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)
Shieldings for X-ray radiotherapy facilities calculated by computer
International Nuclear Information System (INIS)
Pedrosa, Paulo S.; Farias, Marcos S.; Gavazza, Sergio
2005-01-01
This work presents a methodology for calculation of X-ray shielding in facilities of radiotherapy with help of computer. Even today, in Brazil, the calculation of shielding for X-ray radiotherapy is done based on NCRP-49 recommendation establishing a methodology for calculating required to the elaboration of a project of shielding. With regard to high energies, where is necessary the construction of a labyrinth, the NCRP-49 is not very clear, so that in this field, studies were made resulting in an article that proposes a solution to the problem. It was developed a friendly program in Delphi programming language that, through the manual data entry of a basic design of architecture and some parameters, interprets the geometry and calculates the shields of the walls, ceiling and floor of on X-ray radiation therapy facility. As the final product, this program provides a graphical screen on the computer with all the input data and the calculation of shieldings and the calculation memory. The program can be applied in practical implementation of shielding projects for radiotherapy facilities and can be used in a didactic way compared to NCRP-49.
Computer code for shielding calculations of x-rays rooms
International Nuclear Information System (INIS)
Affonso, R.R.W.; Borges, D. da S.; Lava, D.D.; Moreira, M. de L.; Guimarães, A.C.F.
2015-01-01
The building an effective barrier against ionizing radiation present in radiographic rooms requires consideration of many variables. The methodology used for thickness specification of primary and secondary, barrier of a traditional radiographic room, considers the following factors: Use Factor, Occupational Factor, distance between the source and the wall, Workload, Kerma in the air and distance between the patient and the source. With these data it was possible to develop a computer code, which aims to identify and use variables in functions obtained through graphics regressions provided by NCRP-147 (Structural Shielding Design for Medical X-Ray Imaging Facilities) report, for shielding calculation of room walls, and the walls of the dark room and adjacent areas. With the implemented methodology, it was made a code validation by comparison of results with a study case provided by the report. The obtained values for thickness comprise different materials such as concrete, lead and glass. After validation it was made a case study of an arbitrary radiographic room.The development of the code resulted in a user-friendly tool for planning radiographic rooms to comply with the limits established by CNEN-NN-3:01 published in september/2011. (authors)
Neutron and gamma ray transport calculations in shielding system
Energy Technology Data Exchange (ETDEWEB)
Masukawa, Fumihiro; Sakamoto, Hiroki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
1998-03-01
In the shields for radiation in nuclear facilities, the penetrating holes of various kinds and irregular shapes are made for the reasons of operation, control and others. These penetrating holes and gaps are filled with air or the substances with relatively small shielding performance, and radiation flows out through them, which is called streaming. As the calculation techniques for the shielding design or analysis related to the streaming problem, there are the calculations by simplified evaluation, transport calculation and Monte Carlo method. In this report, the example of calculation by Monte Carlo method which is represented by MCNP code is discussed. A number of variance reduction techniques which seem effective for the analysis of streaming problem were tried. As to the investigation of the applicability of MCNP code to streaming analysis, the object of analysis which are the concrete walls without hole and with horizontal hole, oblique hole and bent oblique hole, the analysis procedure, the composition of concrete, and the conversion coefficient of dose equivalent, and the results of analysis are reported. As for variance reduction technique, cell importance was adopted. (K.I.)
Method for calculating required shielding in medical x-ray rooms
International Nuclear Information System (INIS)
Karppinen, J.
1997-10-01
The new annual radiation dose limits - 20 mSv (previously 50 mSv) for radiation workers and 1 mSv (previously 5 mSv) for other persons - implies that the adequacy of existing radiation shielding must be re-evaluated. In principle, one could assume that the thicknesses of old radiation shields should be increased by about one or two half-value layers in order to comply with the new dose limits. However, the assumptions made in the earlier shielding calculations are highly conservative; the required shielding was often determined by applying the maximum high-voltage of the x-ray tube for the whole workload. A more realistic calculation shows that increased shielding is typically not necessary if more practical x-ray tube voltages are used in the evaluation. We have developed a PC-based calculation method for calculating the x-ray shielding which is more realistic than the highly conservative method formerly used. The method may be used to evaluate an existing shield for compliance with new regulations. As examples of these calculations, typical x-ray rooms are considered. The lead and concrete thickness requirements as a function of x-ray tube voltage and workload are also given in tables. (author)
International Nuclear Information System (INIS)
Lukaszek, W.; Kucypera, S.
1982-01-01
The semianalytic method was used for calculating the attenuation of gamma rays and the thickness of biological shield of graphite moderated reactor. A short description of computer code as well as the exemplary results of calculations are given. (A.S.)
A new approximating formula for calculating gamma-ray buildup factors in multilayer shields
International Nuclear Information System (INIS)
Assad, A.; Chiron, M.; Nimal, J.C.; Diop, C.M.; Ridoux, P.
1999-01-01
This study proposes a new approximating formula for calculating gamma-ray buildup factors in multilayer shields. The formula combines the buildup factors of single-layer shields with products and quotients. The feasibility of the formula for reproducing the buildup factors was tested by using point isotropic buildup factors calculated with the SN1D discrete ordinates code as reference data. The dose buildup factors of single-, double-, and multilayer shields composed of water, aluminum, iron, and lead were calculated for a spherical geometry in the energy range between 10 MeV and 40 keV and for total thicknesses of up to 30 mean free paths. The calculation of the buildup factors takes into account the bound electron effect of Compton scattering (incoherent scattering), the coherent scattering, the pair production, and the secondary sources of bremsstrahlung and fluorescence. The tests have shown that the approximating formula reproduces the reference data of double-layer shields very well for most cases. With the same parameters and with a new physical consideration that takes into account in a global way the degradation of the gamma-ray energy spectrum, the buildup factors of three- and five-layer shields were also very well reproduced
JULIA: calculation projection software for primary barriers shielding to X-Rays using barite
International Nuclear Information System (INIS)
Silva, Júlia R.A.S. da; Vieira, José W.; Lima, Fernando R. A.
2017-01-01
The objective was to program a software to calculate the required thicknesses to attenuate X-rays in kilovoltage of 60 kV, 80 kV, 110 kV and 150 kV. The conventional methodological parameters for structural shield calculations established by the NCRP (National Council on Radiation Protection and Measurements) were presented. The descriptive and exploratory methods allowed the construction of the JULIA. In this sense and based on the result obtained, the tool presented is useful for professionals who wish to design structural shielding in radiodiagnostic and/or therapy. The development of calculations in the computational tool corresponds to the accessibility, optimization of time and estimation close to the real. Such heuristic exercise represents improvement of calculations for the estimation of primary barriers with barite
Development of a computer code for shielding calculation in X-ray facilities
International Nuclear Information System (INIS)
Borges, Diogo da S.; Lava, Deise D.; Affonso, Renato R.W.; Moreira, Maria de L.; Guimaraes, Antonio C.F.
2014-01-01
The construction of an effective barrier against the interaction of ionizing radiation present in X-ray rooms requires consideration of many variables. The methodology used for specifying the thickness of primary and secondary shielding of an traditional X-ray room considers the following factors: factor of use, occupational factor, distance between the source and the wall, workload, Kerma in the air and distance between the patient and the receptor. With these data it was possible the development of a computer program in order to identify and use variables in functions obtained through graphics regressions offered by NCRP Report-147 (Structural Shielding Design for Medical X-Ray Imaging Facilities) for the calculation of shielding of the room walls as well as the wall of the darkroom and adjacent areas. With the built methodology, a program validation is done through comparing results with a base case provided by that report. The thickness of the obtained values comprise various materials such as steel, wood and concrete. After validation is made an application in a real case of radiographic room. His visual construction is done with the help of software used in modeling of indoor and outdoor. The construction of barriers for calculating program resulted in a user-friendly tool for planning radiographic rooms to comply with the limits established by CNEN-NN-3:01 published in September / 2011
Handout on shielding calculation
International Nuclear Information System (INIS)
Heilbron Filho, P.F.L.
1991-01-01
In order to avoid the difficulties of the radioprotection supervisors in the tasks related to shielding calculations, is presented in this paper the basic concepts of shielding theory. It also includes exercises and examples. (author)
International Nuclear Information System (INIS)
Santoro, R.T.; Alsmiller, R.G. Jr.; Barnes, J.M.; Chapman, G.T.
