-01-01

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

Development of a new neutron shielding material, TN trademark Resin Vyal for transport/storage casks for radioactive materials

Energy Technology Data Exchange (ETDEWEB)

Abadie, P. [COGEMA Logistics (AREVA Group), Saint-Quentin-en-Yvelines (France)

2004-07-01

TN trademark Resin Vyal, a patent pending composite, is a new neutron shielding material developed by COGEMA LOGISTICS for transport/storage casks of radioactive materials (spent fuel, reprocessed fuel,..). This material is composed of a thermosetting resin (vinylester resin in solution of styrene) and two mineral fillers (alumine hydrate and zinc borate). Its shielding ability for neutron radiation is related to a high hydrogen content (for slowing down neutrons) and a high boron content (for absorbing neutrons). Source of hydrogen is organic matrix (resin) and alumine hydrate; source of boron is zinc borate. Atomic concentrations are equal to 5.10{sup 22} at/cm{sup 3} for hydrogen and 9.10{sup 20} at/cm{sup 3} for boron. Due to the flame retardant character of components, the final material has a good fire resistance (it is auto-extinguishable). Its density is equal to 1,8. The manufacturing process of these material is easy: it consists in mixing the fillers and pouring in-situ (in cask); so, the curing of this composite, which leads to a three-dimensional structure, takes place at ambient temperature. Temperature resistance of this material was evaluated by performing exposition tests of samples at different temperatures (150 C to 170 C). According to tests results, its maximal temperature of use was confirmed equal to 160 C.

Development of a new neutron shielding material, TN trademark Resin Vyal for transport/storage casks for radioactive materials

International Nuclear Information System (INIS)

Abadie, P.

2004-01-01

TN trademark Resin Vyal, a patent pending composite, is a new neutron shielding material developed by COGEMA LOGISTICS for transport/storage casks of radioactive materials (spent fuel, reprocessed fuel,..). This material is composed of a thermosetting resin (vinylester resin in solution of styrene) and two mineral fillers (alumine hydrate and zinc borate). Its shielding ability for neutron radiation is related to a high hydrogen content (for slowing down neutrons) and a high boron content (for absorbing neutrons). Source of hydrogen is organic matrix (resin) and alumine hydrate; source of boron is zinc borate. Atomic concentrations are equal to 5.10 22 at/cm 3 for hydrogen and 9.10 20 at/cm 3 for boron. Due to the flame retardant character of components, the final material has a good fire resistance (it is auto-extinguishable). Its density is equal to 1,8. The manufacturing process of these material is easy: it consists in mixing the fillers and pouring in-situ (in cask); so, the curing of this composite, which leads to a three-dimensional structure, takes place at ambient temperature. Temperature resistance of this material was evaluated by performing exposition tests of samples at different temperatures (150 C to 170 C). According to tests results, its maximal temperature of use was confirmed equal to 160 C

Design of emergency shield

International Nuclear Information System (INIS)

Soliman, S.E.

1993-01-01

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

An Evaluation on Radiation Shielding and Activation Properties of ISOL-bunker Structural Materials for Radiation Safety in RAON Accelerator

Energy Technology Data Exchange (ETDEWEB)

Kim, Do Hyun; Kim, Song Hyun; Woo, Myeong Hyeon; Lee, Jae Yong; Kim, Jong Woo; Shin, Chang Ho [Hanyang University, Seoul (Korea, Republic of); Nam, Shin Woo [Institute for Basic Science, Daejeon (Korea, Republic of)

2015-10-15

RAON heavy ion accelerator has been designed by the Institute for Basic Science (IBS). ISOL is one of RAON facilities to generate and separate rare isotopes. For generating rare isotopes, high intensity proton beam, which has 70 MeV energy, is induced into UCx target. From this reaction, lots of neutrons are concomitantly generated. To meet our design goal, it was required that the structural material of ISOL-bunker should be carefully selected. In this study, to select the structural material which has lower activation property with higher performance for radiation shielding, following aspects were evaluated: (i) residual dose, (ii) radioactive wastes, and (iii) shielding performance in ISOL-bunker. In this study, to effectively design the radiation shielding of the RAON ISOL-bunker, two methods were proposed. No.1 strategy is a method to replace the normal concrete to specific concretes. No.2 strategy is to design dual-layer radiation shields that a specific shielding material is located inner side of the normal concrete. Using the strategies, performance evaluations were evaluated for three aspects, which are residual dose, radioactive waste, and prompt radiation. The results show that the residual radiation can be effectively reduced using B{sub 4}C, borated polyethylene and polyethylene with No.2 strategy. Also, the colemanite concrete and B{sub 4}C shielding give a good ability to reduce the radioactive wastes.

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 ...

Lightweight reduced graphene oxide-Fe3O4 nanoparticle composite in the quest for an excellent electromagnetic interference shielding material

Science.gov (United States)

Singh, Ashwani Kumar; Kumar, Ajit; Kamal Haldar, Krishna; Gupta, Vinay; Singh, Kedar

2018-06-01

This work reports a detailed study of reduced graphene oxide (rGO)-Fe3O4 nanoparticle composite as an excellent electromagnetic (EM) interference shielding material in GHz range. A rGO-Fe3O4 nanoparticle composite was synthesized using a facile, one step, and modified solvothermal method with the reaction of FeCl3, ethylenediamine and graphite oxide powder in the presence of ethylene glycol. Various structural, microstructural and optical characterization tools were used to determine its synthesis and various properties. Dielectric, magnetic and EM shielding parameters were also evaluated to estimate its performance as a shielding material for EM waves. X-ray diffraction patterns have provided information about the structural and crystallographic properties of the as-synthesized material. Scanning electron microscopy micrographs revealed the information regarding the exfoliation of graphite into rGO. Well-dispersed Fe3O4 nanoparticles over the surface of the graphene can easily be seen by employing transmission electron microscopy. For comparison, rGO nanosheets and Fe3O4 nanoparticles have also been synthesized and characterized in a similar fashion. A plot of the dielectric and magnetic characterizations provides some useful information related to various losses and the relaxation process. Shielding effectiveness due to reflection (SER), shielding effectiveness due to absorption (SEA), and total shielding effectiveness (SET) were also plotted against frequency over a broad range (8–12 GHz). A significant change in all parameters (SEA value from 5 dB to 35 dB for Fe3O4 nanoparticles to rGO-Fe3O4 nanoparticle composite) was found. An actual shielding effectiveness (SET) up to 55 dB was found in the rGO-Fe3O4 nanoparticle composite. These graphs give glimpses of how significantly this material shows shielding effectiveness over a broad range of frequency.

Lightweight reduced graphene oxide-Fe3O4 nanoparticle composite in the quest for an excellent electromagnetic interference shielding material.

Science.gov (United States)

Singh, Ashwani Kumar; Kumar, Ajit; Haldar, Krishna Kamal; Gupta, Vinay; Singh, Kedar

2018-06-15

This work reports a detailed study of reduced graphene oxide (rGO)-Fe 3 O 4 nanoparticle composite as an excellent electromagnetic (EM) interference shielding material in GHz range. A rGO-Fe 3 O 4 nanoparticle composite was synthesized using a facile, one step, and modified solvothermal method with the reaction of FeCl 3 , ethylenediamine and graphite oxide powder in the presence of ethylene glycol. Various structural, microstructural and optical characterization tools were used to determine its synthesis and various properties. Dielectric, magnetic and EM shielding parameters were also evaluated to estimate its performance as a shielding material for EM waves. X-ray diffraction patterns have provided information about the structural and crystallographic properties of the as-synthesized material. Scanning electron microscopy micrographs revealed the information regarding the exfoliation of graphite into rGO. Well-dispersed Fe 3 O 4 nanoparticles over the surface of the graphene can easily be seen by employing transmission electron microscopy. For comparison, rGO nanosheets and Fe 3 O 4 nanoparticles have also been synthesized and characterized in a similar fashion. A plot of the dielectric and magnetic characterizations provides some useful information related to various losses and the relaxation process. Shielding effectiveness due to reflection (SE R ), shielding effectiveness due to absorption (SE A ), and total shielding effectiveness (SE T ) were also plotted against frequency over a broad range (8-12 GHz). A significant change in all parameters (SE A value from 5 dB to 35 dB for Fe 3 O 4 nanoparticles to rGO-Fe 3 O 4 nanoparticle composite) was found. An actual shielding effectiveness (SE T ) up to 55 dB was found in the rGO-Fe 3 O 4 nanoparticle composite. These graphs give glimpses of how significantly this material shows shielding effectiveness over a broad range of frequency.

Neutron shielding performance of water-extended polyester

International Nuclear Information System (INIS)

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

2006-01-01

A Monte Carlo study to determine the shielding features to neutrons of water-extended polyester (WEP) was carried out. Materials with low atomic number are predominantly used for neutron shielding because these materials effectively attenuate neutrons, mainly through elastic and inelastic collisions. In addition to neutron attenuation properties, other desirable properties for neutron shielding materials include mechanical strength, stability, low cost, and ease of handling. During the selection of materials to design a neutron shield, prompt gamma production as well as radionuclide induced by neutron activation must be considered. In this investigation the Monte Carlo method (MCNP code) was used to evaluate the performance of a water-extended polyester shield designed for the transportation, storage, and use of a 252 Cf isotopic neutron source, for comparison the calculations were extended to water shielding, the bare source in vacuum and in air. (authors)

Neutron shielding performance of water-extended polyester

Energy Technology Data Exchange (ETDEWEB)

Vega Carrillo, H.R.; Manzanares-Acuna, E.; Hernandez-Davila, V.M. [Zacatecas Univ. Autonoma, Nuclear Studies (Mexico); Vega Carrillo, H.R.; Hernandez-Davila, V.M. [Zacatecas Univ. Autonoma, Electric Engineering Academic Units (Mexico); Gallego, E.; Lorente, A. [Madrid Univ. Politecnica, cNuclear Engineering Department (Mexico)

2006-07-01

A Monte Carlo study to determine the shielding features to neutrons of water-extended polyester (WEP) was carried out. Materials with low atomic number are predominantly used for neutron shielding because these materials effectively attenuate neutrons, mainly through elastic and inelastic collisions. In addition to neutron attenuation properties, other desirable properties for neutron shielding materials include mechanical strength, stability, low cost, and ease of handling. During the selection of materials to design a neutron shield, prompt gamma production as well as radionuclide induced by neutron activation must be considered. In this investigation the Monte Carlo method (MCNP code) was used to evaluate the performance of a water-extended polyester shield designed for the transportation, storage, and use of a {sup 252}Cf isotopic neutron source, for comparison the calculations were extended to water shielding, the bare source in vacuum and in air. (authors)

Effect of particle size of mineral fillers on polymer-matrix composite shielding materials against ionizing electromagnetic radiation

International Nuclear Information System (INIS)

Belgin, E.E.; Aycik, G.A.

2017-01-01

Filler particle size is an important particle that effects radiation attenuation performance of a composite shielding material but the effects of it have not been exploited so far. In this study, two mineral (hematite-ilmenite) with different particle sizes were used as fillers in a polymer-matrix composite and effects of particle size on shielding performance was investigated within a widerange of radiation energy (0-2000 keV). The thermal and structural properties of the composites were also examined. The results showed that as the filler particle size decreased the shielding performance increased. The highest shielding performance reached was 23% with particle sizes being between <7 and <74 µm. (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.

Preparation of a basic data base for shielding design. 3

International Nuclear Information System (INIS)

Takemura, Morio

1998-03-01

With use of a standard groupwise shielding design library JSSTDL produced from the latest evaluated nuclear data library JENDL-3.2, experimental analyses for the In-Vessel Fuel Storage (IVS) Experiment and the Intermediate Heat Exchanger (IHX) Experiment were performed. The results were compared with those obtained by the same analysis method and input data using the JSDJ2 library that had been applied consistently to the JASPER experiment analyses. In general, the results obtained with JSSTDL are higher than those with JSDJ2 as were found in analyses in last two years for the Radial Shield Attenuation Experiment and the Special Materials Experiment and also the Axial Shield Experiment. The calculation-to-experiment ratios of the fast neutron flux just behind deep penetration in sodium were obtained first by these IVS and IHX experimental analyses with the JSSTDL library. However, it was confirmed not to be easy to evaluate the accuracy of sodium cross section because of its dependency on how to model the swelled sodium slabs and tanks. The analyses with JSSTDL library were verified by comparison with other analyses with another library based on JENDL-3.2. Compilation of the input data necessary for future reanalyses of important configurations in JASPER experiments, that were selected at the first stage of this study, were continued and new data were added into the computer disk holding previously accumulated data. (author)

Demonstration test on manufacturing 200 l drum inner shielding material for recycling of reactor operating metal scrap

International Nuclear Information System (INIS)

Umemura, A.; Kimura, K.; Ueno, H.

1993-01-01

Low-level reactor wastes should be safely recycled considering those resource values, the reduction of waste disposal volume and environmental effects. The reasonable recycling system of reactor operating metal scrap has been studied and it was concluded that the 200 liter drum inner shielding material is a very promising product for recycling within the nuclear industry. The drum inner shielding material does not require high quality and so it is expected to be easily manufactured by melting and casting from roughly sorted scrap metals. This means that the economical scrap metal recycling system can be achieved by introducing it. Furthermore its use will ensure safety because of being contained in a drum. In order to realize this recycling system with the drum inner shielding material, the demonstration test program is being conducted. The construction of the test facility, which consists of a melting and refining furnace, a casting apparatus, a machining apparatus etc., was finishing in September, 1992

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

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

International Nuclear Information System (INIS)

1970-07-01

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

SU-F-E-13: Design and Fabrication of Gynacological Brachytherapy Shielding & Non Shielding Applicators Using Indigenously Developed 3D Printing Machine

International Nuclear Information System (INIS)

Shanmugam, S

2016-01-01

Purpose: In this innovative work we have developed Gynecological Brachytherapy shielding & Non Shielding Applicators and compared with the commercially available applicators by using the indigenously developed 3D Printing machine. Methods: We have successfully indigenously developed the 3D printing machine. Which contain the 3 dimensional motion platform, Heater unit, base plate, ect… To fabricate the Gynecological Brachytherapy shielding & non shielding applicators the 3D design were developed in the computer as virtual design. This virtual design is made in a CAD computer file using a 3D modeling program. Separate programme for the shielding & non shielding applicators. We have also provided the extra catheter insert provision in the applicator for the multiple catheter. The DICOM file of the applicator were then converted to stereo Lithography file for the 3D printer. The shielding & Non Shielding Applicators were printed on a indigenously developed 3D printer material. The same dimensions were used to develop the applicators in the acrylic material also for the comparative study. A CT scan was performed to establish an infill-density calibration curve as well as characterize the quality of the print such as uniformity and the infill pattern. To commission the process, basic CT and dose properties of the printing materials were measured in photon beams and compared against water and soft tissue. Applicator were then scanned to confirm the placement of multiple catheter position. Finally dose distributions with rescanned CTs were compared with those computer-generated applicators. Results: The doses measured from the ion Chamber and X-Omat film test were within 2%. The shielded applicator reduce the rectal dose comparatively with the non shielded applicator. Conclusion: As of submission 3 unique cylinders have been designed, printed, and tested dosimetrically. A standardizable workflow for commissioning custom 3D printed applicators was codified and will be

SU-F-E-13: Design and Fabrication of Gynacological Brachytherapy Shielding & Non Shielding Applicators Using Indigenously Developed 3D Printing Machine

Energy Technology Data Exchange (ETDEWEB)

Shanmugam, S

2016-06-15

Purpose: In this innovative work we have developed Gynecological Brachytherapy shielding & Non Shielding Applicators and compared with the commercially available applicators by using the indigenously developed 3D Printing machine. Methods: We have successfully indigenously developed the 3D printing machine. Which contain the 3 dimensional motion platform, Heater unit, base plate, ect… To fabricate the Gynecological Brachytherapy shielding & non shielding applicators the 3D design were developed in the computer as virtual design. This virtual design is made in a CAD computer file using a 3D modeling program. Separate programme for the shielding & non shielding applicators. We have also provided the extra catheter insert provision in the applicator for the multiple catheter. The DICOM file of the applicator were then converted to stereo Lithography file for the 3D printer. The shielding & Non Shielding Applicators were printed on a indigenously developed 3D printer material. The same dimensions were used to develop the applicators in the acrylic material also for the comparative study. A CT scan was performed to establish an infill-density calibration curve as well as characterize the quality of the print such as uniformity and the infill pattern. To commission the process, basic CT and dose properties of the printing materials were measured in photon beams and compared against water and soft tissue. Applicator were then scanned to confirm the placement of multiple catheter position. Finally dose distributions with rescanned CTs were compared with those computer-generated applicators. Results: The doses measured from the ion Chamber and X-Omat film test were within 2%. The shielded applicator reduce the rectal dose comparatively with the non shielded applicator. Conclusion: As of submission 3 unique cylinders have been designed, printed, and tested dosimetrically. A standardizable workflow for commissioning custom 3D printed applicators was codified and will be

Application of MCNP code in shielding calculation of minitype fast reactor

International Nuclear Information System (INIS)

He Keyu; Han Weishi

2008-01-01

An accurate shielding calculation model has been set up for the minitype sodium-cooled fast reactor (MFR) based on MCNP code and particular calculation of its primary shielding parameters has been carried out. The results indicate that the photon and neutron flux density of MFR has rapidly fallen to a low-level. The material for the shielding layer outside of main container is primarily of carbon steel, which can be design as a shielding structure satisfying the safety code. The sodium activation in primary circuit is extremely limited and it is simple to shield from. Both the output of helium in reflector and burn up of boron-10 in control rod are very small. These materials can be used for several cycle lives. (authors)

Validation of SCALE code package on high performance neutron shields

International Nuclear Information System (INIS)

Bace, M.; Jecmenica, R.; Smuc, T.

1999-01-01

The shielding ability and other properties of new high performance neutron shielding materials from the KRAFTON series have been recently published. A comparison of the published experimental and MCNP results for the two materials of the KRAFTON series, with our own calculations has been done. Two control modules of the SCALE-4.4 code system have been used, one of them based on one dimensional radiation transport analysis (SAS1) and other based on the three dimensional Monte Carlo method (SAS3). The comparison of the calculated neutron dose equivalent rates shows a good agreement between experimental and calculated results for the KRAFTON-N2 material.. Our results indicate that the N2-M-N2 sandwich type is approximately 10% inferior as neutron shield to the KRAFTON-N2 material. All values of neutron dose equivalent obtained by SAS1 are approximately 25% lower in comparison with the SAS3 results, which indicates proportions of discrepancies introduced by one-dimensional geometry approximation.(author)

Method of constructing shielding wall

International Nuclear Information System (INIS)

Nagao, Tetsuya.

1990-01-01

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

Neutron shielding performance of water-extended polyester

International Nuclear Information System (INIS)

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

2006-01-01

A Monte Carlo study to determine the shielding features to neutrons of water-extended polyester was carried out. Materials with low atomic number are predominantly used for neutron shielding because these materials effectively attenuate neutrons, mainly through elastic and inelastic collisions. In addition to neutron attenuation properties, other desirable properties for neutron shielding materials include mechanical strength, stability, low cost, and ease of handling. During the selection of materials to design a neutron shield, prompt gamma production as well as radionuclide induced by neutron activation must be considered. In this investigation the Monte Carlo method (M.C.N.P. code) was used to evaluate the performance of a water-extended polyester shield designed for the transportation, storage, and use of a 252 Cf isotopic neutron source, for comparison the calculations were extended to water shielding, the bare source in vacuum and in air. (authors)

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...

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.

Base technology development of new materials for FBR performance innovations

International Nuclear Information System (INIS)

Kano, Shigeki; Koyama, Masahiro; Nomura, Shigeo; Morikawa, Satoru; Ueno, Fumiyoshi

1989-01-01

This paper describes the base technology development of new materials for FBR performance innovations at the Power Reactor and Nuclear Fuel Development Corporation. The contents are as follows: (1) development of sodium and radiation resistant new materials, (2) development of high performance shielding material, (3) development of high performance control material, (4) development of new functional materials for reactor instrumentation. (author)

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 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

A comparative study for different shielding material composition and beam geometry applied to PET facilities: simulated transmission curves

OpenAIRE

Hoff, Gabriela; Costa, Paulo Roberto

2013-01-01

The aim of this work is to simulate transmission data for different beam geometry and material composition in order to evaluate the effect of these parameters on transmission curves. The simulations are focused on outgoing spectra for shielding barriers used in PET facilities. The behavior of the transmission was evaluated as a function of the shielding material composition and thickness using Geant4 Monte Carlo code, version 9.2 p 03.The application was benchmarked for barited mortar and com...