1980-08-01
Integral experiments that measure the transport of approx. 14 MeV D-T neutrons through laminated slabs of proposed fusion reactor shield materials have been carried out. Measured and calculated neutron and gamma ray energy spectra are compared as a function of the thickness and composition of stainless steel type 304, borated polyethylene, and Hevimet (a tungsten alloy), and as a function of detector position behind these materials. The measured data were obtained using a NE-213 liquid scintillator using pulse-shape discrimination methods to resolve neutron and gamma ray pulse height data and spectral unfolding methods to convert these data to energy spectra. The calculated data were obtained using two-dimensional discrete ordinates radiation transport methods in a complex calculational network that takes into account the energy-angle dependence of the D-T neutrons and the nonphysical anomalies of the S/sub n/ method
International Nuclear Information System (INIS)
Elmahroug, Y.; Tellili, B.; Souga, C.; Manai, K.
2015-01-01
Highlights: • Description of the theoretical method used by the ParShield program. • Description of the ParShield program. • Test and validation the ParShield program. - Abstract: This study aims to present a new computer program called ParShield which determines the neutron and gamma-ray shielding parameters. This program can calculate the total mass attenuation coefficients (μ t ), the effective atomic numbers (Z eff ) and the effective electron densities (N eff ) for gamma rays and it can also calculate the effective removal cross-sections (Σ R ) for fast neutrons for mixtures and compounds. The results obtained for the gamma rays by using ParShield were compared with the results calculated by the WinXcom program and the measured results. The obtained values of (Σ R ) were tested by comparing them with the measured results,the manually calculated results and with the results obtained by using MERCSFN program and an excellent agreement was found between them. The ParShield program can be used as a fast and effective tool to choose and compare the shielding materials, especially for the determination of (Z eff ) and (N eff ), there is no other programs in the literature which can calculate
Parameters calculation of shielding experiment
International Nuclear Information System (INIS)
Gavazza, S.
1986-02-01
The radiation transport methodology comparing the calculated reactions and dose rates for neutrons and gama-rays, with experimental measurements obtained on iron shield, irradiated in the YAYOI reactor is evaluated. The ENDF/B-IV and VITAMIN-C libraries and the AMPX-II modular system, for cross sections generation collapsed by the ANISN code were used. The transport calculations were made using the DOT 3.5 code, adjusting the boundary iron shield source spectrum to the reactions and dose rates, measured at the beginning of shield. The neutron and gamma ray distributions calculated on the iron shield presented reasonable agreement with experimental measurements. An experimental arrangement using the IEA-R1 reactor to determine a shielding benchmark is proposed. (Author) [pt
Calculating additional shielding requirements in diagnostics X-ray departments by computer
International Nuclear Information System (INIS)
Rahimi, A.
2004-01-01
This report provides an extension of an existing method for the calculation of the barrier thickness required to reduce the three types of radiation exposure emitted from the source, the primary, secondary and leakage radiation, to a specified weekly design limit (MPD). Since each of these three types of radiation are of different beam quality, having different shielding requirements, NCRP 49 has provided means to calculate the necessary protective barrier thickness for each type of radiation individually. Additionally, barrier requirements specified using the techniques stated at NCRP 49, show enormous variations among users. Part of the variations is due to different assumptions made regarding the use of the examined room and the characteristics of adjoining space. Many of the differences result from the difficulty of accurately relating information from the calculations to graphs and tables involved in the calculation process specified by this report. Moreover, the latest technological developments such as mammography are not addressed and attenuation data for three-phase generators, that are most widely used today, is not provided. The design of shielding barriers in diagnostic X-ray departments generally follow the ALARA principle. That means that, in practice, the exposure levels are kept 'as low as reasonably achievable', taking into account economical and technical factors. Additionally, the calculation of barrier requirements includes many uncertainties (e.g. the workload, the actual kVp used etc.). (author)
Experimental verification of photon: A program for use in x-ray shielding calculations
International Nuclear Information System (INIS)
Brauer, E.; Thomlinson, W.
1987-01-01
At the National Synchrotron Light Source, a computer program named PHOTON has been developed to calculate radiation dose values around a beam line. The output from the program must be an accurate guide to beam line shielding. To test the program, a series of measurements of radiation dose were carried out using existing beam lines; the results were compared to the theoretical calculations of PHOTON. Several different scattering geometries, scattering materials, and sets of walls and shielding materials were studied. Results of the measurements allowed many advances to be made in the program, ultimately resulting in good agreement between the theory and experiment. 3 refs., 6 figs
Sharaf, J. M.; Saleh, H.
2015-05-01
The shielding properties of three different construction styles, and building materials, commonly used in Jordan, were evaluated using parameters such as attenuation coefficients, equivalent atomic number, penetration depth and energy buildup factor. Geometric progression (GP) method was used to calculate gamma-ray energy buildup factors of limestone, concrete, bricks, cement plaster and air for the energy range 0.05-3 MeV, and penetration depths up to 40 mfp. It has been observed that among the examined building materials, limestone offers highest value for equivalent atomic number and linear attenuation coefficient and the lowest values for penetration depth and energy buildup factor. The obtained buildup factors were used as basic data to establish the total equivalent energy buildup factors for three different multilayer construction styles using an iterative method. The three styles were then compared in terms of fractional transmission of photons at different incident photon energies. It is concluded that, in case of any nuclear accident, large multistory buildings with five layers exterior walls, style A, could effectively attenuate radiation more than small dwellings of any construction style.
Shielding calculations using FLUKA
International Nuclear Information System (INIS)
Yamaguchi, Chiri; Tesch, K.; Dinter, H.
1988-06-01
The dose equivalent on the surface of concrete shielding has been calculated using the Monte Carlo code FLUKA86 for incident proton energies from 10 to 800 GeV. The results have been compared with some simple equations. The value of the angular dependent parameter in Moyer's equation has been calculated from the locations where the values of the maximum dose equivalent occur. (author)
International Nuclear Information System (INIS)
Kotegawa, Hiroshi; Tanaka, Shun-ichi
1991-09-01
A point-kernel integral technique code, PKN, and the related data library have been developed to calculate neutron and secondary gamma-ray dose equivalents in water, concrete and iron shields for neutron sources in 3-dimensional geometry. The comparison between calculational results of the present code and those of the 1-dimensional transport code ANISN = JR, and the 2-dimensional transport code DOT4.2 showed a sufficient accuracy, and the availability of the PKN code has been confirmed. (author)
Shielding calculational system for plutonium
International Nuclear Information System (INIS)
Zimmerman, M.G.; Thomsen, D.H.
1975-08-01
A computer calculational system has been developed and assembled specifically for calculating dose rates in AEC plutonium fabrication facilities. The system consists of two computer codes and all nuclear data necessary for calculation of neutron and gamma dose rates from plutonium. The codes include the multigroup version of the Battelle Monte Carlo code for solution of general neutron and gamma shielding problems and the PUSHLD code for solution of shielding problems where low energy gamma and x-rays are important. The nuclear data consists of built in neutron and gamma yields and spectra for various plutonium compounds, an automatic calculation of age effects and all cross-sections commonly used. Experimental correlations have been performed to verify portions of the calculational system. (23 tables, 7 figs, 16 refs) (U.S.)
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
Shielding calculations for NET
International Nuclear Information System (INIS)
Verschuur, K.A.; Hogenbirk, A.
1991-05-01
In the European Fusion Technology Programme there is only a small activity on research and development for fusion neutronics. Never-the-less, looking further than blanket design now, as ECN is getting involved in design of radiation shields for the coils and biological shields, it becomes apparent that fusion neutronics as a whole still needs substantial development. Existing exact codes for calculation of complex geometries like MCNP and DORT/TORT are put over the limits of their numerical capabilities, whilst approximate codes for complex geometries like FURNACE and MERCURE4 are put over the limits of their modelling capabilities. The main objective of this study is just to find out how far we can get with existing codes in obtaining reliable values for the radiation levels inside and outside the cryostat/shield during operation and after shut-down. Starting with a 1D torus model for preliminary parametric studies, more dimensional approximation of the torus or parts of it including the main heterogeneities should follow. Regular contacts with the NET-Team are kept, to be aware of main changes in NET design that might affect our calculation models. Work on the contract started 1 July 1990. The technical description of the contract is given. (author). 14 refs.; 4 figs.; 1 tab
International Nuclear Information System (INIS)
Simpkin, D.J.