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

Preliminary neutron shielding calculations of the electronics in the EAST BES systems focusing on neutron induced displacement damage

Energy Technology Data Exchange (ETDEWEB)

Náfrádi, Gábor, E-mail: nafradi@reak.bme.hu [Institute of Nuclear Techniques (NTI), Budapest University of Technology and Economics (BME), H-1111 Budapest (Hungary); Kovácsik, Ákos, E-mail: kovacsik.akos@reak.bme.hu [Institute of Nuclear Techniques (NTI), Budapest University of Technology and Economics (BME), H-1111 Budapest (Hungary); Németh, József, E-mail: nemeth.jozsef@wigner.mta.hu [Institute for Particle and Nuclear Physics, Wigner Research Centre for Physics (Wigner RCP), Hungarian Academy of Sciences (HAS), POB 49, 1525 Budapest (Hungary); Pór, Gábor, E-mail: por@reak.bme.hu [Institute of Nuclear Techniques (NTI), Budapest University of Technology and Economics (BME), H-1111 Budapest (Hungary); Zoletnik, Sándor, E-mail: zoletnik.sandor@wigner.mta.hu [Institute for Particle and Nuclear Physics, Wigner Research Centre for Physics (Wigner RCP), Hungarian Academy of Sciences (HAS), POB 49, 1525 Budapest (Hungary)

2016-11-15

Monte Carlo N-Particle (MCNP) calculations were carried out to compare neutron shielding capabilities of three frequently used neutron shielding materials: polyethylene without neutron absorbers, polyethylene with boron absorbers and polyethylene with lithium absorbers, according to Non Ionizing Energy Loss (NIEL). The results of 1D shielding calculations showed that simple neutron moderating materials can provide sufficient and cheap shielding against 2.45 MeV and 14.1 MeV fusion neutrons, in terms of 1 MeV neutron equivalent flux, in silicon targets, which is the most commonly used material of electronic components. Based on these results a new shielding concept is proposed which can be taken into consideration where the reduction of displacement damage is the main goal and the free space available for shielding is limited. Based on this shielding concept detailed 3D calculations were carried out to describe the properties of the neutron shielding of the Beam Emission Spectroscopy (BES) system installed at the EAST tokamak.

A shielding chamber for the Rossendorf whole body counter

International Nuclear Information System (INIS)

Beutmann, A.; Ebert, S.; Kaden, M.; Loehnert, D.; Doerfel, H.R.; Schreiber, W.; Helbig, S.

2016-01-01

In connection with the relocation of the incorporation measurement point operated by the VKTA, a new shielding chamber was designed. The development of the new shielding chamber will be shown based on the design study by IDEA System and the inquiries for material availability, procurement of material and assembly technology up to fabrication, assembly and completion of the chamber. The accompanying background measurements through In-situ gamma spectrometry and first experiences with incorporation measurements at the new In-Vivo measurement facility are shown.

Passive magnetic shielding in MRI-Linac systems

Science.gov (United States)

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

2018-04-01

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

Infinite slab-shield dose calculations

International Nuclear Information System (INIS)

Russell, G.J.

1989-01-01

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

Effectiveness of shield materials in the design of the PFBR irradiated fuel subassembly shipping cask

International Nuclear Information System (INIS)

Radhakrishnan, G.

2003-01-01

Fuel subassemblies are irradiated inside the reactor core till they achieve the required burn up and after that they are cooled to permissible decay power level in in-vessel and ex-vessel storage places. Subsequently they are transported to reprocessing plants by means of shipping casks. Shield for the shipping cask has to be designed such a way that it has to comply with the ICRP recommended dose levels of less than 2 mSv/h on contact at the outer surface of the cask and less than 100 mSv/h at 1 m distance from the outer surface of the cask. In this paper, shield design of a typical PFBR irradiated fuel subassembly, which can transport three subassemblies at a time, is narrated. Considering the neutron and fission product and induced gamma rays emitted by typical PFBR irradiated core central subassembly subjected to a maximum burn up, as the source term shield design optimizations have been done. One-dimensional discrete ordinates transport theory computer code ANISN and point kernel computer code QAD-CGGP have been used in complement to carry out the shield design optimizations. Cast-iron, carbon steel, stainless steel 304 and lead and permali have been considered for shield materials. Shield requirements on top, bottom and along the axial height of the shipping cask have also been estimated. (author)

A comparative study for different shielding material composition and beam geometry applied to PET facilities: simulated transmission curves

Energy Technology Data Exchange (ETDEWEB)

Hoff, Gabriela [Pontificia Univ. Catolica do Rio Grande do Sul (PUCRS), Porto Alegre, RS (Brazil). Grupo de Experimentacao e Simulacao Computacional em Fisica Medica; Costa, Paulo Roberto, E-mail: pcosta@if.usp.br [Universidade de Sao Paulo (IF/USP), SP (Brazil). Dept. de Fisica Nuclear. Lab. de Dosimetria das Radiacoes e Fisica Medica

2013-03-15

The aim of this work is to simulate transmission data for different beam geometry and material composition in order to evaluate the effect of these parameters on transmission curves. The simulations are focused on outgoing spectra for shielding barriers used in PET facilities. The behavior of the transmission was evaluated as a function of the shielding material composition and thickness using Geant4 Monte Carlo code, version 9.2 p 03.The application was benchmarked for barited mortar and compared to The American Association of Physicists in Medicine (AAPM) data for lead. Their influence on the transmission curves as well the study of the influence of the shielding material composition and beam geometry on the outgoing spectra were performed. Characteristics of transmitted spectra, such as shape, average energy and Half-Value Layer (HVL), were also evaluated. The Geant4 toolkit benchmark for the energy resulting from the positron annihilation phenomena and its application in transmission curves description shown good agreement between data published by American Association on Physicists in Medicine task group 108 and experimental data published by Brazil. The transmission properties for different material compositions were also studied and have shown low dependency with the considered thicknesses. The broad and narrow beams configuration presented significant differences on the result. The fitting parameter for determining the transmission curves equations, according to Archer model is presented for different material. As conclusion were defined that beam geometry has significant influence and the composition has low influence on transmission curves for shielding design for the range of energy applied to PET. (author)

A comparative study for different shielding material composition and beam geometry applied to PET facilities: simulated transmission curves

International Nuclear Information System (INIS)

Hoff, Gabriela; Costa, Paulo Roberto

2013-01-01

The aim of this work is to simulate transmission data for different beam geometry and material composition in order to evaluate the effect of these parameters on transmission curves. The simulations are focused on outgoing spectra for shielding barriers used in PET facilities. The behavior of the transmission was evaluated as a function of the shielding material composition and thickness using Geant4 Monte Carlo code, version 9.2 p 03.The application was benchmarked for barited mortar and compared to The American Association of Physicists in Medicine (AAPM) data for lead. Their influence on the transmission curves as well the study of the influence of the shielding material composition and beam geometry on the outgoing spectra were performed. Characteristics of transmitted spectra, such as shape, average energy and Half-Value Layer (HVL), were also evaluated. The Geant4 toolkit benchmark for the energy resulting from the positron annihilation phenomena and its application in transmission curves description shown good agreement between data published by American Association on Physicists in Medicine task group 108 and experimental data published by Brazil. The transmission properties for different material compositions were also studied and have shown low dependency with the considered thicknesses. The broad and narrow beams configuration presented significant differences on the result. The fitting parameter for determining the transmission curves equations, according to Archer model is presented for different material. As conclusion were defined that beam geometry has significant influence and the composition has low influence on transmission curves for shielding design for the range of energy applied to PET. (author)

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)

Potential of Nanocellulose Composite for Electromagnetic Shielding

Directory of Open Access Journals (Sweden)

Nabila Yah Nurul Fatihah

2017-01-01

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

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

Overview of first wall/blanket/shield technology

International Nuclear Information System (INIS)

Nygren, R.E.

1983-04-01

This brief overview of first wall, blanket, and shield technology focuses first on changes and trends in important design issues from the 1970's to the 1980's, then on current perceptions of critical issues in first wall, blanket, and shield design and related technology. The emphasis is on base technology rather than either systems engineering or materials development, on the two primary confinement systems, tokamaks and mirrors, and on production of electricity as the primary goal for development

Neutron shielding for a 252 Cf source

International Nuclear Information System (INIS)

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

2006-01-01

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

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

International Nuclear Information System (INIS)

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

2008-01-01

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

MEANS FOR SHIELDING AND COOLING REACTORS

Science.gov (United States)

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

1959-02-10

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

Innovative analytical competence. Optimization of shielding components and lifetime activation calculations

Energy Technology Data Exchange (ETDEWEB)

Boehlke, Steffen; Wortmann, Birgit; Aguilar, Arturo Lizon [STEAG Energy Services GmbH, Essen (Germany)

2014-08-15

Shielding and activation calculations always require a high level of engineering competence and powerful hard- and software tools. With the application of current methods often certain limits were reached in the past. The engineering work for optimization efforts regarding complex components with high shielding requirements exceeded the savings in material. With regard to activation the challenges in size of the geometric model and considered operation time rises constantly and pushes computing time beyond reasonable time frames. These challenges require the application of new and faster methodologies. The application of new and innovative methods is presented for a shielding optimization project to decrease the radiation level, to keep the dose rate limits, and to reduce the amount of used shielding material. In a second case a prediction of the activated materials with it's dose distribution in the surrounding area and classification of waste quantities in the structural materials of a nuclear reactor is presented. For the shielding project the preliminary design CAD model was imported into the software tool, several iterations were run and a significantly reduced radiation exposure together with a significant reduction in shieling material were achieved. For the activation calculations it could be demonstrated that it is possible to determine the activation, waste quantities and dose distribution for the structural materials of a nuclear reactor based on lifetime operational data within reasonable time frames.

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.

X-Ray Micro-Tomography Applied to Nasa's Materials Research: Heat Shields, Parachutes and Asteroids

Science.gov (United States)

Panerai, Francesco; Borner, Arnaud; Ferguson, Joseph C.; Mansour, Nagi N.; Stern, Eric C.; Barnard, Harold S.; Macdowell, Alastair A.; Parkinson, Dilworth Y.

2017-01-01

X-ray micro-tomography is used to support the research on materials carried out at NASA Ames Research Center. The technique is applied to a variety of applications, including the ability to characterize heat shield materials for planetary entry, to study the Earth- impacting asteroids, and to improve broadcloths of spacecraft parachutes. From micro-tomography images, relevant morphological and transport properties are determined and validated against experimental data.

Nuclear reactor shield including magnesium oxide

International Nuclear Information System (INIS)

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

1981-01-01

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

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.

Shielding effectiveness of superconductive particles in plastics

International Nuclear Information System (INIS)

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

1988-09-01

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

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

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.

Program for photon shielding calculations. Examination of approximations on irradiation geometries

International Nuclear Information System (INIS)

Isozumi, Yasuhito; Ishizuka, Fumihiko; Miyatake, Hideo; Kato, Takahisa; Tosaki, Mitsuo

2004-01-01

Penetration factors and related numerical data in 'Manual of Practical Shield Calculation of Radiation Facilities (2000)', which correspond to the irradiation geometries of point isotropic source in infinite thick material (PI), point isotropic source in finite thick material (PF) and vertical incident to finite thick material (VF), have been carefully examined. The shield calculation based on the PI geometry is usually performed with effective dose penetration factors of radioisotopes given in the 'manual'. The present work cleary shows that such a calculation may lead to an overestimate more than twice larger, especially for thick shield of concrete and water. Employing the numerical data in the 'manual', we have fabricated a simple computer program for the estimation of penetration factors and effective doses of radioisotopes in the different irradiation geometries, i.e., PI, PF and VF. The program is also available to calculate the effective dose from a set of radioisotopes in the different positions, which is necessary for the γ-ray shielding of radioisotope facilities. (author)

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

International Nuclear Information System (INIS)

Whitson, W.R.

1977-09-01

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

Shielding calculation for bremsstrahlung from β-emitters

International Nuclear Information System (INIS)

Ichimiya, Tsutomu

1990-01-01

Accompanying the revision of radiation injury prevention law, the shielding calculation method for photon corresponding to the dose equivalent was shown. However, regarding the electron from β decay nuclide and bremsstrahlung caused by shielding material, the shielding calculation method corresponding to the 1 cm dose equivalent has not been reported, hence, in this report, the spectrum of β-ray is calculated and the 1 cm dose equivalent transmission rate of the bremsstrahlung was calculated for three kinds of shielding materials (iron, lead, concrete). As the result of consideration, it is sufficient to think about the bremsstrahlung due to negative electron emission accompanying β-decay. In β-decay, electrons which constitute the continuous spectrum with maximum energy are emitted. The shape of the spectrum differs with nuclides. The maximum energy of β-ray of generally used nuclides is mostly below 3MeV and, besides, the electron ray itself is easily shielded, while the strength of bremsstrahlung depends on the atomic number of shielding materials and its generating mechanism is complicated. In this report, the actual shielding calculation method for bremsstrahlung is shown with regard to the most frequently used β-decay nuclides. (M.T.)

MOSFET Dosimetry for Evaluation of Gonad Shielding during Radiotherapy

International Nuclear Information System (INIS)

Kim, Hwi Young; Choi, Yun Seok; Park, So Yeon; Park, Yang Kyun; Ye, Sung Joon

2011-01-01

In order to confirm feasibility of MOSFET modality in use of in vivo dosimetry, evaluation of gonad shielding in order to minimize gonadal dose of patients undergoing radiotherapy by using MOSFET modality was performed. Gonadal dose of patients undergoing radiotherapy for rectal cancer in the department of radiation oncology of Seoul National University Hospital since 2009 was measured. 6 MV and 15 MV photon beams emitted from Varian 21EX LINAC were used for radiotherapy. In order to minimize exposed dose caused by the scattered ray not only from collimator of LINAC but also from treatment region inside radiation field, we used box.shaped lead shielding material. The shielding material was made of the lead block and consists of 7.5 cm x 9.5 cm x 5.5 cm sized case and 9 cm x 9.5 cm x 1 cm sized cover. Dosimetry for evaluation of gonad shielding was done with MOSFET modality. By protecting with gonad shielding material, average gonadal dose of patients was decreased by 23.07% compared with reference dose outside of the shielding material. Average delivered gonadal dose inside the shielding material was 0.01 Gy. By the result of MOSFET dosimetry, we verified that gonadal dose was decreased by using gonad shielding material. In compare with TLD dosimetry, we could measure the exposed dose easily and precisely with MOSFET modality

MOSFET Dosimetry for Evaluation of Gonad Shielding during Radiotherapy

Energy Technology Data Exchange (ETDEWEB)

Kim, Hwi Young; Choi, Yun Seok; Park, So Yeon; Park, Yang Kyun [Seoul National University College of Medicine, Seoul (Korea, Republic of); Ye, Sung Joon [Seoul National University, Seoul (Korea, Republic of)

2011-03-15

In order to confirm feasibility of MOSFET modality in use of in vivo dosimetry, evaluation of gonad shielding in order to minimize gonadal dose of patients undergoing radiotherapy by using MOSFET modality was performed. Gonadal dose of patients undergoing radiotherapy for rectal cancer in the department of radiation oncology of Seoul National University Hospital since 2009 was measured. 6 MV and 15 MV photon beams emitted from Varian 21EX LINAC were used for radiotherapy. In order to minimize exposed dose caused by the scattered ray not only from collimator of LINAC but also from treatment region inside radiation field, we used box.shaped lead shielding material. The shielding material was made of the lead block and consists of 7.5 cm x 9.5 cm x 5.5 cm sized case and 9 cm x 9.5 cm x 1 cm sized cover. Dosimetry for evaluation of gonad shielding was done with MOSFET modality. By protecting with gonad shielding material, average gonadal dose of patients was decreased by 23.07% compared with reference dose outside of the shielding material. Average delivered gonadal dose inside the shielding material was 0.01 Gy. By the result of MOSFET dosimetry, we verified that gonadal dose was decreased by using gonad shielding material. In compare with TLD dosimetry, we could measure the exposed dose easily and precisely with MOSFET modality.

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.

Cloud immersion building shielding factors for US residential structures

International Nuclear Information System (INIS)

Dickson, E D; Hamby, D M

2014-01-01

This paper presents validated building shielding factors designed for contemporary US housing-stock under an idealized, yet realistic, exposure scenario within a semi-infinite cloud of radioactive material. The building shielding factors are intended for use in emergency planning and level three probabilistic risk assessments for a variety of postulated radiological events in which a realistic assessment is necessary to better understand the potential risks for accident mitigation and emergency response planning. Factors are calculated from detailed computational housing-units models using the general-purpose Monte Carlo N-Particle computational code, MCNP5, and are benchmarked from a series of narrow- and broad-beam measurements analyzing the shielding effectiveness of ten common general-purpose construction materials and ten shielding models representing the primary weather barriers (walls and roofs) of likely US housing-stock. Each model was designed to scale based on common residential construction practices and include, to the extent practical, all structurally significant components important for shielding against ionizing radiation. Calculations were performed for floor-specific locations as well as for computing a weighted-average representative building shielding factor for single- and multi-story detached homes, both with and without basement, as well for single-wide manufactured housing-units. (paper)

Support vector regression model for the estimation of γ-ray buildup factors for multi-layer shields

International Nuclear Information System (INIS)

Trontl, Kresimir; Smuc, Tomislav; Pevec, Dubravko

2007-01-01

The accuracy of the point-kernel method, which is a widely used practical tool for γ-ray shielding calculations, strongly depends on the quality and accuracy of buildup factors used in the calculations. Although, buildup factors for single-layer shields comprised of a single material are well known, calculation of buildup factors for stratified shields, each layer comprised of different material or a combination of materials, represent a complex physical problem. Recently, a new compact mathematical model for multi-layer shield buildup factor representation has been suggested for embedding into point-kernel codes thus replacing traditionally generated complex mathematical expressions. The new regression model is based on support vector machines learning technique, which is an extension of Statistical Learning Theory. The paper gives complete description of the novel methodology with results pertaining to realistic engineering multi-layer shielding geometries. The results based on support vector regression machine learning confirm that this approach provides a framework for general, accurate and computationally acceptable multi-layer buildup factor model

PC based temporary shielding administrative procedure (TSAP)

International Nuclear Information System (INIS)

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

1995-01-01

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

PC based temporary shielding administrative procedure (TSAP)

Energy Technology Data Exchange (ETDEWEB)

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

1995-03-01

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

Background components of Ge(Li) and GeHP-detectors in the passive shield

International Nuclear Information System (INIS)

Buraeva, E.A.; Davydov, M.G.; Zorina, L.V.; Stasov, V.V.

2007-01-01

The gamma-spectrometer Ge(Li)- and the extra pure Ge-detector background components in a specially designed passive shield were subjected to investigation in the land-based laboratory in 1996-2006. The measurement time period varied from 45 up to 240 hours. The detector background is caused by the radionuclides in the shield material, in the shield cells and in the detector materials. The prominence was given to the study of the revealed time dependence of 222 Rn daughter product background including '2 10 Pb 46.5 keV peak [ru

Topogram-based tube current modulation of head computed tomography for optimizing image quality while protecting the eye lens with shielding.

Science.gov (United States)

Lin, Ming-Fang; Chen, Chia-Yuen; Lee, Yuan-Hao; Li, Chia-Wei; Gerweck, Leo E; Wang, Hao; Chan, Wing P

2018-01-01

Background Multiple rounds of head computed tomography (CT) scans increase the risk of radiation-induced lens opacification. Purpose To investigate the effects of CT eye shielding and topogram-based tube current modulation (TCM) on the radiation dose received by the lens and the image quality of nasal and periorbital imaging. Material and Methods An anthropomorphic phantom was CT-scanned using either automatic tube current modulation or a fixed tube current. The lens radiation dose was estimated using cropped Gafchromic films irradiated with or without a shield over the orbit. Image quality, assessed using regions of interest drawn on the bilateral extraorbital areas and the nasal bone with a water-based marker, was evaluated using both a signal-to-noise ratio (SNR) and contrast-noise ratio (CNR). Two CT specialists independently assessed image artifacts using a three-point Likert scale. Results The estimated radiation dose received by the lens was significantly lower when barium sulfate or bismuth-antimony shields were used in conjunction with a fixed tube current (22.0% and 35.6% reduction, respectively). Topogram-based TCM mitigated the beam hardening-associated artifacts of bismuth-antimony and barium sulfate shields. This increased the SNR by 21.6% in the extraorbital region and the CNR by 7.2% between the nasal bones and extraorbital regions. The combination of topogram-based TCM and barium sulfate or bismuth-antimony shields reduced lens doses by 12.2% and 27.2%, respectively. Conclusion Image artifacts induced by the bismuth-antimony shield at a fixed tube current for lenticular radioprotection were significantly reduced by topogram-based TCM, which increased the SNR of the anthropomorphic nasal bones and periorbital tissues.