1989-01-01
A Monte Carlo calculation has been performed to determine the transmission of broad constant-potential x-ray beams through Pb, concrete, gypsum wallboard, steel and plate glass. The EGS4 code system was used with a simple broad-beam geometric model to generate exposure transmission curves for published 70, 100, 120 and 140-kVcp x-ray spectra. These curves are compared to measured three-phase generated x-ray transmission data in the literature and found to be reasonable. For calculation ease the data are fit to an equation previously shown to describe such curves quite well. These calculated transmission data are then used to create three-phase shielding tables for Pb and concrete, as well as other materials not available in Report No. 49 of the NCRP
Simpkin, D J
1989-02-01
A Monte Carlo calculation has been performed to determine the transmission of broad constant-potential x-ray beams through Pb, concrete, gypsum wallboard, steel and plate glass. The EGS4 code system was used with a simple broad-beam geometric model to generate exposure transmission curves for published 70, 100, 120 and 140-kVcp x-ray spectra. These curves are compared to measured three-phase generated x-ray transmission data in the literature and found to be reasonable. For calculation ease the data are fit to an equation previously shown to describe such curves quite well. These calculated transmission data are then used to create three-phase shielding tables for Pb and concrete, as well as other materials not available in Report No. 49 of the NCRP.
A new shielding calculation method for X-ray computed tomography regarding scattered radiation.
Watanabe, Hiroshi; Noto, Kimiya; Shohji, Tomokazu; Ogawa, Yasuyoshi; Fujibuchi, Toshioh; Yamaguchi, Ichiro; Hiraki, Hitoshi; Kida, Tetsuo; Sasanuma, Kazutoshi; Katsunuma, Yasushi; Nakano, Takurou; Horitsugi, Genki; Hosono, Makoto
2017-06-01
The goal of this study is to develop a more appropriate shielding calculation method for computed tomography (CT) in comparison with the Japanese conventional (JC) method and the National Council on Radiation Protection and Measurements (NCRP)-dose length product (DLP) method. Scattered dose distributions were measured in a CT room with 18 scanners (16 scanners in the case of the JC method) for one week during routine clinical use. The radiation doses were calculated for the same period using the JC and NCRP-DLP methods. The mean (NCRP-DLP-calculated dose)/(measured dose) ratios in each direction ranged from 1.7 ± 0.6 to 55 ± 24 (mean ± standard deviation). The NCRP-DLP method underestimated the dose at 3.4% in fewer shielding directions without the gantry and a subject, and the minimum (NCRP-DLP-calculated dose)/(measured dose) ratio was 0.6. The reduction factors were 0.036 ± 0.014 and 0.24 ± 0.061 for the gantry and couch directions, respectively. The (JC-calculated dose)/(measured dose) ratios ranged from 11 ± 8.7 to 404 ± 340. The air kerma scatter factor κ is expected to be twice as high as that calculated with the NCRP-DLP method and the reduction factors are expected to be 0.1 and 0.4 for the gantry and couch directions, respectively. We, therefore, propose a more appropriate method, the Japanese-DLP method, which resolves the issues of possible underestimation of the scattered radiation and overestimation of the reduction factors in the gantry and couch directions.
Calculation of Buildup Factor for Gamma-ray Exposure in Two Layered Shields Made of Water and Lead
International Nuclear Information System (INIS)
Al-Saadi, A.H.
2012-01-01
The buildup factor for gamma ray exposure is most useful in calculations for biological protective shields.The buildup factors for gamma ray exposure were calculated in tow layered shields consist of water-lead and lead-water up to optical Thickness 20 mean free path (mfp) at gamma ray energies 1, 2 and 6MeV by using kalos's formula.The program has been designed to work at any atomic number of the attenuating medium, photon energy, slab thickness and and the arrangement of materials.The results obtained in this search leading to the buildup factor for gamma ray exposure at energies (1and2MeV) in lead-water were higher than the reverse case,while at energy 6 MeV the effect was opposite.The calculated data were parameterized by an empirical formula as a function of optical thickness of tow materials.The results obtained were in reasonable agreement with a previous work
Energy Technology Data Exchange (ETDEWEB)
Lee, Sang Heon; Lee, Eun Joong; Kim, Chan Kyu; Cho, Gyu Seong [Dept. of Nuclear and Quantum Engineering, KAIST, Daejeon (Korea, Republic of); Hur, Sam Suk [Sam Yong Inspection Engineering Co., Ltd., Seoul (Korea, Republic of)
2016-11-15
Radiation generating devices must be properly shielded for their safe application. Although institutes such as US National Bureau of Standards and National Council on Radiation Protection and Measurements (NCRP) have provided guidelines for shielding X-ray tube of various purposes, industry people tend to rely on 'Half Value Layer (HVL) method' which requires relatively simple calculation compared to the case of those guidelines. The method is based on the fact that the intensity, dose, and air kerma of narrow beam incident on shielding wall decreases by about half as the beam penetrates the HVL thickness of the wall. One can adjust shielding wall thickness to satisfy outside wall dose or air kerma requirements with this calculation. However, this may not always be the case because 1) The strict definition of HVL deals with only Intensity, 2) The situation is different when the beam is not 'narrow'; the beam quality inside the wall is distorted and related changes on outside wall dose or air kerma such as buildup effect occurs. Therefore, sometimes more careful research should be done in order to verify the effect of shielding specific radiation generating device. High energy X-ray tubes which is operated at the voltage above 400 kV that are used for 'heavy' nondestructive inspection is an example. People have less experience in running and shielding such device than in the case of widely-used low energy X-ray tubes operated at the voltage below 300 kV. In this study, Air Kerma value per week, outside concrete shielding wall of various thickness surrounding 450 kVp X-ray tube were calculated using MCNP simulation with the aid of Geometry Splitting method which is a famous Variance Reduction technique. The comparison between simulated result, HVL method result, and NCRP Report 147 safety goal 0.02 mGy wk-1 on Air Kerma for the place where the public are free to pass showed that concrete wall of thickness 80 cm is needed to achieve the
International Nuclear Information System (INIS)
Lee, Sang Heon; Lee, Eun Joong; Kim, Chan Kyu; Cho, Gyu Seong; Hur, Sam Suk
2016-01-01
Radiation generating devices must be properly shielded for their safe application. Although institutes such as US National Bureau of Standards and National Council on Radiation Protection and Measurements (NCRP) have provided guidelines for shielding X-ray tube of various purposes, industry people tend to rely on 'Half Value Layer (HVL) method' which requires relatively simple calculation compared to the case of those guidelines. The method is based on the fact that the intensity, dose, and air kerma of narrow beam incident on shielding wall decreases by about half as the beam penetrates the HVL thickness of the wall. One can adjust shielding wall thickness to satisfy outside wall dose or air kerma requirements with this calculation. However, this may not always be the case because 1) The strict definition of HVL deals with only Intensity, 2) The situation is different when the beam is not 'narrow'; the beam quality inside the wall is distorted and related changes on outside wall dose or air kerma such as buildup effect occurs. Therefore, sometimes more careful research should be done in order to verify the effect of shielding specific radiation generating device. High energy X-ray tubes which is operated at the voltage above 400 kV that are used for 'heavy' nondestructive inspection is an example. People have less experience in running and shielding such device than in the case of widely-used low energy X-ray tubes operated at the voltage below 300 kV. In this study, Air Kerma value per week, outside concrete shielding wall of various thickness surrounding 450 kVp X-ray tube were calculated using MCNP simulation with the aid of Geometry Splitting method which is a famous Variance Reduction technique. The comparison between simulated result, HVL method result, and NCRP Report 147 safety goal 0.02 mGy wk-1 on Air Kerma for the place where the public are free to pass showed that concrete wall of thickness 80 cm is needed to achieve the safety goal
Radiation shielding calculation using MCNP
International Nuclear Information System (INIS)
Masukawa, Fumihiro
2001-01-01
To verify the Monte Carlo code MCNP4A as a tool to generate the reference data in the shielding designs and the safety evaluations, various shielding benchmark experiments were analyzed using this code. These experiments were categorized in three types of the shielding subjects; bulk shielding, streaming, and skyshine. For the variance reduction technique, which is indispensable to get meaningful results with the Monte Carlo shielding calculation, we mainly used the weight window, the energy dependent Russian roulette and spitting. As a whole, our analyses performed enough small statistical errors and showed good agreements with these experiments. (author)
International Nuclear Information System (INIS)
Diop, Cheikh M'Backe.
1980-09-01
The conception of the necessary shielding around a conceptual design Tokamak requires to execute an estimated calculation of the doses due to the different radiation sources arising from the machine: the thermonuclear neutron source and the gamma ray source emitted during the interaction of the runaway electrons with the diaphragm. In this study, we propose a theorical method to calculate this gamma source. We tackle also the shielding problem of the conceptual design Tokamak: TORE 2 SUPRA [fr
Energy Technology Data Exchange (ETDEWEB)
Swadling, G. F.; Ross, J. S.; Datte, P.; Moody, J.; Divol, L.; Jones, O.; Landen, O. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States)
2016-11-15
An Optical Thomson Scattering (OTS) diagnostic is currently being developed for the National Ignition Facility (NIF) at Lawrence Livermore National Laboratory. This diagnostic is designed to make measurements of the hohlraum plasma parameters, such as the electron temperature and the density, during inertial confinement fusion (ICF) experiments. NIF ICF experiments present a very challenging environment for optical measurements; by their very nature, hohlraums produce intense soft x-ray emission, which can cause “blanking” (radiation induced opacity) of the radiation facing optical components. The soft x-ray fluence at the surface of the OTS blast shield, 60 cm from the hohlraum, is estimated to be ∼8 J cm{sup −2}. This is significantly above the expected threshold for the onset of “blanking” effects. A novel xenon plasma x-ray shield is proposed to protect the blast shield from x-rays and mitigate “blanking.” Estimates suggest that an areal density of 10{sup 19} cm{sup −2} Xe atoms will be sufficient to absorb 99.5% of the soft x-ray flux. Two potential designs for this shield are presented.
Swadling, G F; Ross, J S; Datte, P; Moody, J; Divol, L; Jones, O; Landen, O
2016-11-01
An Optical Thomson Scattering (OTS) diagnostic is currently being developed for the National Ignition Facility (NIF) at Lawrence Livermore National Laboratory. This diagnostic is designed to make measurements of the hohlraum plasma parameters, such as the electron temperature and the density, during inertial confinement fusion (ICF) experiments. NIF ICF experiments present a very challenging environment for optical measurements; by their very nature, hohlraums produce intense soft x-ray emission, which can cause "blanking" (radiation induced opacity) of the radiation facing optical components. The soft x-ray fluence at the surface of the OTS blast shield, 60 cm from the hohlraum, is estimated to be ∼8 J cm -2 . This is significantly above the expected threshold for the onset of "blanking" effects. A novel xenon plasma x-ray shield is proposed to protect the blast shield from x-rays and mitigate "blanking." Estimates suggest that an areal density of 10 19 cm -2 Xe atoms will be sufficient to absorb 99.5% of the soft x-ray flux. Two potential designs for this shield are presented.
Shielding calculations. Optimization vs. Paradigms
International Nuclear Information System (INIS)
Cornejo Diaz, Nestor; Hernandez Saiz, Alejandro; Martinez Gonzalez, Alina
2005-01-01
Many radiation shielding barriers in Cuba have been designed according to the criterion of Maxi-mum Projected Dose Rates. This fact has created the paradigm of low dose rates. Because of this, dose rate levels greater than units of Sv.h-1 would be considered unacceptable by many specialists, regardless of the real exposure times. Nowadays many shielding barriers are being designed using dose constraints in real exposure times. Behind the new barriers, dose rates could be notably greater than those behind the traditional ones, and it does not imply inadequate designs or constructive errors. In this work were obtained significant differences in dose rate levels and shield-ing thicknesses calculated by both methods for some typical installations. The work concludes that real exposure time approach is more adequate in order to optimise Radiation Protection, although this method should be carefully applied
Methodology of shielding calculation for nuclear reactors
International Nuclear Information System (INIS)
Maiorino, J.R.; Mendonca, A.G.; Otto, A.C.; Yamaguchi, Mitsuo
1982-01-01
A methodology of calculation that coupling a serie of computer codes in a net that make the possibility to calculate the radiation, neutron and gamma transport, is described, for deep penetration problems, typical of nuclear reactor shielding. This net of calculation begining with the generation of constant multigroups, for neutrons and gamma, by the AMPX system, coupled to ENDF/B-IV data library, the transport calculation of these radiations by ANISN, DOT 3.5 and Morse computer codes, up to the calculation of absorbed doses and/or equivalents buy SPACETRAN code. As examples of the calculation method, results from benchmark n 0 6 of Shielding Benchmark Problems - ORNL - RSIC - 25, namely Neutron and Secondary Gamma Ray fluence transmitted through a Slab of Borated Polyethylene, are presented. (Author) [pt
Methods for calculating radiation attenuation in shields
Energy Technology Data Exchange (ETDEWEB)
Butler, J; Bueneman, D; Etemad, A; Lafore, P; Moncassoli, A M; Penkuhn, H; Shindo, M; Stoces, B
1964-10-01
In recent years the development of high-speed digital computers of large capacity has revolutionized the field of reactor shield design. For compact special-purpose reactor shields, Monte-Carlo codes in two- and three dimensional geometries are now available for the proper treatment of both the neutron and gamma- ray problems. Furthermore, techniques are being developed for the theoretical optimization of minimum-weight shield configurations for this type of reactor system. In the design of land-based power reactors, on the other hand, there is a strong incentive to reduce the capital cost of the plant, and economic considerations are also relevant to reactors designed for merchant ship propulsion. In this context simple methods are needed which are economic in their data input and computing time requirements and which, at the same time, are sufficiently accurate for design work. In general the computing time required for Monte-Carlo calculations in complex geometry is excessive for routine design calculations and the capacity of the present codes is inadequate for the proper treatment of large reactor shield systems in three dimensions. In these circumstances a wide range of simpler techniques are currently being employed for design calculations. The methods of calculation for neutrons in reactor shields fall naturally into four categories: Multigroup diffusion theory; Multigroup diffusion with removal sources; Transport codes; and Monte Carlo methods. In spite of the numerous Monte- Carlo techniques which are available for penetration and back scattering, serious problems are still encountered in practice with the scattering of gamma rays from walls of buildings which contain critical facilities and also concrete-lined discharge shafts containing irradiated fuel elements. The considerable volume of data in the unclassified literature on the solution of problems of this type in civil defence work appears not to have been evaluated for reactor shield design. In
Radiation shielding calculations for the vista spacecraft
International Nuclear Information System (INIS)
Sahin, Suemer; Sahin, Haci Mehmet; Acir, Adem
2005-01-01
The VISTA spacecraft design concept has been proposed for manned or heavy cargo deep space missions beyond earth orbit with inertial fusion energy propulsion. Rocket propulsion is provided by fusion power deposited in the inertial confined fuel pellet debris and with the help of a magnetic nozzle. The calculations for the radiation shielding have been revised under the fact that the highest jet efficiency of the vehicle could be attained only if the propelling plasma would have a narrow temperature distribution. The shield mass could be reduced from 600 tons in the original design to 62 tons. Natural and enriched lithium were the principle shielding materials. The allowable nuclear heating in the superconducting magnet coils (up to 5 mW/cm 3 ) is taken as the crucial criterion for dimensioning the radiation shielding structure of the spacecraft. The space craft mass is 6000 tons. Total peak nuclear power density in the coils is calculated as ∼5.0 mW/cm 3 for a fusion power output of 17 500 MW. The peak neutron heating density is ∼2.0 mW/cm 3 , and the peak γ-ray heating density is ∼3.0 mW/cm 3 (on different points) using natural lithium in the shielding. However, the volume averaged heat generation in the coils is much lower, namely 0.21, 0.71 and 0.92 mW/cm 3 for the neutron, γ-ray and total nuclear heating, respectively. The coil heating will be slightly lower if highly enriched 6 Li (90%) is used instead of natural lithium. Peak values are then calculated as 2.05, 2.15 and 4.2 mW/cm 3 for the neutron, γ-ray and total nuclear heating, respectively. The corresponding volume averaged heat generation in the coils became 0.19, 0.58 and 0.77 mW/cm 3
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.