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.

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

Development of a Waterproof Crack-Based Stretchable Strain Sensor Based on PDMS Shielding

Directory of Open Access Journals (Sweden)

Seong Kyung Hong

2018-04-01

Full Text Available This paper details the design of a poly(dimethylsiloxane (PDMS-shielded waterproof crack-based stretchable strain sensor, in which the electrical characteristics and sensing performance are not influenced by changes in humidity. This results in a higher number of potential applications for the sensor. A previously developed omni-purpose stretchable strain (OPSS sensor was used as the basis for this work, which utilizes a metal cracking structure and provides a wide sensing range and high sensitivity. Changes in the conductivity of the OPSS sensor, based on humidity conditions, were investigated along with the potential possibility of using the design as a humidity sensor. However, to prevent conductivity variation, which can decrease the reliability and sensing ability of the OPSS sensor, PDMS was utilized as a shielding layer over the OPSS sensor. The PDMS-shielded OPSS sensor showed approximately the same electrical characteristics as previous designs, including in a high humidity environment, while maintaining its strain sensing capabilities. The developed sensor shows promise for use under high humidity conditions and in underwater applications. Therefore, considering its unique features and reliable sensing performance, the developed PDMS-shielded waterproof OPSS sensor has potential utility in a wide range of applications, such as motion monitoring, medical robotics and wearable healthcare devices.

Development of a Waterproof Crack-Based Stretchable Strain Sensor Based on PDMS Shielding.

Science.gov (United States)

Hong, Seong Kyung; Yang, Seongjin; Cho, Seong J; Jeon, Hyungkook; Lim, Geunbae

2018-04-12

This paper details the design of a poly(dimethylsiloxane) (PDMS)-shielded waterproof crack-based stretchable strain sensor, in which the electrical characteristics and sensing performance are not influenced by changes in humidity. This results in a higher number of potential applications for the sensor. A previously developed omni-purpose stretchable strain (OPSS) sensor was used as the basis for this work, which utilizes a metal cracking structure and provides a wide sensing range and high sensitivity. Changes in the conductivity of the OPSS sensor, based on humidity conditions, were investigated along with the potential possibility of using the design as a humidity sensor. However, to prevent conductivity variation, which can decrease the reliability and sensing ability of the OPSS sensor, PDMS was utilized as a shielding layer over the OPSS sensor. The PDMS-shielded OPSS sensor showed approximately the same electrical characteristics as previous designs, including in a high humidity environment, while maintaining its strain sensing capabilities. The developed sensor shows promise for use under high humidity conditions and in underwater applications. Therefore, considering its unique features and reliable sensing performance, the developed PDMS-shielded waterproof OPSS sensor has potential utility in a wide range of applications, such as motion monitoring, medical robotics and wearable healthcare devices.

Thermal behavior of neutron shielding material, NS-4-FR, under long term storage conditions

International Nuclear Information System (INIS)

Yamada, N.; O-iwa, A.; Asano, R.; Horita, R.; Kusunoki, K.

2004-01-01

NS-4-FR, Epoxy-Resin, has been widely used as a neutron shielding material for casks. It is recognized that the resin will degrade during storage and loose weight under high temperature conditions. Most of the examinations for the resin degrading behavior were conducted with rather small bare resin specimens. However, the actual quantity of neutron shielding is quite large and is covered by the cask body. To confirm the degrading behavior of the resin under the long-term storage conditions, we performed the test on the specimen with the same cross-section as the actual design, Hitz B69. The resin test vessels were made out of stainless steel and equipped with flange

A New Algorithm for Radioisotope Identification of Shielded and Masked SNM/RDD Materials

International Nuclear Information System (INIS)

Jeffcoat, R.

2012-01-01

Detection and identification of shielded and masked nuclear materials is crucial to national security, but vast borders and high volumes of traffic impose stringent requirements for practical detection systems. Such tools must be be mobile, and hence low power, provide a low false alarm rate, and be sufficiently robust to be operable by non-technical personnel. Currently fielded systems have not achieved all of these requirements simultaneously. Transport modeling such as that done in GADRAS is able to predict observed spectra to a high degree of fidelity; our research is focusing on a radionuclide identification algorithm that inverts this modeling within the constraints imposed by a handheld device. Key components of this work include incorporation of uncertainty as a function of both the background radiation estimate and the hypothesized sources, dimensionality reduction, and nonnegative matrix factorization. We have partially evaluated performance of our algorithm on a third-party data collection made with two different sodium iodide detection devices. Initial results indicate, with caveats, that our algorithm performs as good as or better than the on-board identification algorithms. The system developed was based on a probabilistic approach with an improved approach to variance modeling relative to past work. This system was chosen based on technical innovation and system performance over algorithms developed at two competing research institutions. One key outcome of this probabilistic approach was the development of an intuitive measure of confidence which was indeed useful enough that a classification algorithm was developed based around alarming on high confidence targets. This paper will present and discuss results of this novel approach to accurately identifying shielded or masked radioisotopes with radiation detection systems.

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

Science.gov (United States)

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

2016-09-08

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

Requirements for timber and cadmium used in shielding for fissile material transport packaging

International Nuclear Information System (INIS)

1982-02-01

This Code of Practice has been prepared as a guide for designers who require packaging for fissile materials. It should be noted that this document covers design requirements only and it is not a manufacturing specification which can be quoted on a manufacturing contract without qualification. Compliance with the regulations regarding the safe transport of fissile materials may be achieved by the provision of an effective shield embodying:- (a) a moderating material -usually one rich in hydrogen, such as wood - in order to thermalise incoming neutrons, and (b) a material - such as cadmium - with a large absorption cross-section for thermal neutrons, located between the moderator and the fissile material, in order to capture the incoming neutrons. This Code describes the requirements in two sections, one for each of these materials. (author)

Shielding benchmark problems

International Nuclear Information System (INIS)

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

1978-09-01

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

Neutron shielding for a {sup 252} Cf source

Energy Technology Data Exchange (ETDEWEB)

Vega C, H.R.; Manzanares A, E.; Hernandez D, V.M. [Unidades Academicas de Estudios Nucleares e Ingenieria Electrica, Universidad Autonoma de Zacatecas, C. Cipres 10, Fracc. La Penuela, 98068 Zacatecas (Mexico); Eduardo Gallego, Alfredo Lorente [Depto. de Ingenieria Nuclear, ETS Ingenieros Industriales, Universidad Politecnica de Madrid, C. Jose Gutierrez Abascal 2, 28006 Madrid (Spain)]. e-mail: fermineutron@yahoo.com

2006-07-01

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

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

Shield wall evaluation of hot cell facility for advanced spent fuel conditioning process

International Nuclear Information System (INIS)

Cho, I. J.; Kuk, D. H.; Ko, J. H.; Jung, W. M.; Yoo, G. S.; Lee, E. P.; Park, S. W.

2002-01-01

The future hot cell is located in the Irradiated Material Experiment Facility (IMEF) at the Korea Atomic Energy Research Institute (KAERI). It is β-γ type hot cell that was constructed on the base floor in IMEF building for irradiated material testing. And this hot cell will be used for carrying out the Advanced spent fuel Conditioning Process (ACP). The radiation shielding capability of hot cell should be sufficient to meet the radiation dose requirements in the related regulations. Because the radioactive sources of ACP are expected to be higher than radioactive sources of IMEF design criteria, the future hot cell in current status is unsatisfactory to hot test of ACP. So the shielding analysis of the future hot cell is performed to evaluate shielding ability of concrete shield wall. The shielding analysis included (a) identification of ACP source term; (b) photon source spectrum; (c) shielding analysis by QADS and MCNP-4C; and (d) enhancement of concrete shield wall. In this research, dose rates are obtained according to ACP source, geometry and hot cell shield wall thickness. And the evaluation and reinforcement thickness of the shield wall about future hot cell are concluded

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

Experimental simulation and numerical modeling of vapor shield formation and divertor material erosion for ITER typical plasma disruptions

International Nuclear Information System (INIS)

Wuerz, H.; Arkhipov, N.I.; Bakhtin, V.P.; Konkashbaev, I.; Landman, I.; Safronov, V.M.; Toporkov, D.A.; Zhitlukhin, A.M.

1995-01-01

The high divertor heat load during a tokamak plasma disruption results in sudden evaporation of a thin layer of divertor plate material, which acts as vapor shield and protects the target from further excessive evaporation. Formation and effectiveness of the vapor shield are theoretically modeled and are experimentally analyzed at the 2MK-200 facility under conditions simulating the thermal quench phase of ITER tokamak plasma disruptions. ((orig.))

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.

Development of a shield based on Monte-Carlo studies for the COBRA experiment

Energy Technology Data Exchange (ETDEWEB)

Heidrich, Nadine [Institut fuer Experimentalphysik, 22761 Hamburg (Germany); Collaboration: COBRA-Collaboration

2013-07-01

COBRA is a next-generation experiment searching for neutrinoless double beta decay using CdZnTe semiconductor detectors. The main focus is on {sup 116}Cd, with a decay energy of 2813.5 keV well above the highest naturally occurring gamma lines. The concept for a large scale set-up consists of an array of CdZnTe detectors with a total mass of 420 kg enriched in {sup 116}Cd up to 90 %. With a background rate in the order of 10{sup -3} counts/keV/kg/year, the experiment would be sensitive to a half-life larger than 10{sup 26} years, corresponding to a Majorana mass term m{sub ββ} smaller than 50 meV. To achieve the background level, an appropriate shield is necessary. The shield is developed based on Monte-Carlo simulations. For that, different materials and configurations are tested. In the talk the current status of the Monte-Carlo survey is presented and discussed.

Measurement of Charged Particle Interactions in Spacecraft and Planetary Habitat Shielding Materials

Science.gov (United States)

Zeitlin, Cary J.; Heilbronn, Lawrence H.; Miller, Jack; Wilson, John W.; Singleterry, Robert C., Jr.

2003-01-01

Accurate models of health risks to astronauts on long-duration missions outside the geomagnetosphere will require a full understanding of the radiation environment inside a spacecraft or planetary habitat. This in turn requires detailed knowledge of the flux of incident particles and their propagation through matter, including the nuclear interactions of heavy ions that are a part of the Galactic Cosmic Radiation (GCR). The most important ions are likely to be iron, silicon, oxygen, and carbon. Transport of heavy ions through complex shielding materials including self-shielding of tissue modifies the radiation field at points of interest (e.g., at the blood-forming organs). The incident flux is changed by two types of interactions: (1) ionization energy loss, which results in reduced particle velocity and higher LET (Linear Energy Transfer); and (2) nuclear interactions that fragment the incident nuclei into less massive ions. Ionization energy loss is well understood, nuclear interactions less so. Thus studies of nuclear fragmentation at GCR-like energies are needed to fill the large gaps that currently exist in the database. These can be done at only a few accelerator facilities where appropriate beams are available. Here we report results from experiments performed at the Brookhaven National Laboratory s Alternating Gradient Synchrotron (AGS) and the Heavy Ion Medical Accelerator in Chiba, Japan (HIMAC). Recent efforts have focused on extracting charge-changing and fragment production cross sections from silicon beams at 400, 600, and 1200 MeV/nucleon. Some energy dependence is observed in the fragment production cross sections, and as in other data sets the production of fragments with even charge numbers is enhanced relative to those with odd charge numbers. These data are compared to the NASA-LaRC model NUCFRG2. The charge-changing cross section data are compared to recent calculations using an improved model due to Tripathi, which accurately predicts the

Evaluation of radiation shielding performance in sea transport of radioactive material by using simple calculation method

International Nuclear Information System (INIS)

Odano, N.; Ohnishi, S.; Sawamura, H.; Tanaka, Y.; Nishimura, K.

2004-01-01

A modified code system based on the point kernel method was developed to use in evaluation of shielding performance for maritime transport of radioactive material. For evaluation of shielding performance accurately in the case of accident, it is required to preciously model the structure of transport casks and shipping vessel, and source term. To achieve accurate modelling of the geometry and source term condition, we aimed to develop the code system by using equivalent information regarding structure and source term used in the Monte Carlo calculation code, MCNP. Therefore, adding an option to use point kernel method to the existing Monte Carlo code, MCNP4C, the code system was developed. To verify the developed code system, dose rate distribution in an exclusive shipping vessel to transport the low level radioactive wastes were calculated by the developed code and the calculated results were compared with measurements and Monte Carlo calculations. It was confirmed that the developed simple calculation method can obtain calculation results very quickly with enough accuracy comparing with the Monte Carlo calculation code MCNP4C

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)

Verification of using SABINE-3.1 code for calculations of radioactive inventory in reactor shield

International Nuclear Information System (INIS)

Moukhamadeev, R.; Suvorov, A.

2000-01-01

This report presents the results of calculations of radioactive inventory and doses of activation radiation for the International Benchmark Calculations of Radioactive Inventory for Fission Reactor Decommissioning, IAEA, and measurements of activation doses in shield of WWER-440 (Armenian NPP), using one-dimension modified code SABINE-3.1. For decommissioning of NPP it is very important to evaluate in correct manner radioactive inventory in reactor construction and shield materials. One-dimension code SABINE-3.1 (removing-diffusion method for neutron calculation) was modified to perform calculation of radioactive inventory in reactor shield materials and dose from activation photons behind them. These calculations are carried out on the base of nuclear constant system ABBN-78 and new library of activation data for a number of long-lived isotopes, prepared by authors on the base of [9], which present at shield materials as microimpurities and manage radiation situation under the decay more than 1 year. (Authors)

General Corrosion and Localized Corrosion of the Drip Shield

Energy Technology Data Exchange (ETDEWEB)

F. Hua

2004-09-16

The repository design includes a drip shield (BSC 2004 [DIRS 168489]) that provides protection for the waste package both as a barrier to seepage water contact and a physical barrier to potential rockfall. The purpose of the process-level models developed in this report is to model dry oxidation, general corrosion, and localized corrosion of the drip shield plate material, which is made of Ti Grade 7. This document is prepared according to ''Technical Work Plan For: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package'' (BSC 2004 [DIRS 171583]). The models developed in this report are used by the waste package degradation analyses for TSPA-LA and serve as a basis to determine the performance of the drip shield. The drip shield may suffer from other forms of failure such as the hydrogen induced cracking (HIC) or stress corrosion cracking (SCC), or both. Stress corrosion cracking of the drip shield material is discussed in ''Stress Corrosion Cracking of the Drip Shield, the Waste Package Outer Barrier, and the Stainless Steel Structural Material'' (BSC 2004 [DIRS 169985]). Hydrogen induced cracking of the drip shield material is discussed in ''Hydrogen Induced Cracking of Drip Shield'' (BSC 2004 [DIRS 169847]).

General Corrosion and Localized Corrosion of the Drip Shield

International Nuclear Information System (INIS)

F. Hua

2004-01-01

The repository design includes a drip shield (BSC 2004 [DIRS 168489]) that provides protection for the waste package both as a barrier to seepage water contact and a physical barrier to potential rockfall. The purpose of the process-level models developed in this report is to model dry oxidation, general corrosion, and localized corrosion of the drip shield plate material, which is made of Ti Grade 7. This document is prepared according to ''Technical Work Plan For: Regulatory Integration Modeling and Analysis of the Waste Form and Waste Package'' (BSC 2004 [DIRS 171583]). The models developed in this report are used by the waste package degradation analyses for TSPA-LA and serve as a basis to determine the performance of the drip shield. The drip shield may suffer from other forms of failure such as the hydrogen induced cracking (HIC) or stress corrosion cracking (SCC), or both. Stress corrosion cracking of the drip shield material is discussed in ''Stress Corrosion Cracking of the Drip Shield, the Waste Package Outer Barrier, and the Stainless Steel Structural Material'' (BSC 2004 [DIRS 169985]). Hydrogen induced cracking of the drip shield material is discussed in ''Hydrogen Induced Cracking of Drip Shield'' (BSC 2004 [DIRS 169847])

Preparation of a basic data base for shielding design. 4

International Nuclear Information System (INIS)

Nakao, Makoto; Takemura, Morio

1999-03-01

With use of a standard groupwise shielding design library JSSTDL produced from the latest evaluated nuclear data library JENDL-3.2, experimental analyses for the JASPER experiments were performed. In order to verify the new version of JSSTDL, whose cross sections of thermal energy region was updated, the polyethylene transmission experiments of the Special Materials Experiment was analysed again, and also zirconium transmission experiment of the Experiment was newly analysed. JSSTDL was applied to the analysis of neutron multiplicative region of the in-vessel fuel storage mockup configurations in the IVS Experiment. Also it was applied to the analyses of neutron streaming effect through the mockup of sodium window in B 4 C shield in the Flux Monitor Experiment and also the mockup of narrow gaps in thick concrete shield in the Gap streaming Experiment. The results were compared with those obtained by the same analysis method and input data using the JSDJ2 library that had been applied consistently to the JASPER experiment analyses. Although the analysis with the new version of JSSTDL resulted in a little reduction of overestimation in the polyethylene transmission configuration, the results obtained with JSSTDL are, in general, higher than those with JSDJ2 as had been found in analyses in preceding years for the Radial Shield Attenuation Experiment, the Axial Shield Experiment, the Intermediate Heat Exchanger Experiment and so on. Compilation of the input data necessary for future reanalyses of important configurations in JASPER experiments, that were selected at the first stage of this study, were continued and new data were added into the computer disk holding previously accumulated data. (author)

Neutron shielding properties of boron-containing ore and epoxy composites

International Nuclear Information System (INIS)

Li Zhifu; Xue Xiangxin

2011-01-01

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

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.)

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.)

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...

Estimation of temperature distribution in a reactor shield

International Nuclear Information System (INIS)

Agarwal, R.A.; Goverdhan, P.; Gupta, S.K.

1989-01-01

Shielding is provided in a nuclear reactor to absorb the radiations emanating from the core. The energy of these radiations appear in the form of heat. Concrete which is commonly used as a shielding material in nuclear power plants must be able to withstand the temperatures and temperature gradients appearing in the shield due to this heat. High temperatures lead to dehydration of the concrete and in turn reduce the shielding effectiveness of the material. Adequate cooling needs to be provided in these shields in order to limit the maximum temperature. This paper describes a method to estimate steady state and transient temperature distribution in reactor shields. The results due to loss of coolant in the coolant tubes have been studied and presented in the paper. (author). 5 figs

An attenuation Layer for Electromagnetic Shielding in X- Band Frequency

Directory of Open Access Journals (Sweden)

Vida Zaroushani

2015-06-01

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

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)

Evaluation of serpentine ore as a nuclear shielding material using gas chromatographic techniques

International Nuclear Information System (INIS)

Singh, B.N.; Unnikrishnan, E.K.; Kumar, Sangita D.

2007-01-01

Serpentine ore mixed with cement has been recognized as a candidate shielding material for use in nuclear reactors because of its many desirable properties. Therefore the assessment of serpentine ore for release of volatile gases during exposure to elevated temperatures, irradiation and changes in chemical composition, is essential. The present paper deals with the studies on the serpentine ores using gas chromatography and combustion gas chromatographic techniques. (author)

Shield structure for a nuclear reactor

International Nuclear Information System (INIS)

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

1979-01-01

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

Characteristics simulation of wireless power transfer system considering shielding distance

Energy Technology Data Exchange (ETDEWEB)

Lee, Yu Kyeong; Choi, Hyo Sang; Jung, Byung Ik; Jeong, In Sung [Chosun University, Gwangju (Korea, Republic of)

2015-03-15

Wireless power transfer technology is using the magnetic resonance recently drawing increased attention. It uses the resonance between transmitter and receiver coils to transfer power. Thus, it can improve the transfer distance and efficiency compared with the existing magnetic induction technique. The authors found from the previous study that the application of the superconductor coil to the magnetic resonance wireless power transfer system improved its efficiency. Its application to real life, however, requires the additional study on the effects of adjacent materials. In this study, the two resonance coils made by superconductor coils were used to aluminum and plastic shielding materials was placed between the coils. S-parameters were analyzed according to the position of the shielding material between the transmitter and receiver coils. As a result, the plastic of shielding material had no effect, but the aluminum of shielding material affected the wireless power transfer due to the shielding effectiveness.