International Nuclear Information System (INIS)
Santoro, R.T.; Barnes, J.M.
1983-08-01
Neutron and gamma-ray spectra resulting from the interactions of approx. 14-MeV neutrons in laminated slabs of stainless steel type-304 and borated polyethylene have been calculated using the Monte Carlo code MCNP. The calculated spectra are compared with measured data as a function of slab thickness and material composition and as a function of detector location behind the slabs. Comparisons of the differential energy spectra are made for neutrons with energies above 850 keV and for gamma rays with energies above 750 keV. The measured neutron spectra and those calculated using Monte Carlo methods agree within 5% to 50% depending on the slab thickness and composition and neutron energy. The agreement between the measured and calculated gamma-ray energy spectra is also within this range. The MCNP data are also in favorable agreement with attenuated data calculated previously by discrete ordinates transport methods and the Monte Carlo code SAM-CE
Calculation of the BREN house shielding experiments
International Nuclear Information System (INIS)
Woolson, William A.; Gritzner, Michael L.
1987-01-01
The BREN house transmission experiments provide an excellent set of measurements to validate the calculational procedures that will be used to derive house shielding estimates for the revised dosimetry of the survivors of the Hiroshima and Nagasaki A-bombs. The BREN experiments were performed in realistic full scale models of Japanese residences. Although the radiation spectra and relative intensities of neutrons and gamma rays incident on the houses from the HPRR and the 60 Co source are not appropriate for direct application to the A-bomb survivors, they cover the full energy range of importance. The codes and calculations required to compare with BREN experiments are the same as those needed for the A-bomb dosimetry. They consist of a two-dimensional discrete-ordinates calculation of the free field coupled to an adjoint Monte Carlo calculation in detailed house geometry. The agreement obtained between calculations and the experiments is excellent for neutrons and 60 Co gamma rays. Every house transmission calculation spanning simple to complex configurations and detector locations for the 60 Co and HPRR was within an acceptable margin of error. The gamma-ray TF calculations for the reactor source did not agree well with the experiments. Analysis of this discrepancy, however, strongly indicates that the problem probably does not reside in the calculational procedure but in the measurements themselves. In conclusion, it is believed that the excellent agreement of our calculations with the BREN experiments validates the calculational procedure which is planed to be applied o estimating the house shielding for survivors of the Hiroshima and Nagasaki A-bombs. Certainly, the calculations for Hiroshima and Nagasaki will involve modifications to the code used for the computations reported here, but to the extent that these modifications involve increased calculational complexity to treat more realistic materials and configurations, the benchmark established by these
Shielding calculation for treatment rooms of high energy linear accelerator
International Nuclear Information System (INIS)
Elleithy, M.A.
2006-01-01
A review of German Institute of Standardization (DIN) scheme of the shielding calculation and the essential data required has been done for X-rays and electron beam in the energy range from 1 MeV to 50 MeV. Shielding calculation was done for primary and secondary radiations generated during X-ray operation of Linac. In addition, shielding was done against X-rays generated (Bremsstrahlung) by useful electron beams. The calculations also covered the neutrons generated from the interactions of useful X-rays (at energies above 8 MeV) with the surrounding. The present application involved the computation of shielding against the double scattered components of X-rays and neutrons in the maze area and the thickness of the paraffin wax of the room door. A new developed computer program was designed to assist shielding thickness calculations for a new Linac installation or in replacing an existing machine. The program used a combination of published tables and figures in computing the shielding thickness at different locations for all possible radiation situations. The DIN published data of 40 MeV accelerator room was compared with the program calculations. It was found that there is good agreement between both calculations. The developed program improved the accuracy and speed of calculation
Calculation of parameters for an iron shield experiment
International Nuclear Information System (INIS)
Gavazza, S.
1986-01-01
In this text is carreid out the evaluation of radiation transport methodology, comparying the calculated reactions and dose rates, for neutrons and gama-rays, with the experimental measurements obtained on iron shield, irradiated in YAYOI reactor. Were employed the ENDF/B-IV and VITAMIN-C libraries and the AMPX-II modular system for generation of cross sections, collapsed by the ANISN code. The tranpsort calculations were made by using the DOT 3.5 code, adjusting the spectrum of the iron shield boundary source to the reaction and doses rates, measured at the beginning of shield. The distributions calculated for neutrons and gamma-rays, on iron shield, presented reasonable concordance with the experimental measurements. Finally, is presented a proposal for setting up of an experimental arrangement, using the IEA-R1 reactor, with the purpose of lay down a shielding benchmark. (Author) [pt
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)
Shield calculations, optimization vs. paradigm
International Nuclear Information System (INIS)
Cornejo D, N.; Hernandez S, A.; Martinez G, A.
2006-01-01
Many shieldings have been designed under the criteria of 'Maximum dose rates of project'. It has created the paradigm of those 'low dose rates', for the one which not few specialists would consider unacceptable levels of dose rate superior to the units of μSv.h -1 , independently of the exposure times. At the present time numerous shieldings are being designed considering dose restrictions in real times of exposure. After these new shieldings, the dose rates could be notably superior to those after traditional shieldings, without it implies inadequate designs or constructive errors. In the work significant differences in levels of dose rates and thickness of shieldings estimated by both methods for some typical facilities. It was concluded that the use of real times of exposure is more adequate for the optimization of the Radiological Protection, although this method demands bigger care in its application. (Author)
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.)
International Nuclear Information System (INIS)
Kharrati, H.
2005-01-01
The primary, scatter, and leakage doses(in Gy), which constitute the data base for calculating shielding requirements for x-ray facilities, are often converted to the equivalent dose (in sievert) by using a constant of conversion of 1.145Sv/Gy. This constant is used for diagnostic radiology as well as for mammography spectra, and is derived by considering an exposure of 1 R corresponds to an air kerma of 8.73 m Gy, which renders by tradition an equivalent dose of 10 mSv. However, this conversion does not take into account the energy dependence of the conversion coefficients relating air kerma to the equivalent dose as described in ICRU report. Moreover, current radiation protection standards propose the use of the quantity ambient dose equivalent in order to qualify the efficiently of given radiation shielding. Therefore, in this study, a new approach has been introduced for derivation ambient dose equivalent from air kerma to calculate shielding requirements in mammography facilities. This new approach has been used to compute the conversion coefficients relating air kerma to ambient dose equivalent for mammography reference beam series of the Netherlands Metrology Institute Van Swinden Laboratorium (NMi), National Institute of Standards and Technology (NIST), and International Atomic Energy Agency (AIEA) laboratories. The calculation has been performed by the means of two methods which show a maximum deviation less than 10%2 for the primary, scatter, and leakage radiations. The results show that the conversion coefficients vary from 0.242 Sv/ Gy to 0.692 Sv/Gy with an average value of 0.436 Sv/Gy for the primary and the scatter radiations, and form 0.156 Sv/Gy to 1.329 Sv/Gy with an average value of 0.98 Sv/Gy for the leakage radiation. Simpkin et al. using an empirical approach propose a conversion value of 0.50 Sv/Gy for the mammography x-ray spectra. This value approximately coincides with the average conversion value of 0.436 Sv/Gy obtained in this work for
International Nuclear Information System (INIS)
Grimstone, M.J.
1978-06-01
The WRS Modular Programming System has been developed as a means by which programmes may be more efficiently constructed, maintained and modified. In this system a module is a self-contained unit typically composed of one or more Fortran routines, and a programme is constructed from a number of such modules. This report describes one WRS module, the function of which is to calculate the gamma-ray flux, dose, or heating rate in a slab shield using the build-up factor method. The information given in this manual is of use both to the programmer wishing to incorporate the module in a programme, and to the user of such a programme. (author)
PKI, Gamma Radiation Reactor Shielding Calculation by Point-Kernel Method
International Nuclear Information System (INIS)
Li Chunhuai; Zhang Liwu; Zhang Yuqin; Zhang Chuanxu; Niu Xihua
1990-01-01
1 - Description of program or function: This code calculates radiation shielding problem of gamma-ray in geometric space. 2 - Method of solution: PKI uses a point kernel integration technique, describes radiation shielding geometric space by using geometric space configuration method and coordinate conversion, and makes use of calculation result of reactor primary shielding and flow regularity in loop system for coolant
Subsurface Shielding Source Term Specification Calculation
International Nuclear Information System (INIS)
S.Su
2001-01-01
The purpose of this calculation is to establish appropriate and defensible waste-package radiation source terms for use in repository subsurface shielding design. This calculation supports the shielding design for the waste emplacement and retrieval system, and subsurface facility system. The objective is to identify the limiting waste package and specify its associated source terms including source strengths and energy spectra. Consistent with the Technical Work Plan for Subsurface Design Section FY 01 Work Activities (CRWMS M and O 2001, p. 15), the scope of work includes the following: (1) Review source terms generated by the Waste Package Department (WPD) for various waste forms and waste package types, and compile them for shielding-specific applications. (2) Determine acceptable waste package specific source terms for use in subsurface shielding design, using a reasonable and defensible methodology that is not unduly conservative. This calculation is associated with the engineering and design activity for the waste emplacement and retrieval system, and subsurface facility system. The technical work plan for this calculation is provided in CRWMS M and O 2001. Development and performance of this calculation conforms to the procedure, AP-3.12Q, Calculations
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.