Characteristics simulation of wireless power transfer system considering shielding distance

International Nuclear Information System (INIS)

Lee, Yu Kyeong; Choi, Hyo Sang; Jung, Byung Ik; Jeong, In Sung

2015-01-01

Wireless power transfer technology is using the magnetic resonance recently drawing increased attention. It uses the resonance between transmitter and receiver coils to transfer power. Thus, it can improve the transfer distance and efficiency compared with the existing magnetic induction technique. The authors found from the previous study that the application of the superconductor coil to the magnetic resonance wireless power transfer system improved its efficiency. Its application to real life, however, requires the additional study on the effects of adjacent materials. In this study, the two resonance coils made by superconductor coils were used to aluminum and plastic shielding materials was placed between the coils. S-parameters were analyzed according to the position of the shielding material between the transmitter and receiver coils. As a result, the plastic of shielding material had no effect, but the aluminum of shielding material affected the wireless power transfer due to the shielding effectiveness

Advanced BorobondTM Shields for Nuclear Materials Containment and BorobondTM Immobilization of Volatile Fission Products - Final CRADA Report

International Nuclear Information System (INIS)

Wagh, Arun S.

2016-01-01

Borobond is a company-proprietary material developed by the CRADA partner in collaboration with Argonne, and is based on Argonne's Ceramicrete technology. It is being used by DOE for nuclear materials safe storage, and Boron Products, LLC is the manufacturer and supplier of Borobond. The major objective of this project was to produce a more versatile composition of this material and find new applications. Major target applications were use for nuclear radiation shields, such as in dry storage casks; use in immobilization of most difficult waste streams, such as Hanford K-Basin waste; use for soluble and volatile fission products, such as Cs, Tc, Sr, and I; and use for corrosion and fire protection applications in nuclear facilities.

Study on photon sensitivity of silicon diodes related to materials used for shielding

International Nuclear Information System (INIS)

Moiseev, T.

1999-01-01

Large area silicon diodes used in electronic neutron dosemeters have a significant over-response to X- and gamma-rays, highly non-linear at photon energies below 200 keV. This over-response to photons is proportional to the diode's active area and strongly affects the neutron sensitivity of such dosemeters. Since silicon diodes are sensitive to light and electromagnetic fields, most diode detector assemblies are provided with a shielding, sometimes also used as radiation filter. In this paper, the influence of materials covering the diode's active area is investigated using the MCNP-4A code by estimating the photon induced pulses in a typical silicon wafer (300 μm thickness and 1 cm diameter) when provided with a front case cover. There have been simulated small-size diode front covers made of several materials with low neutron interaction cross-sections like aluminium, TEFLON, iron and lead. The estimated number of induced pulses in the silicon wafer is calculated for each type of shielding at normal photon incidence for several photon energies from 9.8 keV up to 1.15 MeV and compared with that in a bare silicon wafer. The simulated pulse height spectra show the origin of the photon-induced pulses in silicon for each material used as protective cover: the photoelectric effect for low Z front case materials at low-energy incident photons (up to about 65 keV) and the Compton and build-up effects for high Z case materials at higher photon energies. A simple means to lower and flatten the photon response of silicon diodes over an extended X- and gamma rays energy range is proposed by designing a composed photon filter. (author)

Study on Photon Sensitivity of Silicon Diodes Related to Materials Used for Shielding

International Nuclear Information System (INIS)

Moiseev, T.

2000-01-01

Large area Silicon diodes used in electronic neutron dosemeters have a significant over-response to X and gamma rays, highly non-linear at photon energies below 200 keV. This over-response to photons is proportional to the diodes active area and strongly affects the neutron sensitivity of such dosemeters. Since Silicon diodes are sensitive to light and electromagnetic fields, most diode detector assemblies are provided with a shielding, sometimes also used as radiation filter. In this paper, the influence of materials covering the diode's active area is investigated using the MCNP-4A code by estimating the photon induced pulses in a typical silicon wafer (300 μm thickness and 1 cm diameter) when provided with a front case cover. There have been simulated small-size diode front covers made of several materials with low neutron interaction cross-sections like aluminium, TEFLON, iron and lead. The estimated number of induced pulses in the silicon wafer is calculated for each type of shielding at normal photon incidence for several photon energies from 9.8 keV up to 1.15 MeV and compared with that in a bare silicon wafer. The simulated pulse height spectra show the origin of the photon induced pulses in silicon for each material used as protective cover: the photoelectric effect for low Z front case materials at low energy incident photons (up to about 65 keV) and the Compton and build-up effects for high Z case materials at higher photon energies. A simple means to lower and flatten the photon response of silicon diodes over an extended X and gamma rays energy range is proposed by designing a composed photon filter. (author)

Flux trapping and shielding in irreversible superconductors

International Nuclear Information System (INIS)

Frankel, D.J.

1978-05-01

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

Material and electromagnetic properties of Faraday shields for ion cyclotron heating antennas

International Nuclear Information System (INIS)

Hoffman, D.J.; Baity, F.W.; Becraft, W.R.; Caughman, J.B.O.; Tsai, C.C.

1985-01-01

The Faraday shields for ion cyclotron antennas must transmit magnetic waves and absorb little RF power. To investigate these properties, we have constructed 27 Faraday shields in many configurations, including chevrons, tubes, straps, concentric rings, various layered shields, conventionally leafed straps, and replicas of the Faraday shields for ASDEX, the Joint European Torus (JET), TEXTOR, and Alcator-C. We have measured the magnetic flux and observed loading at various operating resistances by using dielectric sheets or magnetic-coupled loads. Each Faraday shield effects a net change in the characteristic inductance of the antenna, resulting in a reduction of wave coupling. However, the load experienced by the antenna is not always reduced because the Faraday shield itself acts as a load. We differentiate between these effects experimentally. The net result of the study is that the Faraday shields now in use cost up to a factor of 50% of coupling. This, of course, reduces the power handling capability by 50% as well. However, configurations exist that are easily cooled and result in a reduction of less than 5% in loading

Material and electromagnetic properties of Faraday shields for ion cyclotron heating antennas

International Nuclear Information System (INIS)

Hoffman, D.J.; Becraft, W.R.; Baity, F.W.; Caughman, J.B.O.; Tsai, C.C.

1985-01-01

The Faraday shields for ion cyclotron antennas must transmit magnetic waves and adsorb little rf power. To investigate these properties, we have constructed 27 Faraday shields in many configurations, including chevrons, tubes, straps, concentric rings, various layered shields, conventionally leafed straps, and replicas of the Faraday shields for ASDEX, the Joint European Torus (JET), TEXTOR, and Alcator-C. We have measured the magnetic flux and observed loading at various operating resistances by using dielectric sheets or magnetic-coupled loads. Each Faraday shield effects a net change in the characteristic inductance of the antenna, resulting in a reduction of wave coupling. However, the load experienced by the antenna is not always reduced because the Faraday shield itself acts as a load. We differentiate between these effects experimentally. The net result of the study is that the Faraday shields now in use cost up to a factor of 50% of coupling. This, of course, reduces the power handling capability by 50% as well. However, configurations exist that are easily cooled and result in a reduction of less than 5% in loading

Materials development for ITER shielding and test blanket in China

Energy Technology Data Exchange (ETDEWEB)

Chen, J.M., E-mail: Chenjm@swip.ac.cn [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041 (China); Wu, J.H.; Liu, X.; Wang, P.H. [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041 (China); Wang, Z.H.; Li, Z.N. [Ningxia Orient Non-ferrous Metals Group Co. Ltd., P.O. Box 105, Shizuishan (China); Wang, X.S.; Zhang, P.C. [China Academy of Engineering Physics, P.O. Box 919-71, Mianyang 621900 (China); Zhang, N.M.; Fu, H.Y.; Liu, D.H. [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041 (China)

2011-10-01

China is a member of the ITER program and is developing her own materials for its shielding and test blanket modules. The materials include vacuum-hot-pressing (VHP) Be, CuCrZr alloy, 316L(N) and China low activation ferritic/martensitic (CLF-1) steels. Joining technologies including Be/Cu hot isostatic pressing (HIP) and electron beam (EB) weldability of 316L(N) were investigated. Chinese VHP-Be showed good properties, with BeO content and ductility that satisfy the ITER requirements. Be/Cu mock-ups were fabricated for Be qualification tests at simulated ITER vertical displacement event (VDE) and heat flux cycling conditions. Fine microstructure and good mechanical strength of the CuCrZr alloy were achieved by a pre-forging treatment, while the weldability of 316L(N) by EB was demonstrated for welding depths varying from 5 to 80 mm. Fine microstructure, high strength, and good ductility were achieved in CLF-1 steel by an optimized normalizing, tempering and aging procedure.

General and crevice corrosion study of the in-wall shielding materials for ITER vacuum vessel

Science.gov (United States)

Joshi, K. S.; Pathak, H. A.; Dayal, R. K.; Bafna, V. K.; Kimihiro, Ioki; Barabash, V.

2012-11-01

Vacuum vessel In-Wall Shield (IWS) will be inserted between the inner and outer shells of the ITER vacuum vessel. The behaviour of IWS in the vacuum vessel especially concerning the susceptibility to crevice of shielding block assemblies could cause rapid and extensive corrosion attacks. Even galvanic corrosion may be due to different metals in same electrolyte. IWS blocks are not accessible until life of the machine after closing of vacuum vessel. Hence, it is necessary to study the susceptibility of IWS materials to general corrosion and crevice corrosion under operations of ITER vacuum vessel. Corrosion properties of IWS materials were studied by using (i) Immersion technique and (ii) Electro-chemical Polarization techniques. All the sample materials were subjected to a series of examinations before and after immersion test, like Loss/Gain weight measurement, SEM analysis, and Optical stereo microscopy, measurement of surface profile and hardness of materials. After immersion test, SS 304B4 and SS 304B7 showed slight weight gain which indicate oxide layer formation on the surface of coupons. The SS 430 material showed negligible weight loss which indicates mild general corrosion effect. On visual observation with SEM and Metallography, all material showed pitting corrosion attack. All sample materials were subjected to series of measurements like Open Circuit potential, Cyclic polarization, Pitting potential, protection potential, Critical anodic current and SEM examination. All materials show pitting loop in OC2 operating condition. However, its absence in OC1 operating condition clearly indicates the activity of chloride ion to penetrate oxide layer on the sample surface, at higher temperature. The critical pitting temperature of all samples remains between 100° and 200°C.

Experimental studies of dynamic impact response with scale models of lead shielded radioactive material shipping containers

International Nuclear Information System (INIS)

Robinson, R.A.; Hadden, J.A.; Basham, S.J.

1978-01-01

Preliminary experimental studies of dynamic impact response of scale models of lead-shielded radioactive material shipping containers are presented. The objective of these studies is to provide DOE/ECT with a data base to allow the prediction of a rational margin of confidence in overviewing and assessing the adequacy of the safety and environmental control provided by these shipping containers. Replica scale modeling techniques were employed to predict full scale response with 1/8, 1/4, and 1/2 scale models of shipping containers that are used in the shipment of spent nuclear fuel and high level wastes. Free fall impact experiments are described for scale models of plain cylindrical stainless steel shells, stainless steel shells filled with lead, and replica scale models of radioactive material shipping containers. Dynamic induced strain and acceleration measurements were obtained at several critical locations on the models. The models were dropped from various heights, attitudes to the impact surface, with and without impact limiters and at uniform temperatures between -40 and 175 0 C. In addition, thermal expansion and thermal gradient induced strains were measured at -40 and 175 0 C. The frequency content of the strain signals and the effect of different drop pad compositions and stiffness were examined. Appropriate scale modeling laws were developed and scaling techniques were substantiated for predicting full scale response by comparison of dynamic strain data for 1/8, 1/4, and 1/2 scale models with stainless steel shells and lead shielding

Recent developments in fusion first wall, blanket, and shield technology

International Nuclear Information System (INIS)

Nygren, R.E.

1983-01-01

This brief overview of first wall, blanket and shield technology reviews the changes and trends in important design issues in first wall, blanket and shield design and related technology from the 1970's to the 1980's. The emphasis is on base technology rather than either systems engineering or materials development. The review is limited to the two primary confinement systems, tokamaks and mirrors, and production of electricity as the primary goal for development

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.)

Shielding requirements for particle bed propulsion systems

Science.gov (United States)

Gruneisen, S. J.

1991-06-01

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

Optimal design of a composite space shield based on numerical simulations

International Nuclear Information System (INIS)

Son, Byung Jin; Yoo, Jeong Hoon; Lee, Min Hyung

2015-01-01

In this study, optimal design of a stuffed Whipple shield is proposed by using numerical simulations and new penetration criterion. The target model was selected based on the shield model used in the Columbus module of the international space station. Because experimental results can be obtained only in the low velocity region below 7 km/s, it is required to derive the Ballistic limit curve (BLC) in the high velocity region above 7 km/s by numerical simulation. AUTODYN-2D, the commercial hydro-code package, was used to simulate the nonlinear transient analysis for the hypervelocity impact. The Smoothed particle hydrodynamics (SPH) method was applied to projectile and bumper modeling to represent the debris cloud generated after the impact. Numerical simulation model and selected material properties were validated through a quantitative comparison between numerical and experimental results. A new criterion to determine whether the penetration occurs or not is proposed from kinetic energy analysis by numerical simulation in the velocity region over 7 km/s. The parameter optimization process was performed to improve the protection ability at a specific condition through the Design of experiment (DOE) method and the Response surface methodology (RSM). The performance of the proposed optimal design was numerically verified.

Self Shielding in Nuclear Fissile Assay Using LSDS

International Nuclear Information System (INIS)

Lee, Yong Deok; Park, Chang Je; Park, Geun Il; Song, Kee Chan

2012-01-01

The new technology for isotopic fissile material contents assay is under development at KAERI using lead slowing down spectrometer(LSDS). LSDS is very sensitive to distinguish fission signals from each fissile isotope in spent and recycled fuel. The accumulation of spent fuel is current big issue. The amount of spent fuels will reach the maximum storage capacity of the pools soon. Therefore, an interim storage must be searched and it should be optimized in design by applying accurate fissile content. When the storage has taken effect, all the nuclear materials must be also specified and verified for safety, economics and management. Generally, the spent fuel from PWR has unburned ∼1 % U235, produced ∼0.5 % plutonium from decay chain, ∼3 % fission products, ∼ 0.1 % minor actinides (MA) and uranium remainder. About 1.5 % fissile materials still exist in the spent fuel. Therefore, for reutilization of fissile materials in spent fuel at SFR, resource material is produced through pyro process. Fissile material contents in resource material must be analyzed before fabricating SFR fuel for reactor safety and economics. In assay of fissile content of spent fuel and recycled fuel, intense radiation background gives limitation on the direct analysis of fissile materials. However, LSDS is not influenced by such a radiation background in fissile assay. Based on the decided geometry setup, self shielding parameter was calculated at the fuel assay zone by introducing spent fuel or pyro produced nuclear material. When nuclear material is inserted into the assay area, the spent fuel assembly or pyro recycled fuel material perturbs the spatial distribution of the slowing down neutrons in lead and the prompt fast fission neutrons produced by fissile materials are also perturbed. The self shielding factor is interpreted as that how much of absorption is created inside the fuel area when it is in the lead. Self shielding effect provides a non-linear property in the isotopic

Effect of metal shielding on a wireless power transfer system

Science.gov (United States)

Li, Jiacheng; Huang, Xueliang; Chen, Chen; Tan, Linlin; Wang, Wei; Guo, Jinpeng

2017-05-01

In this paper, the effect of non-ferromagnetic metal shielding (NFMS) material on the resonator of wireless power transfer (WPT) is studied by modeling, simulation and experimental analysis. And, the effect of NFMS material on the power transfer efficiency (PTE) of WPT systems is investigated by circuit model. Meanwhile, the effect of ferromagnetic metal shielding material on the PTE of WPT systems is analyzed through simulation. A double layer metal shield structure is designed. Experimental results demonstrate that by applying the novel double layer metal shielding method, the system PTE increases significantly while the electromagnetic field of WPT systems declines dramatically.

Gonadal shield.

Science.gov (United States)

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

1975-10-01

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

Beam loss reduction by magnetic shielding using beam pipes and bellows of soft magnetic materials

Science.gov (United States)

Kamiya, J.; Ogiwara, N.; Hotchi, H.; Hayashi, N.; Kinsho, M.

2014-11-01

One of the main sources of beam loss in high power accelerators is unwanted stray magnetic fields from magnets near the beam line, which can distort the beam orbit. The most effective way to shield such magnetic fields is to perfectly surround the beam region without any gaps with a soft magnetic high permeability material. This leads to the manufacture of vacuum chambers (beam pipes and bellows) with soft magnetic materials. A Ni-Fe alloy (permalloy) was selected for the material of the pipe parts and outer bellows parts, while a ferritic stainless steel was selected for the flanges. An austenitic stainless steel, which is non-magnetic material, was used for the inner bellows for vacuum tightness. To achieve good magnetic shielding and vacuum performances, a heat treatment under high vacuum was applied during the manufacturing process of the vacuum chambers. Using this heat treatment, the ratio of the integrated magnetic flux density along the beam orbit between the inside and outside of the beam pipe and bellows became small enough to suppress beam orbit distortion. The outgassing rate of the materials with this heat treatment was reduced by one order magnitude compared to that without heat treatment. By installing the beam pipes and bellows of soft magnetic materials as part of the Japan Proton Accelerator Research Complex 3 GeV rapid cycling synchrotron beam line, the closed orbit distortion (COD) was reduced by more than 80%. In addition, a 95.5% beam survival ratio was achieved by this COD improvement.

Paddle-based rotating-shield brachytherapy

Energy Technology Data Exchange (ETDEWEB)

Liu, Yunlong; Xu, Weiyu [Department of Electrical and Computer Engineering, University of Iowa, 4016 Seamans Center, Iowa City, Iowa 52242 (United States); Flynn, Ryan T.; Kim, Yusung; Bhatia, Sudershan K.; Buatti, John M. [Department of Radiation Oncology, University of Iowa, 200 Hawkins Drive, Iowa City, Iowa 52242 (United States); Dadkhah, Hossein [Department of Biomedical Engineering, University of Iowa, 1402 Seamans Center, Iowa City, Iowa 52242 (United States); Wu, Xiaodong, E-mail: xiaodong-wu@uiowa.edu [Department of Electrical and Computer Engineering, University of Iowa, 4016 Seamans Center, Iowa City, Iowa 52242 and Department of Radiation Oncology, University of Iowa, 200 Hawkins Drive, Iowa City, Iowa 52242 (United States)

2015-10-15

Purpose: The authors present a novel paddle-based rotating-shield brachytherapy (P-RSBT) method, whose radiation-attenuating shields are formed with a multileaf collimator (MLC), consisting of retractable paddles, to achieve intensity modulation in high-dose-rate brachytherapy. Methods: Five cervical cancer patients using an intrauterine tandem applicator were considered to assess the potential benefit of the P-RSBT method. The P-RSBT source used was a 50 kV electronic brachytherapy source (Xoft Axxent™). The paddles can be retracted independently to form multiple emission windows around the source for radiation delivery. The MLC was assumed to be rotatable. P-RSBT treatment plans were generated using the asymmetric dose–volume optimization with smoothness control method [Liu et al., Med. Phys. 41(11), 111709 (11pp.) (2014)] with a delivery time constraint, different paddle sizes, and different rotation strides. The number of treatment fractions (fx) was assumed to be five. As brachytherapy is delivered as a boost for cervical cancer, the dose distribution for each case includes the dose from external beam radiotherapy as well, which is 45 Gy in 25 fx. The high-risk clinical target volume (HR-CTV) doses were escalated until the minimum dose to the hottest 2 cm{sup 3} (D{sub 2cm{sup 3}}) of either the rectum, sigmoid colon, or bladder reached their tolerance doses of 75, 75, and 90 Gy{sub 3}, respectively, expressed as equivalent doses in 2 Gy fractions (EQD2 with α/β = 3 Gy). Results: P-RSBT outperformed the two other RSBT delivery techniques, single-shield RSBT (S-RSBT) and dynamic-shield RSBT (D-RSBT), with a properly selected paddle size. If the paddle size was angled at 60°, the average D{sub 90} increases for the delivery plans by P-RSBT on the five cases, compared to S-RSBT, were 2.2, 8.3, 12.6, 11.9, and 9.1 Gy{sub 10}, respectively, with delivery times of 10, 15, 20, 25, and 30 min/fx. The increases in HR-CTV D{sub 90}, compared to D-RSBT, were 16

Pre-evaluation of fusion shielding benchmark experiment

International Nuclear Information System (INIS)

Hayashi, K.; Handa, H.; Konno, C.