Calculated shielding factors for selected European houses
International Nuclear Information System (INIS)
Hedemann Jensen, P.
1984-12-01
Shielding factors for gamma radiation from activity deposited on structures and ground surfaces have been calculated with the computer model DEPSHIELD for single-family and multi-storey buildings in France, United Kingdom and Denmark. For all three countries it was found that the shielding factors for single-family houses are approximately a factor of 2 - 10 higher that those for buildings with five or more storeys. Away from doors and windows the shielding factors for French, British, and Danish single-family houses are in the range 0.03 - 0.1, 0.06 - 0.4, and 0.07 - 0.3, respectively. The uncertainties of the calculations are discussed and DEPSHIELD-results are compared with other methods as well as with experimental results. (author)
Parameters calculation of a shielding experiment and evaluation of calculation methodology
International Nuclear Information System (INIS)
Gavazza, S.; Otto, A.C.; Gomes, I.C.; Maiorino, J.R.
1986-01-01
In this text is carried out the evaluation of radiation transport methodology, comparying the calculated reactions and dose rates, for neutrons and gamma-rays, with the experimental measurements obtained on iron shield, irradiated in YAYOI reactor. Were employed the ENDF/B-IV and VITAMIN-C libraries and the AMPX-II modular system for generation of cross sections, collapsed by the ANISN code. The transport calculation were made by using the DOT 3.5 code, adjusting the spectrum of the iron shield boundary source to the reactions and dose rates, measured at the beginning of shield. The distributions calculated for neutrons and gamma-rays, on iron shield, presented coherence with the experimental measurements. (Author) [pt
FLUKA shielding calculations for the FAIR project
International Nuclear Information System (INIS)
Fehrenbacher, Georg; Kozlova, Ekaterina; Radon, Torsten; Sokolov, Alexey
2015-01-01
FAIR is an international accelerator project being in construction at GSI Helmholtz center for heavy ion research in Darmstadt. The Monte Carlo program FLUKA is used to study radiation protection problems. The contribution deals with general application possibilities of FLUKA and for FAIR with respect the radiation protection planning. The necessity to simulate the radiation transport through shielding of several meters thickness and to determine the equivalent doses outside the shielding with sufficient accuracy is demonstrated using two examples under consideration of the variance reduction. Results of simulation calculations for activation estimation in accelerator facilities are presented.
Equivalent-spherical-shield neutron dose calculations
International Nuclear Information System (INIS)
Russell, G.J.; Robinson, H.
1988-01-01
Neutron doses through 162-cm-thick spherical shields were calculated to be 1090 and 448 mrem/h for regular and magnetite concrete, respectively. These results bracket the measured data, for reinforced regular concrete, of /approximately/600 mrem/h. The calculated fraction of the high-energy (>20 MeV) dose component also bracketed the experimental data. The measured and calculated doses were for a graphite beam stop bombarded with 100 nA of 800-MeV protons. 6 refs., 2 figs., 1 tab
Benchmarks for evaluation of shielding calculations
International Nuclear Information System (INIS)
Coelho, P.R.P.; Maiorino, J.R.
1989-01-01
The spatial-energy neutron distribution emerging from a laminated shielding (stainless, polyethylene and lead) were measured by a fast neutron spectrometer and some experimental results were compared with those calculated by a network of codes. The source neutrons incident in the shielding were 14 MeV neutrons from a H-3(d,n)He-4 reaction coming from a Van de Graaff accelerator. Experimentally was verified a good radial symmetry of neutron energy-spectrum, and also a moderation and attenuation effect for points located out of the central axis of symmetry. These results indicate that the experiment can be well modelated by R-Z geometry. A neutron-energy spectra calculated by DOT 3.5 was compared with the measured spectra, showing a good agreement in the shape and value of the spectra (12% for an integrated spectrum from 2 to 16 MeV). (author) [pt
Monte Carlo methods for shield design calculations
International Nuclear Information System (INIS)
Grimstone, M.J.
1974-01-01
A suite of Monte Carlo codes is being developed for use on a routine basis in commercial reactor shield design. The methods adopted for this purpose include the modular construction of codes, simplified geometries, automatic variance reduction techniques, continuous energy treatment of cross section data, and albedo methods for streaming. Descriptions are given of the implementation of these methods and of their use in practical calculations. 26 references. (U.S.)
International Nuclear Information System (INIS)
Kharrati, Hedi
2005-01-01
In this study, a new approach has been introduced for derivation of the effective dose from air kerma to calculate shielding requirements in mammography facilities. This new approach has been used to compute the conversion coefficients relating air kerma to the effective dose for the mammography reference beam series of the Netherlands Metrology Institute Van Swinden Laboratorium, National Institute of Standards and Technology, and International Atomic Energy Agency laboratories. The results show that, in all cases, the effective dose in mammography energy range is less than 25% of the incident air kerma for the primary and the scatter radiations and does not exceed 75% for the leakage radiation
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.
Software Tools for Measuring and Calculating Electromagnetic Shielding Effectiveness
National Research Council Canada - National Science Library
Tesny, Neal
2005-01-01
The evaluation and the analysis of high-altitude electromagnetic pulse response of shielded enclosures require the availability of software tools able to acquire data and calculate shielding effectiveness...
Radiological shielding of low power compact reactor: calculation and design
International Nuclear Information System (INIS)
Marino, Raul
2004-01-01
The development of compact reactors becoming a technology that offers great projection and innumerable use possibilities, both in electricity generation and in propulsion.One of the requirements for the operation of this type of reactor is that it must include a radiological shield that will allow for different types of configurations and that, may be moved with the reactor if it needs to be transported.The nucleus of a reactor emits radiation, mainly neutrons and gamma rays in the heat of power, and gamma radiation during the radioactive decay of fission products.This radiation must be restrained in both conditions of operation to avoid it affecting workers or the public.The combination of different materials and properties in layers results in better performance in the form of a decrease in radiation, hence causing the dosage outside the reactor, whether in operation or shut down, to fall within the allowed limits.The calculations and design of radiological shields is therefore of paramount importance in reactor design.The choice of material and the design of the shield have a strong impact on the cost and the load capacity, the latter being one of the characteristics to optimize.The imposed condition of design is that the reactor can be transported together with the decay shield in a standard container of 40 foot [es
Shielding calculations for the SNO detector
International Nuclear Information System (INIS)
Earle, E.D.; Wong, P.Y.
1987-05-01
The gamma-ray background into the central D 2 O vessel of the SNO detector due to Th and U in the rock, concrete, and photomultipliers is calculated. A cylindrical geometry and concrete thicknesses of 0.5 and 1 m are assumed. The effect of adding boron to the concrete is also considered. It is concluded that backgrounds from (α,n) reactions can be reduced to the required level. These calculations will assist in finalizing the detector design but additional calculations will be required as new design details become known
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)
Calculation and design for SSRF's bulk shield
Energy Technology Data Exchange (ETDEWEB)
Fang, K.M. [Shanghai Institute of Applied Physics, Chinese Academy of Science (China)]. E-mail: fangkm@sinap.ac.cn; Xu, X.J. [Shanghai Institute of Applied Physics, Chinese Academy of Science (China); Cai, J.H. [Shanghai Institute of Applied Physics, Chinese Academy of Science (China)
2006-12-15
Shielding design objectives for the SSRF are chosen, assumptions for beam loss rates are given, the methods used on the APS by Moe are summarized and introduced to make calculation and design on bulk shield, the factor of skyshine is also considered, design thicknesses for SSRF's bulk shield are presented.