1994-01-01

Shielding benchmark experiment is very useful to test the design code and nuclear data for fusion devices. There are many types of benchmark experiments that should be done in fusion shielding problems, but time and budget are limited. Therefore it will be important to select and determine the effective experimental configurations by precalculation before the experiment. The authors did three types of pre-evaluation to determine the experimental assembly configurations of shielding benchmark experiments planned in FNS, JAERI. (1) Void Effect Experiment - The purpose of this experiment is to measure the local increase of dose and nuclear heating behind small void(s) in shield material. Dimension of the voids and its arrangements were decided as follows. Dose and nuclear heating were calculated both for with and without void(s). Minimum size of the void was determined so that the ratio of these two results may be larger than error of the measurement system. (2) Auxiliary Shield Experiment - The purpose of this experiment is to measure shielding properties of B 4 C, Pb, W, and dose around superconducting magnet (SCM). Thickness of B 4 C, Pb, W and their arrangement including multilayer configuration were determined. (3) SCM Nuclear Heating Experiment - The purpose of this experiment is to measure nuclear heating and dose distribution in SCM material. Because it is difficult to use liquid helium as a part of SCM mock up material, material composition of SCM mock up are surveyed to have similar nuclear heating property of real SCM composition

Computing Moment-Based Probability Tables for Self-Shielding Calculations in Lattice Codes

International Nuclear Information System (INIS)

Hebert, Alain; Coste, Mireille

2002-01-01

As part of the self-shielding model used in the APOLLO2 lattice code, probability tables are required to compute self-shielded cross sections for coarse energy groups (typically with 99 or 172 groups). This paper describes the replacement of the multiband tables (typically with 51 subgroups) with moment-based tables in release 2.5 of APOLLO2. An improved Ribon method is proposed to compute moment-based probability tables, allowing important savings in CPU resources while maintaining the accuracy of the self-shielding algorithm. Finally, a validation is presented where the absorption rates obtained with each of these techniques are compared with exact values obtained using a fine-group elastic slowing-down calculation in the resolved energy domain. Other results, relative to the Rowland's benchmark and to three assembly production cases, are also presented

SU-E-T-523: Investigation of Various MR-Compatible Shielding Materials for Direction Modulated Brachytherapy (DMBT) Tandem Applicator for Cervical Cancer Treatment

Energy Technology Data Exchange (ETDEWEB)

Safigholi, H; Soliman, A; Song, W [Sunnybrook Research Institute, Sunnybrook Health Sciences Centre, U of T, Toronto, Ontario (Canada); Han, D [Sunnybrook Research Institute, Sunnybrook Health Sciences Centre, U of T, Toronto, Ontario (Canada); University of California, San Diego, La Jolla, CA (United States); Meigooni, A Soleimani [Comprehensive Cancer Center of Nevada, Las Vegas, NV (United States); Scanderbeg, D [UCSD Medical Center, La Jolla, CA (United States)

2015-06-15

Purpose: To evaluate various shielding materials such as Gold (Au), Osmium (Os), Tantalum (Ta), and Tungsten (W) based alloys for use with a novel intensity modulation capable direction modulated brachytherapy (DMBT) tandem applicator for image guided cervical cancer HDR brachytherapy. Methods: The novel MRI-compatible DMBT tandem, made from nonmagnetic tungsten-alloy rod with diameter of 5.4 mm, has 6 symmetric peripheral holes of 1.3 mm diameter with 2.05 mm distance from the center for a high degree intensity modulation capacity. The 0.3 mm thickness of bio-compatible plastic tubing wraps the tandem. MCNPX was used for Monte Carlo simulations of the shields and the mHDR Ir-192 V2 source. MC-generated 3D dose matrices of different shielding materials of Au, Os, Ta, and W with 1 mm3 resolution were imported into an in-house-coded inverse optimization planning system to evaluate 19 clinical patient plans. Prescription dose was 15Gy. All plans were normalized to receive the same HRCTV D90. Results: In general, the plan qualities for various shielding materials were similar. The OAR D2cc for bladder was very similar for Au, Os, and Ta with 11.64±2.30Gy. For W, it was very close 11.65±2.30Gy. The sigmoid D2cc was 9.82±2.46Gy for Au and Os while it was 9.84±2.48Gy for Ta and W. The rectum D2cc was 7.44±3.06Gy for Au, 7.43±3.07Gy for Os, 7.48±3.05Gy for Ta, and 7.47±3.05Gy for W. The HRCTV D98 and V100 were very close with 16.37±1.87 Gy and 97.37±1.93 Gy, on average, respectively. Conclusion: Various MRI-compatible shielding alloys were investigated for the DMBT tandem applicator. The clinical plan qualities were not significantly different among these various alloys, however. Therefore, the candidate metals (or in combination) can be used to select best alloys for MRI image guided cervical cancer brachytherapy using the novel DMBT applicator that is capable of unprecedented level of intensity modulation.

SU-E-T-523: Investigation of Various MR-Compatible Shielding Materials for Direction Modulated Brachytherapy (DMBT) Tandem Applicator for Cervical Cancer Treatment

International Nuclear Information System (INIS)

Safigholi, H; Soliman, A; Song, W; Han, D; Meigooni, A Soleimani; Scanderbeg, D

2015-01-01

Purpose: To evaluate various shielding materials such as Gold (Au), Osmium (Os), Tantalum (Ta), and Tungsten (W) based alloys for use with a novel intensity modulation capable direction modulated brachytherapy (DMBT) tandem applicator for image guided cervical cancer HDR brachytherapy. Methods: The novel MRI-compatible DMBT tandem, made from nonmagnetic tungsten-alloy rod with diameter of 5.4 mm, has 6 symmetric peripheral holes of 1.3 mm diameter with 2.05 mm distance from the center for a high degree intensity modulation capacity. The 0.3 mm thickness of bio-compatible plastic tubing wraps the tandem. MCNPX was used for Monte Carlo simulations of the shields and the mHDR Ir-192 V2 source. MC-generated 3D dose matrices of different shielding materials of Au, Os, Ta, and W with 1 mm3 resolution were imported into an in-house-coded inverse optimization planning system to evaluate 19 clinical patient plans. Prescription dose was 15Gy. All plans were normalized to receive the same HRCTV D90. Results: In general, the plan qualities for various shielding materials were similar. The OAR D2cc for bladder was very similar for Au, Os, and Ta with 11.64±2.30Gy. For W, it was very close 11.65±2.30Gy. The sigmoid D2cc was 9.82±2.46Gy for Au and Os while it was 9.84±2.48Gy for Ta and W. The rectum D2cc was 7.44±3.06Gy for Au, 7.43±3.07Gy for Os, 7.48±3.05Gy for Ta, and 7.47±3.05Gy for W. The HRCTV D98 and V100 were very close with 16.37±1.87 Gy and 97.37±1.93 Gy, on average, respectively. Conclusion: Various MRI-compatible shielding alloys were investigated for the DMBT tandem applicator. The clinical plan qualities were not significantly different among these various alloys, however. Therefore, the candidate metals (or in combination) can be used to select best alloys for MRI image guided cervical cancer brachytherapy using the novel DMBT applicator that is capable of unprecedented level of intensity modulation

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.

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

Directory of Open Access Journals (Sweden)

Suwimon Ruengsri

2014-01-01

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

Radiation shield for PWR reactors

International Nuclear Information System (INIS)

Esenov, Amra; Pustovgar, Andrey

2013-01-01

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

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)

Radiation shielding material characterization by non-destructive neutron radiography technique

International Nuclear Information System (INIS)

Hafizal Yazid; Azali Muhammad; Abdul Aziz Mohamed; Rafhayudi Jamro; Hishamuddin Husain

2007-01-01

Shielding property of boronated rubber was characterized easily by the use of neutron radiography technique. For 10 phr of boron carbide in the natural rubber composite, the ability to completely shield against neutron was found to have 8mm thickness and above for the neutron flux of 1.04 x 10 5 n/cm 2 s (author)

Shielded scanning electron microscope for radioactive samples

International Nuclear Information System (INIS)

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

1977-01-01

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

Effect of metal shielding on a wireless power transfer system

Directory of Open Access Journals (Sweden)

Jiacheng Li

2017-05-01

Full Text Available In this paper, the effect of non-ferromagnetic metal shielding (NFMS material on the resonator of wireless power transfer (WPT is studied by modeling, simulation and experimental analysis. And, the effect of NFMS material on the power transfer efficiency (PTE of WPT systems is investigated by circuit model. Meanwhile, the effect of ferromagnetic metal shielding material on the PTE of WPT systems is analyzed through simulation. A double layer metal shield structure is designed. Experimental results demonstrate that by applying the novel double layer metal shielding method, the system PTE increases significantly while the electromagnetic field of WPT systems declines dramatically.

Non-contact radio frequency shielding and wave guiding by multi-folded transformation optics method.

Science.gov (United States)

Madni, Hamza Ahmad; Zheng, Bin; Yang, Yihao; Wang, Huaping; Zhang, Xianmin; Yin, Wenyan; Li, Erping; Chen, Hongsheng

2016-11-14

Compared with conventional radio frequency (RF) shielding methods in which the conductive coating material encloses the circuits design and the leakage problem occurs due to the gap in such conductive material, non-contact RF shielding at a distance is very promising but still impossible to achieve so far. In this paper, a multi-folded transformation optics method is proposed to design a non-contact device for RF shielding. This "open-shielded" device can shield any object at a distance from the electromagnetic waves at the operating frequency, while the object is still physically open to the outer space. Based on this, an open-carpet cloak is proposed and the functionality of the open-carpet cloak is demonstrated. Furthermore, we investigate a scheme of non-contact wave guiding to remotely control the propagation of surface waves over any obstacles. The flexibilities of such multi-folded transformation optics method demonstrate the powerfulness of the method in the design of novel remote devices with impressive new functionalities.

CAD-Based Shielding Analysis for ITER Port Diagnostics

Directory of Open Access Journals (Sweden)

Serikov Arkady

2017-01-01

Full Text Available Radiation shielding analysis conducted in support of design development of the contemporary diagnostic systems integrated inside the ITER ports is relied on the use of CAD models. This paper presents the CAD-based MCNP Monte Carlo radiation transport and activation analyses for the Diagnostic Upper and Equatorial Port Plugs (UPP #3 and EPP #8, #17. The creation process of the complicated 3D MCNP models of the diagnostics systems was substantially accelerated by application of the CAD-to-MCNP converter programs MCAM and McCad. High performance computing resources of the Helios supercomputer allowed to speed-up the MCNP parallel transport calculations with the MPI/OpenMP interface. The found shielding solutions could be universal, reducing ports R&D costs. The shield block behind the Tritium and Deposit Monitor (TDM optical box was added to study its influence on Shut-Down Dose Rate (SDDR in Port Interspace (PI of EPP#17. Influence of neutron streaming along the Lost Alpha Monitor (LAM on the neutron energy spectra calculated in the Tangential Neutron Spectrometer (TNS of EPP#8. For the UPP#3 with Charge eXchange Recombination Spectroscopy (CXRS-core, an excessive neutron streaming along the CXRS shutter, which should be prevented in further design iteration.

CAD-Based Shielding Analysis for ITER Port Diagnostics

Science.gov (United States)

Serikov, Arkady; Fischer, Ulrich; Anthoine, David; Bertalot, Luciano; De Bock, Maartin; O'Connor, Richard; Juarez, Rafael; Krasilnikov, Vitaly

2017-09-01

Radiation shielding analysis conducted in support of design development of the contemporary diagnostic systems integrated inside the ITER ports is relied on the use of CAD models. This paper presents the CAD-based MCNP Monte Carlo radiation transport and activation analyses for the Diagnostic Upper and Equatorial Port Plugs (UPP #3 and EPP #8, #17). The creation process of the complicated 3D MCNP models of the diagnostics systems was substantially accelerated by application of the CAD-to-MCNP converter programs MCAM and McCad. High performance computing resources of the Helios supercomputer allowed to speed-up the MCNP parallel transport calculations with the MPI/OpenMP interface. The found shielding solutions could be universal, reducing ports R&D costs. The shield block behind the Tritium and Deposit Monitor (TDM) optical box was added to study its influence on Shut-Down Dose Rate (SDDR) in Port Interspace (PI) of EPP#17. Influence of neutron streaming along the Lost Alpha Monitor (LAM) on the neutron energy spectra calculated in the Tangential Neutron Spectrometer (TNS) of EPP#8. For the UPP#3 with Charge eXchange Recombination Spectroscopy (CXRS-core), an excessive neutron streaming along the CXRS shutter, which should be prevented in further design iteration.

Self-shielding characteristics of aqueous self-cooled blankets for next generation fusion devices

International Nuclear Information System (INIS)

Pelloni, S.; Cheng, E.T.; Embrechts, M.J.

1987-11-01

The present study examines self-shielding characteristics for two aqueous self-cooled tritium producing driver blankets for next generation fusion devices. The aqueous Self-Cooled Blanket concept (ASCB) is a very simple blanket concept that relies on just structural material and coolant. Lithium compounds are dissolved in water to provide for tritium production. An ASCB driver blanket would provide a low technology and low temperature environment for blanket test modules in a next generation fusion reactor. The primary functions of such a blanket would be shielding, energy removal and tritium production. One driver blanket considered in this study concept relates to the one proposed for the Next European Torus (NET), while the second concept is indicative for the inboard shield design for the Engineering Test Reactor proposed by the USA (TIBER II/ETR). The driver blanket for NET is based on stainless steel for the structural material and aqueous solution, while the inboard shielding blanket for TIBER II/ETR is based on a tungsten/aqueous solution combination. The purpose of this study is to investigate self-shielding and heterogeneity effects in aqueous self-cooled blankets. It is found that no significant gains in tritium breeding can be achieved in the stainless steel blanket if spatial and energy self-shielding effects are considered, and the heterogeneity effects are also insignificant. The tungsten blanket shows a 5 percent increase in tritium production in the shielding blanket when energy and spatial self-shielding effects are accounted for. However, the tungsten blanket shows a drastic increase in the tritium breeding ratio due to heterogeneity effects. (author) 17 refs., 9 figs., 9 tabs

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

International Nuclear Information System (INIS)

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

1984-04-01

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

Annotated references on shielding experiment and calculation of high energy particles

International Nuclear Information System (INIS)

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

1990-12-01

The literature on shielding experiment and calculation of high energy particles above 20 MeV has been surveyed. The survey covers thirteen journals, from 1965 up to 1989. For each paper, applicable information is listed on type and energy of the projectile, the accelerator used, composition and thickness of the target and shielding materials, shielding geometry, the experimental and calculational methods, and the quantities obtained. The references on shielding experiment and on shielding calculation are accessed through two indices which list the projectile-target and shielding material combination, shielding geometry and the projectile energy range. The literature on neutron, photon and hadron production from thick target bombarded by charged particles has been surveyed mainly from 1984 as a complement of the previous work. (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)

Neutron shieldings

International Nuclear Information System (INIS)

Tarutani, Kohei

1979-01-01

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

Guide to shielding calculations for the design of fluoroscopic laboratory at 503 workshop AVN base Rawalpindi

International Nuclear Information System (INIS)

Din, J.U.; Ahmad, M.; Ashraf, M.M.; Khan, A.R.; Khan, A.A.

1986-11-01

Non-destructive testing plays an important role in assessing the quality of materials. Various methods are used for this purpose. Radiography by X-rays and gamma-rays is one of the NDT methods used. There are number of mathematical formulae used to estimate the required shielding for an X-ray tube operating at maximum rated voltage or a gamma radiation source having fixed energies. This report covers the shielding requirements for a 150 KV constant potential X-ray unit operating at maximum rated voltage. In addition, the report is a guide for the design of shielded enclosure required for X-rays machines in general. (orig./A.B.)

Shielding designs and tests of a new exclusive ship for transporting spent nuclear fuels

International Nuclear Information System (INIS)

Ono, M.; Ito, D.; Kitano, T.; Ueki, K.; Akiyama, H.; Obara, I.; Sanui, T.

2000-01-01

The Rokuei-Maru, a ship built specially for the transport of spent nuclear fuels in casks, was launched April in 1996. She is the first ship to comply with special Japanese regulations, KAISA 520, based on the INF code. DOT3.5 and MCNP-4A were used for the evaluation of dose equivalent rates of her shielding structures. On-board gamma-ray shielding tests were executed to confirm the effectiveness of the ship's shielding performance. The tests confirmed that effective shielding has been achieved and the dose equivalent rate in the accommodation and other inhabited spaces is sufficiently lower than the regulated limitations. This was achieved by employing the appropriate calculation methods and shielding materials. (author)

Role of Outgassing of ITER Vacuum Vessel In-Wall Shielding Materials in Leak Detection of ITER Vacuum Vessel

Science.gov (United States)

Maheshwari, A.; Pathak, H. A.; Mehta, B. K.; Phull, G. S.; Laad, R.; Shaikh, M. S.; George, S.; Joshi, K.; Khan, Z.

2017-04-01

ITER Vacuum Vessel is a torus-shaped, double wall structure. The space between the double walls of the VV is filled with In-Wall Shielding Blocks (IWS) and Water. The main purpose of IWS is to provide neutron shielding during ITER plasma operation and to reduce ripple of Toroidal Magnetic Field (TF). Although In-Wall Shield Blocks (IWS) will be submerged in water in between the walls of the ITER Vacuum Vessel (VV), Outgassing Rate (OGR) of IWS materials plays a significant role in leak detection of Vacuum Vessel of ITER. Thermal Outgassing Rate of a material critically depends on the Surface Roughness of material. During leak detection process using RGA equipped Leak detector and tracer gas Helium, there will be a spill over of mass 3 and mass 2 to mass 4 which creates a background reading. Helium background will have contribution of Hydrogen too. So it is necessary to ensure the low OGR of Hydrogen. To achieve an effective leak test it is required to obtain a background below 1 × 10-8 mbar 1 s-1 and hence the maximum Outgassing rate of IWS Materials should comply with the maximum Outgassing rate required for hydrogen i.e. 1 x 10-10 mbar 1 s-1 cm-2 at room temperature. As IWS Materials are special materials developed for ITER project, it is necessary to ensure the compliance of Outgassing rate with the requirement. There is a possibility of diffusing the gasses in material at the time of production. So, to validate the production process of materials as well as manufacturing of final product from this material, three coupons of each IWS material have been manufactured with the same technique which is being used in manufacturing of IWS blocks. Manufacturing records of these coupons have been approved by ITER-IO (International Organization). Outgassing rates of these coupons have been measured at room temperature and found in acceptable limit to obtain the required Helium Background. On the basis of these measurements, test reports have been generated and got

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.

Summary of Blast Shield and Material Testing for Development of Solid Debris Collection at the National Ignition Facility (NIF)

International Nuclear Information System (INIS)

Shaughnessy, D.A.; Gostic, J.M.; Moody, K.J.; Grant, P.M.; Lewis, L.A.; Hutcheon, I.D.

2011-01-01

this report and will be written up at a later time. Based on the results of the aluminum blast shield analysis, it was determined that additional materials needed to be tested as potential collectors in the NIF chamber. 1-2 mm thick pieces of tantalum, niobium, vanadium, silver, titanium, molybdenum, and graphite foil were fielded in the Wedge Range Filter (WRF) mount at a distance of 50 cm from target chamber center during the shock timing campaign. The pieces were subsequently removed and analyzed in a similar fashion to the aluminum shields. As of this writing, the pieces are still under analysis, but initial results indicate that gold debris was collected on the various materials. Currently, the pieces are being cross-sectioned so that the melt depths of each material can be compared. In addition, NAA and/or mass spectrometry will be performed in order to determine the total gold mass that was collected on each surface.

Shielding benchmark problems, (2)

International Nuclear Information System (INIS)

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

1980-02-01

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

Preliminary shielding analysis of VHTR reactors

International Nuclear Information System (INIS)

Flaspoehler, Timothy M.; Petrovic, Bojan

2011-01-01

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

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)

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.)

Tests of shielding effectiveness of Kevlar and Nextel onboard the International Space Station and the Foton-M3 capsule.

Science.gov (United States)

Pugliese, M; Bengin, V; Casolino, M; Roca, V; Zanini, A; Durante, M

2010-08-01

Radiation assessment and protection in space is the first step in planning future missions to the Moon and Mars, where mission and number of space travelers will increase and the protection of the geomagnetic shielding against the cosmic radiation will be absent. In this framework, the shielding effectiveness of two flexible materials, Kevlar and Nextel, were tested, which are largely used in the construction of spacecrafts. Accelerator-based tests clearly demonstrated that Kevlar is an excellent shield for heavy ions, close to polyethylene, whereas Nextel shows poor shielding characteristics. Measurements on flight performed onboard of the International Space Station and of the Foton-M3 capsule have been carried out with special attention to the neutron component; shielded and unshielded detectors (thermoluminescence dosemeters, bubble detectors) were exposed to a real radiation environment to test the shielding properties of the materials under study. The results indicate no significant effects of shielding, suggesting that thin shields in low-Earth Orbit have little effect on absorbed dose.

HPGe detector shielding adjustment

International Nuclear Information System (INIS)

Trnkova, L.; Rulik, P.

2008-01-01

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

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.

Shield Optimization and Formulation of Regression Equations for Split-Ring Resonator

Directory of Open Access Journals (Sweden)

Tahir Ejaz

2016-01-01

Full Text Available Microwave resonators are widely used for numerous applications including communication, biomedical and chemical applications, material testing, and food grading. Split-ring resonators in both planar and nonplanar forms are a simple structure which has been in use for several decades. This type of resonator is characterized with low cost, ease of fabrication, moderate quality factor, low external noise interference, high stability, and so forth. Due to these attractive features and ease in handling, nonplanar form of structure has been utilized for material characterization in 1–5 GHz range. Resonant frequency and quality factor are two important parameters for determination of material properties utilizing perturbation theory. Shield made of conducting material is utilized to enclose split-ring resonator which enhances quality factor. This work presents a novel technique to develop shield around a predesigned nonplanar split-ring resonator to yield optimized quality factor. Based on this technique and statistical analysis regression equations have also been formulated for resonant frequency and quality factor which is a major outcome of this work. These equations quantify dependence of output parameters on various factors of shield made of different materials. Such analysis is instrumental in development of devices/designs where improved/optimum result is required.

Contaminant deposition building shielding factors for US residential structures.

Science.gov (United States)

Dickson, Elijah; Hamby, David; Eckerman, Keith

2017-10-10

This paper presents validated building shielding factors designed for contemporary US housing-stock under an idealized, yet realistic, exposure scenario from contaminant deposition on the roof and surrounding surfaces. The building shielding factors are intended for use in emergency planning and level three probabilistic risk assessments for a variety of postulated radiological events in which a realistic assessment is necessary to better understand the potential risks for accident mitigation and emergency response planning. Factors are calculated from detailed computational housing-units models using the general-purpose Monte Carlo N-Particle computational code, MCNP5, and are benchmarked from a series of narrow- and broad-beam measurements analyzing the shielding effectiveness of ten common general-purpose construction materials and ten shielding models representing the primary weather barriers (walls and roofs) of likely US housing-stock. Each model was designed to scale based on common residential construction practices and include, to the extent practical, all structurally significant components important for shielding against ionizing radiation. Calculations were performed for floor-specific locations from contaminant deposition on the roof and surrounding ground as well as for computing a weighted-average representative building shielding factor for single- and multi-story detached homes, both with and without basement as well for single-wide manufactured housing-unit. © 2017 IOP Publishing Ltd.

Contaminant deposition building shielding factors for US residential structures

International Nuclear Information System (INIS)

Dickson, E D; Hamby, D M; Eckerman, K F

2015-01-01

This paper presents validated building shielding factors designed for contemporary US housing-stock under an idealized, yet realistic, exposure scenario from contaminant deposition on the roof and surrounding surfaces. The building shielding factors are intended for use in emergency planning and level three probabilistic risk assessments for a variety of postulated radiological events in which a realistic assessment is necessary to better understand the potential risks for accident mitigation and emergency response planning. Factors are calculated from detailed computational housing-units models using the general-purpose Monte Carlo N-Particle computational code, MCNP5, and are benchmarked from a series of narrow- and broad-beam measurements analyzing the shielding effectiveness of ten common general-purpose construction materials and ten shielding models representing the primary weather barriers (walls and roofs) of likely US housing-stock. Each model was designed to scale based on common residential construction practices and include, to the extent practical, all structurally significant components important for shielding against ionizing radiation. Calculations were performed for floor-specific locations from contaminant deposition on the roof and surrounding ground as well as for computing a weighted-average representative building shielding factor for single- and multi-story detached homes, both with and without basement as well for single-wide manufactured housing-unit. (paper)

An analytical solution to the shielding of Co 60 teletherapy rooms based on a semiempirical equation of photon attenuation

International Nuclear Information System (INIS)

Saez, D.G.; Hernandez, L.; Borroto, M.; Figueredo, M.

1996-01-01

A semiempirical equation of polynomial-exponential type is presented to describe the transmission data of Co-60 gamma radiation in finite materials of concrete and lead. This equation and the expression obtained for the relationship of scatter-to-incident exposure made easy the developing in computer of an analytical solution for shielding calculations of Co 60 teletherapy rooms, based on the procedures of the NCRP 49 and Simpkin's method. The standard error in the estimation of parameters is less than 1.7 % except for the attenuation of 150 'o' scattered radiation in concrete that resulted in 6.3 % for one of them. The shielding calculations were compared with the data in NCRP 49 for the same conditions with a correlation better than 99 %

3D carbon fiber mats/nano-Fe3O4 hybrid material with high electromagnetic shielding performance

Science.gov (United States)

Zhan, Yingqing; Long, Zhihang; Wan, Xinyi; Zhang, Jiemin; He, Shuangjiang; He, Yi

2018-06-01

To obtain high-performance electromagnetic shielding materials, structure and morphology are two key factors. We here developed an efficient and facial method to prepare high-performance 3D carbon nanofiber mats (CFM)/Fe3O4 hybrid electromagnetic shielding materials. For this purpose, the CFM were chemically modified by mussel-inspired poly-dopamine coating, which were further used as templates for decoration of Fe3O4 nanoparticles via solvothermal route. It was found that the Fe3O4 nano-spheres with diameters of 200-250 nm were uniformly coated on the surface of 3D carbon nanofibers. More importantly, the morphology and structure of resulting 3D carbon nanofiber mats/Fe3O4 hybrids could be easily controlled by altering the experiment parameters, which were examined by FT-IR, XPS, TGA, XRD, SEM, and TEM. The measured magnetic properties showed that saturation magnetism and coercivity increased from 13.4 to 39.7 emu/g and 85.3 to 104.6 Oe, respectively. The lowest reflectivity of resulting hybrid was calculated to be -47 dB at 10.0 GHz (2.5 mm). In addition, the reflectivity of 3D carbon nanofiber mats/Fe3O4 hybrid was less than -25 dB in the range of 7-13 GHz. Moreover, the resulting 3D carbon nanofiber mats/Fe3O4 hybrid exhibited an EMI shielding performance of -62.6 dB in the frequency range of 8.2-12.4 GHz. Therefore, 3D carbon fiber mats/Fe3O4 hybrids can be ideal EMI materials with strong absorption, low density, and wide absorption range.

Preparation and Study of Electromagnetic Interference Shielding Materials Comprised of Ni-Co Coated on Web-Like Biocarbon Nanofibers via Electroless Deposition

Directory of Open Access Journals (Sweden)

Xiaohu Huang

2015-01-01

Full Text Available Electromagnetic interference (EMI shielding materials made of Ni-Co coated on web-like biocarbon nanofibers were successfully prepared by electroless plating. Biocarbon nanofibers (CF with a novel web-like structure comprised of entangled and interconnected carbon nanoribbons were obtained using bacterial cellulose pyrolyzed at 1200°C. Paraffin wax matrix composites filled with different loadings (10, 20, and 30 wt%, resp. of CF and Ni-Co coated CF (NCCF were prepared. The electrical conductivities and electromagnetic parameters of the composites were investigated by the four-probe method and vector network analysis. From these results, the EMI shielding efficiencies (SE of NCCF composites were shown to be significantly higher than that of CF at the same mass fraction. The paraffin wax composites containing 30 wt% NCCF showed the highest EMI SE of 41.2 dB (99.99% attenuation, which are attributed to the higher electrical conductivity and permittivity of the NCCF composites than the CF composites. Additionally, EMI SE increased with an increase in CF and NCCF loading and the absorption was determined to be the primary factor governing EMI shielding. This study conclusively reveals that NCCF composites have potential applications as EMI shielding materials.

A Reliability Comparison of Classical and Stochastic Thickness Margin Approaches to Address Material Property Uncertainties for the Orion Heat Shield

Science.gov (United States)

Sepka, Steve; Vander Kam, Jeremy; McGuire, Kathy

2018-01-01

The Orion Thermal Protection System (TPS) margin process uses a root-sum-square approach with branches addressing trajectory, aerothermodynamics, and material response uncertainties in ablator thickness design. The material response branch applies a bond line temperature reduction between the Avcoat ablator and EA9394 adhesive by 60 C (108 F) from its peak allowed value of 260 C (500 F). This process is known as the Bond Line Temperature Material Margin (BTMM) and is intended to cover material property and performance uncertainties. The value of 60 C (108 F) is a constant, applied at any spacecraft body location and for any trajectory. By varying only material properties in a random (monte carlo) manner, the perl-based script mcCHAR is used to investigate the confidence interval provided by the BTMM. In particular, this study will look at various locations on the Orion heat shield forebody for a guided and an abort (ballistic) trajectory.

Shielding of manned space stations against Van Allen Belt protons: a preliminary scoping study

International Nuclear Information System (INIS)

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

1986-09-01

Calculated results are presented to aid in the design of the shielding required to protect astronauts in a space station that is orbiting through the Van Allen proton belt. The geometry considered - a spherical shell shield with a spherical tissue phantom at its center - is only a very approximate representation of an actual space station, but this simple geometry makes it possible to consider a wide range of possible shield materials. Both homogeneous and laminated shields are considered. Also, an approximation procedure - the equivalent thickness approximation - that allows dose rates to be estimated for any shield material or materials from the dose rates for an aluminum shield is presented and discussed

A polynomial–based function approach to point isotropic gamma-ray buildup factor data in double layered spherical shield of water and lead

Directory of Open Access Journals (Sweden)

M. H. Alamatsaz

2014-03-01

Full Text Available As the input of MCNP code (Monte Carlo N - Particle code system, a monoenergetic and isotropic point source with the energy rangeg from 0.3 to 10 MeV was placed at the center of a spherical material surrounded by another one. The first shielding material was water and the second one was lead. The total thickness of the shield varied between 2 to 10 mfp. Then, using the output of MCNCP, exposure build up factor was calculated. The MCNP computed data were analyzed by plotting the buildup factor as a function of each independent variable (energy, first material thickness and second material thickness and observing the trends. Based on the trends, we examined many different expressions with different number of constants. By MINUIT the FORTRAN program, the constants were calculated, which gave the best agreement between the MCNP-computed exposure buildup factors and those obtained by the formula. At last, we developed a polynomial formula with 11 constants that reproduced exposure buildup factor with a relative error below 2% (in comparison with the MCNP result.

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.

Evaluation of rubber composites as shielding materials against ionizing radiation

International Nuclear Information System (INIS)

Atia, M.K.

2010-01-01

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

The performance test of anti-scattering x-ray grid with inclined shielding material by MCNP code simulation

Energy Technology Data Exchange (ETDEWEB)

Bae, Jun Woo; Kim, Hee Reyoung [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

2016-06-15

The scattered photons cause reduction of the contrast of radiographic image and it results in the degradation of the quality of the image. In order to acquire better quality image, an anti-scattering x-ray gird should be equipped in radiography system. The X-ray anti-scattering grid of the inclined type based on the hybrid concept for that of parallel and focused type was tested by MCNP code. The MCNPX 2.7.0 was used for the simulation based test. The geometry for the test was based on the IEC 60627 which was an international standard for diagnostic X-ray imaging equipment-Characteristics of general purpose and mammographic anti-scatter grids. The performance of grids with four inclined shielding material types was compared with that of the parallel type. The grid with completely tapered type the best performance where there were little performance difference according to the degree of inclination.

A shielding chamber for the Rossendorf whole body counter; Eine Abschirmkammer fuer den Rossendorfer Ganzkoerperzaehler

Energy Technology Data Exchange (ETDEWEB)

Beutmann, A.; Ebert, S.; Kaden, M.; Loehnert, D. [VKTA, Dresden (Germany); Doerfel, H.R. [IDEA System GmbH, Karlsruhe (Germany); Schreiber, W. [Ingenieurbedarf G. Schoene und W. Schreiber GmbH, Grosserkmannsdorf (Germany); Helbig, S.

2016-07-01

In connection with the relocation of the incorporation measurement point operated by the VKTA, a new shielding chamber was designed. The development of the new shielding chamber will be shown based on the design study by IDEA System and the inquiries for material availability, procurement of material and assembly technology up to fabrication, assembly and completion of the chamber. The accompanying background measurements through In-situ gamma spectrometry and first experiences with incorporation measurements at the new In-Vivo measurement facility are shown.

Dependences of microstructure on electromagnetic interference shielding properties of nano-layered Ti3AlC2 ceramics.

Science.gov (United States)

Tan, Yongqiang; Luo, Heng; Zhou, Xiaosong; Peng, Shuming; Zhang, Haibin

2018-05-21

The microstructure dependent electromagnetic interference (EMI) shielding properties of nano-layered Ti 3 AlC 2 ceramics were presented in this study by comparing the shielding properties of various Ti 3 AlC 2 ceramics with distinct microstructures. Results indicate that Ti 3 AlC 2 ceramics with dense microstructure and coarse grains are more favourable for superior EMI shielding efficiency. High EMI shielding effectiveness over 40 dB at the whole Ku-band frequency range was achieved in Ti 3 AlC 2 ceramics by microstructure optimization, and the high shielding effectiveness were well maintained up to 600 °C. A further investigation reveals that only the absorption loss displays variations upon modifying microstructure by allowing more extensive multiple reflections in coarse layered grains. Moreover, the absorption loss of Ti 3 AlC 2 was found to be much higher than those of highly conductive TiC ceramics without layered structure. These results demonstrate that nano-layered MAX phase ceramics are promising candidates of high-temperature structural EMI shielding materials and provide insightful suggestions for achieving high EMI shielding efficiency in other ceramic-based shielding materials.

Monte-Carlo simulations of neutron shielding for the ATLAS forward region

CERN Document Server

Stekl, I; Kovalenko, V E; Vorobel, V; Leroy, C; Piquemal, F; Eschbach, R; Marquet, C

2000-01-01

The effectiveness of different types of neutron shielding for the ATLAS forward region has been studied by means of Monte-Carlo simulations and compared with the results of an experiment performed at the CERN PS. The simulation code is based on GEANT, FLUKA, MICAP and GAMLIB. GAMLIB is a new library including processes with gamma-rays produced in (n, gamma), (n, n'gamma) neutron reactions and is interfaced to the MICAP code. The effectiveness of different types of shielding against neutrons and gamma-rays, composed from different types of material, such as pure polyethylene, borated polyethylene, lithium-filled polyethylene, lead and iron, were compared. The results from Monte-Carlo simulations were compared to the results obtained from the experiment. The simulation results reproduce the experimental data well. This agreement supports the correctness of the simulation code used to describe the generation, spreading and absorption of neutrons (up to thermal energies) and gamma-rays in the shielding materials....

The properties of neutron shielding and flame retardant of EVA polymer after modified by EB accelerator

Science.gov (United States)

Wang, Guo-hui; He, Man-li; Jiang, Dan-feng; He, Fan; Chang, Shu-quan; Dai, Yao-dong

2017-11-01

According to the requirements for neutron shielding and flame retardant properties of some nuclear devices, a new kind of polymer composite materials based on ethylene and vinyl acetate (EVA) polymer have been studied. EVA is the copolymer of ethylene and vinyl acetate, It can be used as materials for applications due to its flexibility, good processability, and low cost. Insulating EVA can be used for cable sheath, automotive sound damping and many other appication. Boron nitride (BN), zinc borate (ZB), magnesium hydroxide (MH) and EVA consisted the compounds with the properties of neutron shielding and flame retardant. With increasing of the contents of BN and ZB, the neutron shielding performance of materials increased up to 33.08%. With the increasing contents of MH and ZB as flame retardant, oxygen index of material have been improved. The elongation at break and tensile strength of material decreased with the increasing of filler powders. Sheet E was chosen and modified by electron beam accelerator in different doses. After modification by electron beam irradiation the sheets showed varying degrees of transformation in the OI, neutron shielding rate and mechanical properties.

Prediction of melanoma metastasis by the Shields index based on lymphatic vessel density

Directory of Open Access Journals (Sweden)

Metcalfe Chris

2010-05-01

Full Text Available Abstract Background Melanoma usually presents as an initial skin lesion without evidence of metastasis. A significant proportion of patients develop subsequent local, regional or distant metastasis, sometimes many years after the initial lesion was removed. The current most effective staging method to identify early regional metastasis is sentinel lymph node biopsy (SLNB, which is invasive, not without morbidity and, while improving staging, may not improve overall survival. Lymphatic density, Breslow's thickness and the presence or absence of lymphatic invasion combined has been proposed to be a prognostic index of metastasis, by Shields et al in a patient group. Methods Here we undertook a retrospective analysis of 102 malignant melanomas from patients with more than five years follow-up to evaluate the Shields' index and compare with existing indicators. Results The Shields' index accurately predicted outcome in 90% of patients with metastases and 84% without metastases. For these, the Shields index was more predictive than thickness or lymphatic density. Alternate lymphatic measurement (hot spot analysis was also effective when combined into the Shields index in a cohort of 24 patients. Conclusions These results show the Shields index, a non-invasive analysis based on immunohistochemistry of lymphatics surrounding primary lesions that can accurately predict outcome, is a simple, useful prognostic tool in malignant melanoma.

Nuclear data for radiation shielding

International Nuclear Information System (INIS)

Miyasaka, Shunichi; Takahashi, Hiroshi.

1976-01-01

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

Shielding in electron beams used in radiotherapy

International Nuclear Information System (INIS)

Sentenac, Irenee.

1979-01-01

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

Neutron data error estimate of criticality calculations for lattice in shielding containers with metal fissionable materials

International Nuclear Information System (INIS)

Vasil'ev, A.P.; Krepkij, A.S.; Lukin, A.V.; Mikhal'kova, A.G.; Orlov, A.I.; Perezhogin, V.D.; Samojlova, L.Yu.; Sokolov, Yu.A.; Terekhin, V.A.; Chernukhin, Yu.I.

1991-01-01

Critical mass experiments were performed using assemblies which simulated one-dimensional lattice consisting of shielding containers with metal fissile materials. Calculations of the criticality of the above assemblies were carried out using the KLAN program with the BAS neutron constants. Errors in the calculations of the criticality for one-, two-, and three-dimensional lattices are estimated. 3 refs.; 1 tab

Self shielding in cylindrical fissile sources in the APNea system

International Nuclear Information System (INIS)

Hensley, D.

1997-01-01

In order for a source of fissile material to be useful as a calibration instrument, it is necessary to know not only how much fissile material is in the source but also what the effective fissile content is. Because uranium and plutonium absorb thermal neutrons so Efficiently, material in the center of a sample is shielded from the external thermal flux by the surface layers of the material. Differential dieaway measurements in the APNea System of five different sets of cylindrical fissile sources show the various self shielding effects that are routinely encountered. A method for calculating the self shielding effect is presented and its predictions are compared with the experimental results

System for detecting and processing abnormality in electromagnetic shielding

International Nuclear Information System (INIS)

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

1991-01-01

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

A shielding design for an accelerator-based neutron source for boron neutron capture therapy

Energy Technology Data Exchange (ETDEWEB)

Hawk, A.E.; Blue, T.E. E-mail: blue.1@osu.edu; Woollard, J.E

2004-11-01

Research in boron neutron capture therapy (BNCT) at The Ohio State University Nuclear Engineering Department has been primarily focused on delivering a high quality neutron field for use in BNCT using an accelerator-based neutron source (ABNS). An ABNS for BNCT is composed of a proton accelerator, a high-energy beam transport system, a {sup 7}Li target, a target heat removal system (HRS), a moderator assembly, and a treatment room. The intent of this paper is to demonstrate the advantages of a shielded moderator assembly design, in terms of material requirements necessary to adequately protect radiation personnel located outside a treatment room for BNCT, over an unshielded moderator assembly design.