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)
Shielding of the child's head during x-ray studies
International Nuclear Information System (INIS)
Tolmach, Eh.U.
1985-01-01
Three devices for X-ray shielding of child's head are suggested; the first one is a protective attachment for shielding a child being in horizontal position on an X-ray table; the second one is a protective stand for shielding head and body at roentgenofraphy of upper extremities of a child sitting near the X-ray table; the third one is a prot ctive suspension for shielding the head of a child being in vertical position
Determination of gamma ray shielding parameters of rocks and concrete
Obaid, Shamsan S.; Gaikwad, Dhammajyot K.; Pawar, Pravina P.
2018-03-01
Gamma shielding parameters such as mass attenuation coefficient (μ/ρ), effective atomic number (Zeff) and electron density (Neff) have been measured and calculated for rocks and concrete in the energy range 122-1330 keV. The measurements have been carried out at 122, 356, 511, 662, 1170, 1275, 1330 keV gamma ray energies using a gamma spectrometer includes a NaI(Tl) scintillation detector and MCA card. The atomic and electronic cross sections have also been investigated. Experimental and calculated (WinXCom) values were compared, and good agreement has been observed within the experimental error. The obtained results showed that feldspathic basalt, compact basalt, volcanic rock, dolerite and pink granite are more efficient than the sandstone and concrete for gamma ray shielding applications.
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
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.
International Nuclear Information System (INIS)
Barros, Gustavo A.S.J.; Caldas, Linda V.E.
2009-01-01
This work performed the shield calculation of the future rooms walls of the five X-ray equipment of the Instrument Calibration Laboratory of the IPEN, Sao Paulo, Brazil, which will be constructed in project of laboratory enlargement. The obtained results by application of a calculation methodology from an international regulation have shown that the largest thickness of shielding (25.7 cm of concrete or 7.1 mm of lead) will be of the wall which will receive the primary beam of the equipment with a 320 kV voltage. The cost/benefit analysis indicated the concrete as the best material option for the shielding
nmr spectroscopic study and dft calculations of giao nmr shieldings
African Journals Online (AJOL)
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3Department of Physics, Arts and Science Faculty, Dumlupinar University, Kütahya, ... 1H, 13C NMR chemical shifts and 1JCH coupling constants of .... then estimated using the corresponding TMS shieldings calculated in advance at the same.
Shielding calculations for the Intense Neutron Source Facility. Final report
International Nuclear Information System (INIS)
Battat, M.E.; Henninger, R.J.; Macdonald, J.L.; Dudziak, D.J.
1978-06-01
Results of shielding calculations for the Intnse Neutron Source (INS) facility are presented. The INS facility is designed to house two sources, each of which will produce D--T neutrons with intensities in the range from 1 to 3 x 10 15 n/s on a continuous basis. Topics covered include the design of the biological shield, use of two-dimensional discrete-ordinates results to specify the source terms for a Monte Carlo skyshine calculation, air activation, and dose rates in the source cell (after shutdown) due to activation of the biological shield
Alternative methodology for irradiation reactor experimental shielding calculation
International Nuclear Information System (INIS)
Vellozo, Sergio de Oliveira; Vital, Helio de Carvalho
1996-01-01
Due to a change in the project of the Experimental Irradiation Reactor, its shielding design had to be recalculated according to an alternative simplified analytical approach, since the standard transport calculations were temporarily unavailable. In the calculation of the new width for the shielding made up of steel and high-density concrete layers, the following radiation components were considered: fast neutrons and primary gammas (produced by fission and beta decay), from the core; and secondary gammas, produced by thermal neutron capture in the shielding. (author)
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)
Shielding design at Fermilab: Calculations and measurements
International Nuclear Information System (INIS)
Cossairt, J.D.
1986-11-01
The development of the Fermilab accelerator complex during the past two decades from its concept as the ''200 BeV accelerator'' to that of the present tevatron, designed to operate at energies as high as 1 TeV, has required a coincidental refinement and development in methods of shielding design. In this paper I describe these methods as used by the radiation protection staff of Fermilab. This description will review experimental measurements which substantiate these techniques in realistic situations. Along the way, observations will be stated which likely are applicable to other protron accelerators in the multi-hundred GeV energy region, including larger ones yet to be constructed
Radiation shielding calculation for the MOX fuel fabrication plant Melox
International Nuclear Information System (INIS)
Lee, Y.K.; Nimal, J.C.; Chiron, M.
1994-01-01
Radiation shielding calculation is an important engineering work in the design of the MOX fuel fabrication plant MELOX. Due to the recycle of plutonium and uranium from UO2 spent fuel reprocessing and the large capacity of production (120t HM/yr.), the shielding design requires more attention in this LWR fuel plant. In MELOX, besides several temporary storage facilities of massive fissile material, about one thousand radioactive sources with different geometries, forms, densities, quantities and Pu concentrations, are distributed through different workshops from the PuO 2 powder reception unit to the fuel assembly packing room. These sources, with or without close shield, stay temporarily in different locations, containers and glove boxes. In order to optimize the dimensions, the material and the cost of shield as well as to limit the calculation work in a reasonable engineer-hours, a calculation scheme for shielding design of MELOX is developed. This calculation scheme has been proved to be useful in consideration of the feedback from the evolutionary design and construction. The validated shielding calculations give a predictive but reliable radiation doses information. (authors). 2 figs., 10 refs
Simplified shielding calculation system for high-intensity proton accelerators
Energy Technology Data Exchange (ETDEWEB)
Masumura, Tomomi; Nakashima, Hiroshi; Nakane, Yoshihiro; Sasamoto, Nobuo [Center for Neutron Science, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan)
2000-06-01
A simplified shielding calculation system is developed for applying conceptual shielding design of facilities in the joint project for high-intensity proton accelerators. The system is composed of neutron transmission calculation part for bulk shielding using simplified formulas: Moyer model and Tesch's formula, and neutron skyshine calculation part using an empirical formula: Stapleton's formula. The system is made with the Microsoft Excel software for user's convenience. This report provides a manual for the system as well as calculation conditions used in the calculation such as Moyer model's parameters. In this report preliminary results based on data at December 8, 1999, are also shown as an example. (author)
International Nuclear Information System (INIS)
Diop, C.M'B.
1984-01-01
This thesis presents a global shielding calculation scheme for neutron and gamma rays arising from the Tokamak TORE SUPRA fusion device, in which a deuterium plasma is used. To study the shield parameters we have elabored a important chaining of neutron and gamma transport codes, TRIPOLI, ANISN, MERCURE 4, allowing to evaluate the radial and skyshine components of the dose rate behind the concrete shield. The study of thermonuclear neutron activation is fundamental to define a tokamak exploitation strategy. For this, two formalisme have been developed. They are based on a modelization of the activation reaction rates according to TRIPOLI, ANISN, and MERCURE 4 codes capabilities. The first one calculates, in one dimensional geometry, the desactivation gamma dose rate inside the vacuum chamber. The second one is a tridimensional model which determines the spatial variation of the gamma dose rate in the machine room. The problem of the existence of runaway electrons and associated secondaries radiations, bremsstrahlung gamma rays particularly, is approched. The results which are presented have contributed to define the parameters of the concrete shield and a strategy for TORE SUPRA Tokamak exploitation [fr
ASOP, Shield Calculation, 1-D, Discrete Ordinates Transport
International Nuclear Information System (INIS)
1993-01-01
1 - Nature of physical problem solved: ASOP is a shield optimization calculational system based on the one-dimensional discrete ordinates transport program ANISN. It has been used to design optimum shields for space applications of SNAP zirconium-hydride-uranium- fueled reactors and uranium-oxide fueled thermionic reactors and to design beam stops for the ORELA facility. 2 - Method of solution: ASOP generates coefficients of linear equations describing the logarithm of the dose and dose-weight derivatives as functions of position from data obtained in an automated sequence of ANISN calculations. With the dose constrained to a design value and all dose-weight derivatives required to be equal, the linear equations may be solved for a new set of shield dimensions. Since changes in the shield dimensions may cause the linear functions to change, the entire procedure is repeated until convergence is obtained. The detailed calculations of the radiation transport through shield configurations for every step in the procedure distinguish ASOP from other shield optimization computer code systems which rely on multiple component sources and attenuation coefficients to describe the transport. 3 - Restrictions on the complexity of the problem: Problem size is limited only by machine size
Calculation of a toroidal labyrinth shields
International Nuclear Information System (INIS)
Sul'kin, A.G.