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.

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

Pre-installation empirical testing of room shielding for high dose rate remote afterloaders

International Nuclear Information System (INIS)

Klein, E.E.; Grigsby, P.W.; Williamson, J.F.; Meigooni, A.S.

1993-01-01

PURPOSE: Many facilities are acquiring high dose rate remote afterloading units. It is economical that these units be placed in existing shielded teletherapy rooms. Scatter-radiation barriers marginally protect uncontrolled areas from a high dose rate source especially in a room that houses a non-dynamic Cobalt-60 unit. In addition the exact thickness and material composition of the barriers are unknown and therefore, a calculation technique may give misleading results. Also, it would be impossible to evaluate an entire wall barrier by taking isolated core samples in order to assist in the calculations. A quick and inexpensive measurement of dose equivalent using a rented high activity 192Ir source evaluates the barriers and locates shielding deficiencies. METHODS AND MATERIALS: We performed transmission calculations for primary and scattered radiation based on National Council on Radiation Protection and Measurements Reports 49 and 51, respectively. We then rented a high activity 21.7 Ci (8.03 x 10(11) Bq) Ir-192 source to assess our existing teletherapy room shielding for adequacy and voids. This source was placed at the proposed location for clinical high dose rate treatment and measurements were performed. RESULTS: No deficiencies were found in controlled areas surrounding the room, but large differences were found between the calculated and measured values. Our survey located a region in the uncontrolled area above the room requiring augmented shielding which was not predicted by the calculations. A canopy shield was designed to potentially augment the shielding in the ceiling direction. CONCLUSION: Pre-installation testing by measurement is an invaluable method for locating shielding deficiencies and avoiding unnecessary enhancement of shielding particularly when there is lack of information of the inherent shielding

Penetration portion shielding structure

International Nuclear Information System (INIS)

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

1994-01-01

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

Shielding modification design of the N.S. Mutsu

International Nuclear Information System (INIS)

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

1983-01-01

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

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

Superhydrophilic poly (styrene co acrylonitrile)-ZnO nanocomposite surfaces for UV shielding and self-cleaning applications

Science.gov (United States)

Singh, Rajender; Sharma, Ramesh; Barman, P. B.; Sharma, Dheeraj

2017-11-01

UV shielding based super hydrophilic material is developed in the present formulation by in situ emulsion polymerization of poly (styrene-acrylonitrile) with ZnO nanoparticles. The ESI-MS technique confirms the structure of polymer nanocomposite by their mass fragments. The XRD study confirms the presence of ZnO phase in polymer matrix. PSAN/ZnO nanocomposite leads to give effective UV shielding (upto 375 nm) and visible luminescence with ZnO content in polymer matrix. The FESEM and TEM studies confirm the symmetrical, controlled growth of PNs. The incorporation of ZnO nanofillers into PSAN matrix lead to restructuring the PNs surfaces into superhydrophilic surfaces in water contact angle (WCA) from 70° to 10°. We believe our synthesized PSAN/ZnO nanocomposite could be potential as UV shielding, luminescent and super hydrophilic nature based materials in related commercial applications.

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

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.

Dosimetry and shielding

International Nuclear Information System (INIS)

Farinelli, U.

1977-01-01

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

An investigation of safety aspects of operating the end-shields in a brittle condition

International Nuclear Information System (INIS)

Seth, V.K.; Patwardhan, V.M.

1975-01-01

Published data on radiation embrittlement of 3.5% Ni steels (material for RAPP-1, RAPP-2 and MAPP-1 end shields - with charpy V notch value of 2.074 gm at -101 0 C) indicates that the nil ductility transition temperature rise would be of the order of 205 0 C to 260 0 C at the end of 30 year reactor life, against earlier figure of around 120 0 C. Surveillance programme on radiation embrittlement of the end-shields is being conducted to get an idea of the actual condition of the material at any required time. A study has been made to investigate safety aspects of operating the end shields in 'Brittle condition' of the material under the presently designed operating conditions. This study is based on the concept of crack arrest approach (employing fracture analysis diagram; FAD and linear elastic fracture mechanics (using possible correlation between Ksub(Ic) and CVN values). (author)

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

Science.gov (United States)

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

2016-09-01

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

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

Normalization of shielding structure quality and the method of its studying

International Nuclear Information System (INIS)

Bychkov, Ya.A.; Lavdanskij, P.A.

1987-01-01

Method for evaluation of nuclear facility radiation shield quality is suggested. Indexes of shielding structure radiation efficiency and face efficiency are used as the shielding structure quality indexes. The first index is connected with radiation dose rate during personnel irradiation behind the shield, and the second one - with the stresses in shielding structure introduction of the indexes presented allows to evaluate objectively the quality of nuclear facility shielding structure quality design construction and operation and to economize labour and material resources

Shielding tests for a new ship for the transport of spent nuclear fuels

International Nuclear Information System (INIS)

Ito, D.; Kitano, T.; Akiyama, H.; Ueki, K.; Sanui, T.

1998-01-01

a new ship for the transport of spent nuclear fuels which uses serpentine concrete as its major shielding material has been constructed. The shielding calculations are based on DOT3.5 code (CCC-276) and the DLC23). Experiments with Cf-252 and Co-60 sources were carried out to confirm the validity of this method of calculating the shielding effectiveness of serpentine concrete. In these experiments, neutron and gamma-ray dose equivalent rates were measured in various arrangements to simulate the shielding structures of the ship, the calculations for each arrangement were performed by this shielding calculation method. For both neutron and gamma-rays, the calculation results agreed with the experiments very well, confirming that this calculation method used in the ship's shielding design is valid. Two kinds of on-board gamma-ray shielding tests were performed to confirm the ship's actual shielding effectiveness. In one kind of test, gamma-ray dose equivalent rates were measured for each shielding wall using Co-60 sources. In the other kind of test, gamma-ray dose equivalent rates in the ship's accommodation area were measured when a strong Co-60 source was placed in a loaded shipping cask's position. In both gamma-ray shielding tests all measured dose equivalent rates were less than the calculated values, confirming that the ship's actual shielding is sufficient to meet safety requirements. (authors)

Transmission of broad W/Rh and W/Al (target/filter) x-ray beams operated at 25-49 kVp through common shielding materials.

Science.gov (United States)

Li, Xinhua; Zhang, Da; Liu, Bob

2012-07-01

To provide transmission data for broad 25-39 kVp (kilovolt peak) W/Rh and 25-49 kVp W/Al (target/filter, W-tungsten, Rh-rhodium, and Al-aluminum) x-ray beams through common shielding materials, such as lead, concrete, gypsum wallboard, wood, steel, and plate glass. The unfiltered W-target x-ray spectra measured on a Selenia Dimensions system (Hologic Inc., Bedford, MA) set at 20-49 kVp were, respectively, filtered using 50-μm Rh and 700-μm Al, and were subsequently used for Monte Carlo calculations. The transmission of broad x-ray beams through shielding materials was simulated using Geant4 low energy electromagnetic physics package with photon- and electron-processes above 250 eV, including photoelectric effect, Compton scattering, and Rayleigh scattering. The calculated transmission data were fitted using Archer equation with a robust fitting algorithm. The transmission of broad x-ray beams through the above-mentioned shielding materials was calculated down to about 10(-5) for 25-39 kVp W/Rh and 25-49 kVp W/Al. The fitted results of α, β, and γ in Archer equation were provided. The α values of kVp ≥ 40 were approximately consistent with those of NCRP Report No. 147. These data provide inputs for the shielding designs of x-ray imaging facilities with W-anode x-ray beams, such as from Selenia Dimensions.

MFTF-α+T end cell vacuum vessel and nuclear shield trade studies

International Nuclear Information System (INIS)

Kirchner, J.

1984-01-01

Three separate and distinct vacuum vessel and nuclear shield trade studies were performed in series. The studies are: vacuum topology, nuclear shield location and composition, and water bulk shield location and material selection

Electromagnetic interference shielding effectiveness of polypropylene/conducting fiber composites

International Nuclear Information System (INIS)

Lee, Pyoung-Chan; Kim, Bo-Ram; Jeoung, Sun Kyoung; Kim, Yeung Keun

2016-01-01

Electromagnetic released from the automotive electronic parts is harmful to human body. Electromagnetic interference (EMT) shielding refers to the reflection and/or adsorption of electromagnetic radiation by a material, which thereby acts as a shield against the penetration of the radiation through the shield. Polypropylene (PP)/conductive micro fiber composites containing various fiber contents and fiber length were injection-molded. The effect of fiber content and length on electrical properties of the composites was studied by electrical resistivity and EMT shielding measurements. The through-plane electrical conductivity and dielectric permittivity were obtained by measuring dielectric properties. The EMT shielding effectiveness (SE) was investigated by using S-parameter in the range of 100 ~ 1500 MHz. Reflection, absorption and multiple-reflection are the EMT attenuation mechanisms. From the measurement of S-Parameters, the absorption coefficient, reflection coefficient, and the shielding efficiency of the materials were calculated. The EMT SE of PP/conducing fiber composites is 40 dB over a wide frequency range up to 1.5 GHz, which is higher than that of PP/talc composite used automotive parts, viz. 0 dB.

7 CFR 1755.406 - Shield or armor ground resistance measurements.

Science.gov (United States)

2010-01-01

...) The insulation resistance test set should have an output voltage not to exceed 500 volts dc and may be... 7 Agriculture 11 2010-01-01 2010-01-01 false Shield or armor ground resistance measurements. 1755... MATERIALS, AND STANDARD CONTRACT FORMS § 1755.406 Shield or armor ground resistance measurements. (a) Shield...

Shielded container

International Nuclear Information System (INIS)

Fries, B.A.

1978-01-01

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

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

Optimization of the National Ignition Facility primary shield design

International Nuclear Information System (INIS)

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

1993-10-01

Minimum cost design concepts of the primary shield for the National Ignition laser fusion experimental Facility (NIF) are searched with the help of the optimization code SWAN. The computational method developed for this search involves incorporating the time dependence of the delayed photon field within effective delayed photon production cross sections. This method enables one to address the time-dependent problem using relatively simple, time-independent transport calculations, thus significantly simplifying the design process. A novel approach was used for the identification of the optimal combination of constituents that will minimize the shield cost; it involves the generation, with SWAN, of effectiveness functions for replacing materials on an equal cost basis. The minimum cost shield design concept was found to consist of a mixture of polyethylene and low cost, low activation materials such as SiC, with boron added near the shield boundaries

Biological shield design for a 10 MeV Rhodotron

International Nuclear Information System (INIS)

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

2012-01-01

Highlights: ► We evaluate the produced radiations of the Rhodotron-TT200 and their attenuation to the permitted level. ► We apply analytical calculations to determine the shield material and thickness. ► We simulate the Rhodotron accelerator and its shielding using MCNPX code to make sure of results accuracy. -- Abstract: Radiation field of the Rhodotron-TT200 electron accelerator is determined in this study. Regarding the interactions of electron with matter, the produced radiations and their attenuation to the permitted level (i.e. 0.01 mrem/h) are evaluated and calculated. For this purpose analytical calculations are applied to determine the biological shield material and thickness. In order to make sure of results accuracy, Rhodotron accelerator and its shielding are simulated using MCNPX code and the results of analytical calculations and MCNPX code are compared with the experimental ones.

Nanostructured composite layers for electromagnetic shielding in the GHz frequency range

Energy Technology Data Exchange (ETDEWEB)

Suchea, M. [Center of Materials Technology and Photonics, School of Engineering, Technological Educational Institute of Crete, Heraklion (Greece); Chemistry and Physics, “Al.I. Cuza” University of Iasi, Iasi (Romania); Tudose, I.V. [Chemistry and Physics, “Al.I. Cuza” University of Iasi, Iasi (Romania); Center of Materials Technology and Photonics, School of Engineering, Technological Educational Institute of Crete, Heraklion (Greece); Tzagkarakis, G. [Center of Materials Technology and Photonics, School of Engineering, Technological Educational Institute of Crete, Heraklion (Greece); Electrical Engineering Department, School of Engineering, Technological Educational Institute of Crete, Heraklion (Greece); Kenanakis, G. [Center of Materials Technology and Photonics, School of Engineering, Technological Educational Institute of Crete, Heraklion (Greece); Institute of Electronic Structure & Laser (IESL), Foundation for Research and Technology (FORTH) Hellas, Heraklion (Greece); Katharakis, M. [Center of Materials Technology and Photonics, School of Engineering, Technological Educational Institute of Crete, Heraklion (Greece); Drakakis, E. [Electrical Engineering Department, School of Engineering, Technological Educational Institute of Crete, Heraklion (Greece); Koudoumas, E., E-mail: koudoumas@staff.teicrete.gr [Center of Materials Technology and Photonics, School of Engineering, Technological Educational Institute of Crete, Heraklion (Greece); Electrical Engineering Department, School of Engineering, Technological Educational Institute of Crete, Heraklion (Greece)

2015-10-15

Graphical abstract: - Highlights: • Paint-like nanocomposite layers consisting of graphene nanoplatelets, PANI:HCl and PEDOT:PSS present very effective attenuation of electromagnetic radiation in the frequency range 4–20 GHz. • The shielding performance is based mostly on the graphene nanoplatelets and supported by PANI:HCl. In contrast, PEDOT:PSS plays mainly the role of the binder. • Increasing resistivity was observed to reduce the shielding effect, while increasing thickness to favor it. - Abstract: We report on preliminary results regarding the applicability of nanostructured composite layers for electromagnetic shielding in the frequency range of 4–20 GHz. Various combinations of materials were employed including poly(3,4-ethylenedioxythiophene)-poly(styrenesulfonate) (PEDOT:PSS), polyaniline, graphene nanoplatelets, carbon nanotubes, Cu nanoparticles and Poly(vinyl alcohol). As shown, paint-like nanocomposite layers consisting of graphene nanoplatelets, polyaniline PEDOT:PSS and Poly(vinyl alcohol) can offer quite effective electromagnetic shielding, similar or even better than that of commercial products, the response strongly depending on their thickness and resistivity.

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...

Nanostructured composite layers for electromagnetic shielding in the GHz frequency range

International Nuclear Information System (INIS)

Suchea, M.; Tudose, I.V.; Tzagkarakis, G.; Kenanakis, G.; Katharakis, M.; Drakakis, E.; Koudoumas, E.

2015-01-01

Graphical abstract: - Highlights: • Paint-like nanocomposite layers consisting of graphene nanoplatelets, PANI:HCl and PEDOT:PSS present very effective attenuation of electromagnetic radiation in the frequency range 4–20 GHz. • The shielding performance is based mostly on the graphene nanoplatelets and supported by PANI:HCl. In contrast, PEDOT:PSS plays mainly the role of the binder. • Increasing resistivity was observed to reduce the shielding effect, while increasing thickness to favor it. - Abstract: We report on preliminary results regarding the applicability of nanostructured composite layers for electromagnetic shielding in the frequency range of 4–20 GHz. Various combinations of materials were employed including poly(3,4-ethylenedioxythiophene)-poly(styrenesulfonate) (PEDOT:PSS), polyaniline, graphene nanoplatelets, carbon nanotubes, Cu nanoparticles and Poly(vinyl alcohol). As shown, paint-like nanocomposite layers consisting of graphene nanoplatelets, polyaniline PEDOT:PSS and Poly(vinyl alcohol) can offer quite effective electromagnetic shielding, similar or even better than that of commercial products, the response strongly depending on their thickness and resistivity.

Characterization of serpentine. A potential nuclear shielding material

International Nuclear Information System (INIS)

Sengupta, A.; Rajeswari, B.; Kadam, R.M.; Kshirsagar, R.J.

2012-01-01

The use of serpentine as a potential nuclear shielding material necessitates a chemical quality control of the samples before its use in reactors. With this in view, characterization of these mineral samples was carried out using inductively coupled plasma atomic emission spectrometry (ICP-AES) and Instrumental neutron activation analysis (INAA) methods. The analytical results obtained by both ICP-AES and NAA techniques were found to be comparable. Na, Cr, Co, Zn, and Cu were found to be present in all samples of Indian origin while Ga, Ag, Ni, and Cd were found to below the limits of detection. A comparison on the detection limits of elements of interest was also carried out by both the analytical techniques and found to be in good agreement. An infrared spectroscopic investigation was also carried out on all the mineral samples. Bands at 3,689 and 3,648 cm -1 were attributed to inner and outer hydroxyl stretching of Mg-OH, respectively. The weak and broad band centered around 3,416 cm -1 was assigned due to the stretching vibrations of the adsorbed water molecules while three bands at 1076, 1022 and 968 cm -1 were prescribed to the vibrations of the SiO 4 tetrahedra. (author)

A study on the shielding element using Monte Carlo simulation

Energy Technology Data Exchange (ETDEWEB)

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

2017-06-15

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

Production of a datolite-based heavy concrete for shielding nuclear reactors and megavoltage radiotherapy rooms

International Nuclear Information System (INIS)

Mortazavi, S. M. J.; Mosleh-Shirazi, M.A.; Baradaran-Ghahfarokhi, M.; Siavashpour, Z.; Farshadi, A.; Ghafoori, M.; Shahvar, A.

2010-01-01

Biological shielding of nuclear reactors has always been a great concern and decreasing the complexity and expense of these installations is of great interest. In this study, we used datolite and galena minerals for production of a high performance heavy concrete. Materials and Methods: Datolite and galena minerals which can be found in many parts of Iran were used in the concrete mix design. To measure the gamma radiation attenuation of the Datolite and galena concrete samples, they were exposed to both narrow and wide beams of gamma rays emitted from a cobalt-60 radiotherapy unit. An Am-Be neutron source was used for assessing the shielding properties of the samples against neutrons. To test the compression strengths, both types of concrete mixes (Datolite and galena and ordinary concrete) were investigated. Results: The concrete samples had a density of 4420-4650 kg/m 3 compared to that of ordinary concrete (2300-2500 kg/m 3 ) or barite high density concrete (up to 3500 kg/m 3 ). The measured half value layer thickness of the Datolite and galena concrete samples for cobalt-60 gamma rays was much less than that of ordinary concrete (2.56 cm compared to 6.0 cm). Furthermore, the galena concrete samples had a significantly higher compressive strength as well as 20% more neutron absorption. Conclusion: The Datolite and galena concrete samples showed good shielding/engineering properties in comparison with other reported samples made, using high-density materials other than depleted uranium. It is also more economic than the high-density concretes. Datolite and galena concrete may be a suitable option for shielding nuclear reactors and megavoltage radiotherapy rooms.

Shielding practice

International Nuclear Information System (INIS)

Sauermann, P.F.

1985-08-01

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

Study of ceramic mixed boron element as a neutron shielding

International Nuclear Information System (INIS)

Ismail Mustapha; Mohd Reusmaazran Yusof; Md Fakarudin Ab Rahman; Nor Paiza Mohamad Hasan; Samihah Mustaffha; Yusof Abdullah; Mohamad Rabaie Shari; Airwan Affandi Mahmood; Nurliyana Abdullah; Hearie Hassan

2012-01-01

Shielding upon radiation should not be underestimated as it can causes hazard to health. Precautions on the released of radioactive materials should be well concerned and considered. Therefore, the combination of ceramic and boron make them very useful for shielding purpose in areas of low and intermediate neutron. A six grades of ceramic tile have been produced namely IMN05 - 5 % boron, IMN06 - 6 % boron, IMN07 - 7 % boron, IMN08 - 8 % boron, IMN09 - 9 % boron, IMN10 - 10 % boron from mixing, press and sintered process. Boron is a material that capable of absorbing and capturing neutron, so that neutron and gamma test were conducted to analyze the effectiveness of boron material in combination with ceramic as shielding. From the finding, percent reduction number of count per minute shows the ceramic tiles are capable to capture neutron. Apart from all the percentage of boron used, 10 % is the most effective shields since the percent reduction indicating greater neutron captured increased. (author)

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

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.

Spectral perturbations from silicon diode detector encapsulation and shielding in photon fields.

Science.gov (United States)

Eklund, Karin; Ahnesjö, Anders

2010-11-01

Silicon diodes are widely used as detectors for relative dose measurements in radiotherapy. The common manufacturing practice is to encapsulate the diodes in plastic for protection and to facilitate mounting in scanning devices. Diodes intended for use in photon fields commonly also have a shield of a high atomic number material (usually tungsten) integrated into the encapsulation to selectively absorb low-energy photons to which silicon diodes would otherwise over-response. However, new response models based on cavity theories and spectra calculations have been proposed for direct correction of the readout from unshielded (e.g., "electron") diodes used in photon fields. This raises the question whether it is correct to assume that the spectrum in a water phantom at the location of the detector cavity is not perturbed by the detector encapsulation materials. The aim of this work is to investigate the spectral effects of typical encapsulations, including shielding, used for clinical diodes. The effects of detector encapsulation of an unshielded and a shielded commercial diode on the spectra at the detector cavity location are studied through Monte Carlo simulations with PENELOPE-2005. Variance reduction based on correlated sampling is applied to reduce the CPU time needed for the simulations. The use of correlated sampling is found to be efficient and to not introduce any significant bias to the results. Compared to reference spectra calculated in water, the encapsulation for an unshielded diode is demonstrated to not perturb the spectrum, while a tungsten shielded diode caused not only the desired decrease in low-energy scattered photons but also a large increase of the primary electron fluence. Measurements with a shielded diode in a 6 MV photon beam proved that the shielding does not completely remove the field-size dependence of the detector response caused by the over-response from low-energy photons. Response factors of a properly corrected unshielded diode

Evaluation of Shielding Wall Optimization in Lead Slowing Down Spectrometer System

Energy Technology Data Exchange (ETDEWEB)

Jeon, Ju Young; Kim, Jeong Dong; Lee, Yong Deok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

A Lead Slowing Down Spectrometer (LSDS) system is nondestructive technology for analyzing isotope fissile content in spent fuel and pyro processed material, in real time and directly. The high intensity neutron and gamma ray were generated from a nuclear material (Pyro, Spent nuclear fuel), electron beam-target reaction and fission of fissile material. Therefore, shielding analysis of LSDS system should be carried out. In this study, Borax, B{sub 4}C, Li{sub 2}Co{sub 3}, Resin were chosen for shielding analysis. The radiation dose limit (<0.1 μSv/hr) was adopted conservatively at the outer wall surface. The covering could be able to reduce the concrete wall thickness from 5cm to 15cm. The optimized shielding walls evaluation will be used as an important data for future real LSDS facility design and shielding door assessment.

Neutron shielding characteristics of nano-B2O3 dispersed Poly Vinyl Alcohol

International Nuclear Information System (INIS)

Kim, Jae Woo; Uhm, Young Rang; Lee, Min Ku; Lee, Hee Min; Rhee, Chang Kyu

2008-01-01

Neutron is sometimes beneficiary to human beings while they are unwanted for most cases same as the other radiations such as gamma, beta, and alpha, etc. do. Shielding for neutrons therefore is extremely important to keep the radiation environment safe. Especially, it is critical to absorb (or shield) neutrons generated from the spent fuel in a container/storage, nuclear reactor, and cyclotron, etc. In this regard, light materials containing neutron absorbers such as borated-polymers are very useful to shield neutrons in those radiation environments. This investigation is focused on the development of borated polymer-based materials whose neutron shielding efficiency is greatly enhanced by using nano sized boron compounds. Boron is well known as a thermal neutron absorber due to its large thermal neutron absorption cross-section (σ th = 760 b, b = 10 -2 - 4 cm 2 ). Although absorption of neutrons in the medium is mainly dependent on the boron atomic weight concentration, we firstly observed the size of boron particles also has an important role in neutron shielding. Mean free path of neutrons colliding with the smaller particles dispersed in the medium might be decreased when it is compared to the larger particles at the same atomic weight concentration. This means that the neutron shielding efficiency of a polymer mixed with the smaller boron compounds is higher than that of a polymer mixed with the larger boron compounds at the same atomic weight boron concentration

Measurements and Monte-Carlo simulations of the particle self-shielding effect of B4C grains in neutron shielding concrete

Science.gov (United States)

DiJulio, D. D.; Cooper-Jensen, C. P.; Llamas-Jansa, I.; Kazi, S.; Bentley, P. M.

2018-06-01

A combined measurement and Monte-Carlo simulation study was carried out in order to characterize the particle self-shielding effect of B4C grains in neutron shielding concrete. Several batches of a specialized neutron shielding concrete, with varying B4C grain sizes, were exposed to a 2 Å neutron beam at the R2D2 test beamline at the Institute for Energy Technology located in Kjeller, Norway. The direct and scattered neutrons were detected with a neutron detector placed behind the concrete blocks and the results were compared to Geant4 simulations. The particle self-shielding effect was included in the Geant4 simulations by calculating effective neutron cross-sections during the Monte-Carlo simulation process. It is shown that this method well reproduces the measured results. Our results show that shielding calculations for low-energy neutrons using such materials would lead to an underestimate of the shielding required for a certain design scenario if the particle self-shielding effect is not included in the calculations.

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.

DEVELOPMENT OF EXTREMELY LOW FREQUENCY PASSIVE SHIELDING APPLICATION USING MAGNETIC AQUEOUS SUBSTRATE

Directory of Open Access Journals (Sweden)

NOOR ASHIKIN MOHD RASHID

2016-04-01

Full Text Available Public concerns on Extremely Low Frequency (ELF Electromagnetic Field (EMF exposure have been elongated since the last few decades. Electrical substations and high tension rooms in commercial buildings were among the contributing factors emanating ELF magnetic fields. This paper discussed various shielding methods conventionally used in mitigating the ELF exposure. Nevertheless, the standard methods were found to be impractical and incapable of meeting currents shielding demands. In response to that, remarkable researches were conducted in effort to invent novel methods which is more convenient and efficient such as magnetic aqueous shielding or paint, textiles and papers shielding. A magnetic aqueous substrate, Manganese Zinc Ferrite was used as shielding material. The magnetic field and flux distribution inside the aqueous magnetic material are evaluated to optimize shielding against ELFEMF exposure, as to mitigate its exposure.

Design report for shielded glove box

International Nuclear Information System (INIS)

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

1999-05-01

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

Safety analysis and lay-out aspects of shieldings against particle radiation at the example of spallation facilities in the megawatt range

International Nuclear Information System (INIS)

Hanslik, R.

2006-08-01

This paper discusses the shielding of particle radiation from high current accelerators, spallation neutron sources and so called ADS-facilities (Accelerator Driven Systems). ADS-facilities are expected to gain importance in the future for transmutation of long-lived isotopes from fission reactors as well as for energy production. In this paper physical properties of the radiation as well as safety relevant requirements and corresponding shielding concepts are discussed. New concepts for the layout and design of such shielding are presented. Focal point of this work will be the fundamental difference between conventional fission reactor shielding and the safety relevant issues of shielding from high-energy radiation. Key point of this paper is the safety assessment of shielding issues of high current accelerators, spallation targets and ADS-blanket systems as well as neutron scattering instruments at spallation neutron sources. Safety relevant shielding requirements are presented and discussed. For the layout and design of the shielding for spallation sources computer base calculations methods are used. A discussion and comparison of the most important methods like semi-empirical, deterministic and stochastic codes are presented. Another key point within the presented paper is the discussion of shielding materials and their shielding efficiency concerning different types of radiation. The use of recycling material, as a cost efficient solution, is discussed. Based on the conducted analysis, flowcharts for a systematic layout and design of adequate shielding for targets and accelerators have been developed and are discussed in this paper. By use of these flowcharts layout and engineering design of future ADS-facilities can be performed. (orig.)

High Kinetic Energy Penetrator Shielding and High Wear Resistance Materials Fabricated with Boron Nitride Nanotubes (BNNTS) and BNNT Polymer Composites

Science.gov (United States)

Kang, Jin Ho (Inventor); Sauti, Godfrey (Inventor); Smith, Michael W. (Inventor); Jordan, Kevin C. (Inventor); Park, Cheol (Inventor); Bryant, Robert George (Inventor); Lowther, Sharon E. (Inventor)

2015-01-01

Boron nitride nanotubes (BNNTs), boron nitride nanoparticles (BNNPs), carbon nanotubes (CNTs), graphites, or combinations, are incorporated into matrices of polymer, ceramic or metals. Fibers, yarns, and woven or nonwoven mats of BNNTs are used as toughening layers in penetration resistant materials to maximize energy absorption and/or high hardness layers to rebound or deform penetrators. They can be also used as reinforcing inclusions combining with other polymer matrices to create composite layers like typical reinforcing fibers such as Kevlar.RTM., Spectra.RTM., ceramics and metals. Enhanced wear resistance and usage time are achieved by adding boron nitride nanomaterials, increasing hardness and toughness. Such materials can be used in high temperature environments since the oxidation temperature of BNNTs exceeds 800.degree. C. in air. Boron nitride based composites are useful as strong structural materials for anti-micrometeorite layers for spacecraft and space suits, ultra strong tethers, protective gear, vehicles, helmets, shields and safety suits/helmets for industry.

Multi-Layered Effects of Fe on EMI Shielding of Sn-Al Hotel Architectural Powder

Directory of Open Access Journals (Sweden)

Hung Fei-Shuo

2016-01-01

Full Text Available No evident effect in shielding efficiency is observed when the electromagnetic wave-absorbing coating materials were applied in single layers because of the dispersing nature of the powder. When increased to two-layer coating, shielding effects were evident at both high and low frequencies, with greater shielding efficiency at low frequencies over high frequencies. It is worth noting that when increased to three-layer coating, as the weight percentage of powdered Fe increased from 5% to 8% , the shielding efficiency of the powdered-Fe composite material was raised to −35 dB This shows that, as the weight percentage gets higher, the powder shows the resonance phenomenon of permeability spectrum, and at high frequencies, the electromagnetic wave shielding efficiency of the composite materials were greatly increased. As the weight percentage of the powered Fe were increased to 8% , we were unable to spread the powder evenly in the epoxy because of the dispersing characteristic in the electromagnetic properties of Fe and the anisotropic and heterogeneous nature of a powered composite material. During production, the powder aggregates often resulted in greater heterogeneity in the materials and consequently, lowered shielding efficiency at 3GHz.

Physical, mechanical and neutron shielding properties of h-BN/Gd2O3/HDPE ternary nanocomposites

Science.gov (United States)

İrim, Ş. Gözde; Wis, Abdulmounem Alchekh; Keskin, M. Aker; Baykara, Oktay; Ozkoc, Guralp; Avcı, Ahmet; Doğru, Mahmut; Karakoç, Mesut

2018-03-01

In order to prepare an effective neutron shielding material, not only neutron but also gamma absorption must be taken into account. In this research, a polymer nanocomposite based novel type of multifunctional neutron shielding material is designed and fabricated. For this purpose, high density polyethylene (HDPE) was compounded with different amounts of hexagonal boron nitride (h-BN) and Gd2O3 nanoparticles having average particle size of 100 nm using melt-compounding technique. The mechanical, thermal and morphological properties of nanocomposites were investigated. As filler content increased, the absorption of both neutron and gamma fluxes increased despite fluctuating neutron absorption curves. Adding h-BN and Gd2O3 nano particles had a significant influence on both neutron and gamma attenuation properties (Σ, cm-1 and μ/ρ, cm-2/g) of ternary shields and they show an enhancement of 200-280%, 14-52% for neutron and gamma radiations, respectively, in shielding performance.

EXPERIMENTAL EVALUATION OF THE THERMAL PERFORMANCE OF A WATER SHIELD FOR A SURFACE POWER REACTOR

International Nuclear Information System (INIS)

Reid, Robert S.; Pearson, J. Bosie; Stewart, Eric T.

2007-01-01

Water based reactor shielding is being investigated for use on initial lunar surface power systems. A water shield may lower overall cost (as compared to development cost for other materials) and simplify operations in the setup and handling. The thermal hydraulic performance of the shield is of significant interest. The mechanism for transferring heat through the shield is natural convection. Natural convection in a 100 kWt lunar surface reactor shield design is evaluated with 2 kW power input to the water in the Water Shield Testbed (WST) at the NASA Marshall Space Flight Center. The experimental data from the WST is used to validate a CFD model. Performance of the water shield on the lunar surface is then predicted with a CFD model anchored to test data. The experiment had a maximum water temperature of 75 C. The CFD model with 1/6-g predicts a maximum water temperature of 88 C with the same heat load and external boundary conditions. This difference in maximum temperature does not greatly affect the structural design of the shield, and demonstrates that it may be possible to use water for a lunar reactor shield.

Experimental Evaluation of the Thermal Performance of a Water Shield for a Surface Power Reactor

International Nuclear Information System (INIS)

Pearson, J. Boise; Stewart, Eric T.; Reid, Robert S.

2007-01-01

Water based reactor shielding is being investigated for use on initial lunar surface power systems. A water shield may lower overall cost (as compared to development cost for other materials) and simplify operations in the setup and handling. The thermal hydraulic performance of the shield is of significant interest. The mechanism for transferring heat through the shield is natural convection. Natural convection in a 100 kWt lunar surface reactor shield design is evaluated with 2 kW power input to the water in the Water Shield Testbed (WST) at the NASA Marshall Space Flight Center. The experimental data from the WST is used to validate a CFD model. Performance of the water shield on the lunar surface is then predicted with a CFD model anchored to test data. The experiment had a maximum water temperature of 75 deg. C. The CFD model with 1/6-g predicts a maximum water temperature of 88 deg. C with the same heat load and external boundary conditions. This difference in maximum temperature does not greatly affect the structural design of the shield, and demonstrates that it may be possible to use water for a lunar reactor shield

EXPERIMENTAL EVALUATION OF THE THERMAL PERFORMANCE OF A WATER SHIELD FOR A SURFACE POWER REACTOR

Energy Technology Data Exchange (ETDEWEB)

REID, ROBERT S. [Los Alamos National Laboratory; PEARSON, J. BOSIE [Los Alamos National Laboratory; STEWART, ERIC T. [Los Alamos National Laboratory

2007-01-16

Water based reactor shielding is being investigated for use on initial lunar surface power systems. A water shield may lower overall cost (as compared to development cost for other materials) and simplify operations in the setup and handling. The thermal hydraulic performance of the shield is of significant interest. The mechanism for transferring heat through the shield is natural convection. Natural convection in a 100 kWt lunar surface reactor shield design is evaluated with 2 kW power input to the water in the Water Shield Testbed (WST) at the NASA Marshall Space Flight Center. The experimental data from the WST is used to validate a CFD model. Performance of the water shield on the lunar surface is then predicted with a CFD model anchored to test data. The experiment had a maximum water temperature of 75 C. The CFD model with 1/6-g predicts a maximum water temperature of 88 C with the same heat load and external boundary conditions. This difference in maximum temperature does not greatly affect the structural design of the shield, and demonstrates that it may be possible to use water for a lunar reactor shield.

Effect of lithological variations of mine roof on chock shield support using numerical modeling technique

Energy Technology Data Exchange (ETDEWEB)

NONE

2006-09-15

Interaction between chock shield supports, the most popular powered supports in Indian longwall mines, and surrounding coal measure strata is analyzed using finite element models. Thickness and material properties of the main roof, the immediate roof and the coal seam are varied to simulate various geological conditions of Indian coal measure strata. Contact/gap elements are inserted in between the main roof and overburden layer to allow strata separation. Nonlinear material properties are applied with plastic corrections based on Drucker-Prager yield criterion. This paper illustrates effects of lithological variations on shield load, abutment stress, yield zone and longwall face convergence.

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.

Magnetic Shielding Design for Coupler of Wireless Electric Vehicle Charging Using Finite Element Analysis

Science.gov (United States)

Zhao, W. N.; Yang, X. J.; Yao, C.; Ma, D. G.; Tang, H. J.

2017-10-01

Inductive power transfer (IPT) is a practical and preferable method for wireless electric vehicle (EV) charging which proved to be safe, convenient and reliable. Due to the air gap between the magnetic coupler, the magnetic field coupling decreases and the magnetic leakage increases significantly compared to traditional transformer, and this may lead to the magnetic flux density around the coupler more than the safety limit for human. So magnetic shielding should be adding to the winding made from litz wire to enhance the magnetic field coupling effect in the working area and reduce magnetic field strength in non-working area. Magnetic shielding can be achieved by adding high-permeability material or high-conductivity material. For high-permeability material its magnetic reluctance is much lower than the surrounding air medium so most of the magnetic line goes through the high-permeability material rather than surrounding air. For high-conductivity material the eddy current in the material can produce reverse magnetic field to achieve magnetic shielding. This paper studies the effect of the two types of shielding material on coupler for wireless EV charging and designs combination shielding made from high-permeability material and high-conductivity material. The investigation of the paper is done with the help of finite element analysis.

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)

Development of a double plasma gun device for investigation of effects of vapor shielding on erosion of PFC materials under ELM-like pulsed plasma bombardment

Science.gov (United States)

Sakuma, I.; Iwamoto, D.; Kitagawa, Y.; Kikuchi, Y.; Fukumoto, N.; Nagata, M.

2012-10-01

It is considered that thermal transient events such as type I edge localized modes (ELMs) could limit the lifetime of plasma-facing components (PFCs) in ITER. We have investigated surface damage of tungsten (W) materials under transient heat and particle loads by using a magnetized coaxial plasma gun (MCPG) device at University of Hyogo. The capacitor bank energy for the plasma discharge is 144 kJ (2.88 mF, 10 kVmax). Surface melting of a W material was clearly observed at the energy density of ˜2 MJ/m2. It is known that surface melting and evaporation during a transient heat load could generate a vapor cloud layer in front of the target material [1]. Then, the subsequent erosion could be reduced by the vapor shielding effect. In this study, we introduce a new experiment using two MCPG devices (MCPG-1, 2) to understand vapor shielding effects of a W surface under ELM-like pulsed plasma bombardment. The capacitor bank energy of MCPG-2 is almost same as that of MCPG-1. The second plasmoid is applied with a variable delay time after the plasmoid produced by MCPG-1. Then, a vapor cloud layer could shield the second plasma load. To verify the vapor shielding effects, surface damage of a W material is investigated by changing the delay time. In the conference, the preliminary experimental results will be shown.[4pt] [1] A. Hassanein et al., J. Nucl. Mater. 390-391, pp. 777-780 (2009).

Monte Carlo simulation using MCNP4B for an optimal shielding design of a 252 Cf source

International Nuclear Information System (INIS)

Silva, Ademir X. da; Crispim, Verginia R.

2001-01-01

This study aim to investigate an optimum shielding design against neutrons and gamma-rays from a source of 252 Cf, using Monte Carlo simulation. The shielding materials studied were: borated polyethylene, borated-lead polyethylene and stainless steel. The Monte Carlo code MCNP, version 4B, was used to design shielding for 252 Cf based neutron irradiator systems. By normalizing the dose equivalent rate values presented to the neutron production rate of the source, the resulting calculations are independents of the intensity of actual 252 Cf source. The results shown what the total dose equivalent rates were reduced significantly by the shielding system optimization. (author)

Material characterisation and preliminary mechanical design for the HL-LHC shielded beam screens operating at cryogenic temperatures

CERN Document Server

Garion, C; Koettig, T; Machiocha, W; Morrone, M

2015-01-01

The High Luminosity LHC project (HL-LHC) aims at increasing the luminosity (rate of collisions) in the Large Hadron Collider (LHC) experiments by a factor of 10 beyond the original design value (from 300 to 3000 fb-1). It relies on new superconducting magnets, installed close to the interaction points, equipped with new beam screen. This component has to ensure the vacuum performance together with shielding the cold mass from physics debris and screening the cold bore cryogenic system from beam induced heating. The beam screen operates in the range 40-60 K whereas the magnet cold bore temperature is 1.9 K. A tungsten-based material is used to absorb the energy of particles. In this paper, measurements of the mechanical and physical properties of such tungsten material are shown at room and cryogenic temperature. In addition, the design and the thermal mechanical behaviour of the beam screen assembly are presented also. They include the heat transfer from the tungsten absorbers to the cooling pipes and the sup...

The assembly of the disk shielding is finished.

CERN Multimedia

Vincent Hedberg

At the end of March, the shielding project engineer, Jan Palla, could draw a sigh of relief when the fourth and final rotation of the disk shielding was carried out without incident. The two 80-ton heavy shielding assemblies were built in a horizontal position and they had to be first turned upside-down and then rotated to a vertical position during the assembly. The relatively thin disk plate with a diameter of 9 meters, made this operation quite delicate and a lot of calculation work and strengthening of the shielding was carried out before the rotations could take place. The disk shielding is being turned upside-down. The stainless steel cylinder in the centre supports the shielding as well as the small muon wheel. The two disk shielding assemblies consist of different materials such as bronze, gray steel, cast iron, stainless steel, boron doped polyethylene and lead. The project is multinational with the major pieces having been made by companies in Armenia, Serbia, Spain, Bulgaria, Italy, Slovaki...

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.