1979-01-01
Calculation of protective case with a toroidal labyrinth channel, being one of the main design elements of hose gamma-devices, is presented. The case provides relative isotropic distribution of radiation outside protection limits. The main geometric parameters of the channel are determined: r-radius of the channel hole, rho-bend radius of the channel axis, β-angle of the channel bend. General exposure dose rate of γ-radiation in the detection point at l distance (usually l=100 m during calculations), is also calculated. Differential current dose albedo values have been found for certain combinations of parameters of the labyrinth channel. It is considered for simplification of labyrinth channel calculations, that backward radiation scattering passes, without energy change and isotropically, due to which differential current albedo values of γ-radiation for any incidence angle may be determined from integral albedo current values by the empirie formula
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.
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.
Shielding calculation techniques used in the design of storage systems
International Nuclear Information System (INIS)
Wang, S.S.; Massey, J.V.
1986-01-01
The shielding design and analysis of a concrete modular spent fuel storage system are discussed. Particular attention is given to comparing various computation techniques in determining bulk shielding thickness, and also in dealing with the radiation streaming effect through the air exist penetration openings in the module. Three computer codes QADMOD, ANISN, and DOT-IV were used to solve the same problem. In addition, hand albedo calculation were done to augment the result of the QADMOD calculation to properly deal with the surface scattering
X-ray face mask and chest shield device
International Nuclear Information System (INIS)
Moti, S.
1981-01-01
A protective face mask is designed to shield an x-ray technician or machine operator primarily from random secondary or scatter x-rays deflected towards his face, head and neck by the table, walls, equipment and other reflecting elements in an x-ray room or chamber. The face mask and chest shield device can be mounted on a patient's shoulders in reverse attitude to protect the back of a patient's head and neck from the x-ray beam. The face mask is relatively or substantially transparent and contains lead in combination with a plastic ionomer or comonomer, which to a degree absorbs or resists penetration of the random deflected secondary or scatter x-rays or the x-ray beam through the mask. The face mask is removably attachable to the chest shield for easy application of the device to and support upon the shoulders of the technician or the patient. (author)
Comparison of calculational methods for liquid metal reactor shields
International Nuclear Information System (INIS)
Carter, L.L.; Moore, F.S.; Morford, R.J.; Mann, F.M.
1985-09-01
A one-dimensional comparison is made between Monte Carlo (MCNP), discrete ordinances (ANISN), and diffusion theory (MlDX) calculations of neutron flux and radiation damage from the core of the Fast Flux Test Facility (FFTF) out to the reactor vessel. Diffusion theory was found to be reasonably accurate for the calculation of both total flux and radiation damage. However, for large distances from the core, the calculated flux at very high energies is low by an order of magnitude or more when the diffusion theory is used. Particular emphasis was placed in this study on the generation of multitable cross sections for use in discrete ordinates codes that are self-shielded, consistent with the self-shielding employed in the generation of cross sections for use with diffusion theory. The Monte Carlo calculation, with a pointwise representation of the cross sections, was used as the benchmark for determining the limitations of the other two calculational methods. 12 refs., 33 figs
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)
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
Tire inspection system with shielded x-ray source
International Nuclear Information System (INIS)
Heisner, D.N.; Palermo, A. Jr.; Loyer, P.K.
1976-01-01
An automated tire inspection system is described which employs a penetrative radiation, such as x-radiation, to inspect the integrity of portions of tires fed sequentially along a feed path through a centering station and into a shielded enclosure where an inspection station is defined. Features of the system include a continuously operating x-ray source movable between inspection and retracted positions, and an x-ray shield for covering the source when it is retracted to permit the doors of the shielded enclosure to be opened without danger from escaping radiation. 19 Claims, 38 Drawing Figures
Shielding calculations for ships carrying irradiated nuclear fuel
International Nuclear Information System (INIS)
Dean, M.H.
1985-01-01
A number of ships have been constructed to carry irradiated fuel from Japan to the U.K. and France, for reprocessing. About 20 transport flasks may be carried on each voyage. Permanent shielding must be provided on the ships to ensure that no member of the crew receives an annual dose greater than a specified limit. As the fuel is of varying type and radiation history, and as flasks of differing designs are used, many shielding calculations are needed. There are a number of difficulties in making shielding calculations for the ships. The geometry is complex, dimensions are large and considerable air spaces are involved. The paper considers possible methods of calculation. The line-of-sight method is chosen for most of the calculations, for both γ-radiation and neutrons. The basic data which is used in the calculations is described. As the methods of calculation are somewhat approximate, it is necessary to provide confirmation that they are sufficiently accurate. Validation has been provided in two ways. First, measurements have been made on board one of the ships, Pacific Crane, and these have been checked against calculation. Second, a simplified model of the flasks and ship has been set up, and calculations checked against more sophisticated methods. Results of the validation checks are presented, and it is shown that adequate accuracy is achieved. (author)
RZ calculations for self shielded multigroup cross sections
Energy Technology Data Exchange (ETDEWEB)
Li, M.; Sanchez, R.; Zmijarevic, I.; Stankovski, Z. [Commissariat a l' Energie Atomique CEA, Direction de l' Energie Nucleaire, DEN/DM2S/SERMA/LENR, 91191 Gif-sur-Yvette Cedex (France)
2006-07-01
A collision probability method has been implemented for RZ geometries. The method accounts for white albedo, specular and translation boundary condition on the top and bottom surfaces of the geometry and for a white albedo condition on the outer radial surface. We have applied the RZ CP method to the calculation of multigroup self shielded cross sections for Gadolinia absorbers in BWRs. (authors)
RZ calculations for self shielded multigroup cross sections
International Nuclear Information System (INIS)
Li, M.; Sanchez, R.; Zmijarevic, I.; Stankovski, Z.
2006-01-01
A collision probability method has been implemented for RZ geometries. The method accounts for white albedo, specular and translation boundary condition on the top and bottom surfaces of the geometry and for a white albedo condition on the outer radial surface. We have applied the RZ CP method to the calculation of multigroup self shielded cross sections for Gadolinia absorbers in BWRs. (authors)
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.
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
International Nuclear Information System (INIS)
Wu Chuan; Guo Fanqing; Purdy, James A
2006-01-01
This paper reports a step-by-step shielding calculation recipe for a helical tomotherapy unit (TomoTherapy Inc., Madison, WI, USA), recently installed in an existing Varian 600C treatment room. Both primary and secondary radiations (leakage and scatter) are explicitly considered. A typical patient load is assumed. Use factor is calculated based on an analytical formula derived from the tomotherapy rotational beam delivery geometry. Leakage and scatter are included in the calculation based on corresponding measurement data as documented by TomoTherapy Inc. Our calculation result shows that, except for a small area by the therapists' console, most of the existing Varian 600C shielding is sufficient for the new tomotherapy unit. This work cautions other institutions facing the similar situation, where an HT unit is considered for an existing LINAC treatment room, more secondary shielding might be considered at some locations, due to the significantly increased secondary shielding requirement by HT. (note)
Gamma-ray shielding design and performance test of WASTEF
International Nuclear Information System (INIS)
Matsumoto, Seiichiro; Aoyama, Saburo; Tashiro, Shingo; Nagai, Shiro
1984-06-01
The Waste Safety Testing Facility (WASTEF) was planned in 1978 to test the safety performance of HLW vitrified forms under the simulated conditions of long term storage and disposal, and completed in August 1981. The designed feature of the facility is to treat the vitrified forms contain actual high-level wastes of 5 x 10 4 Ci in maximum with 5 units of concrete shilded hot cells (3 units : Bate-Gamma cells, 2 units : Alpha-Gamma cells) and one units of Alpha-Gamma lead shielded cell, and to store radioactivity of 10 6 Ci in maximum. The safety performance of this facility is fundamentally maintained with confinement of radioactivity and shielding of the radiation. This report describes the method of gamma-ray shielding design, evaluation of the shielding test performed by using sealded gamma-ray sources(Co-60). (author)
Shielding calculations for ships carrying irradiated nuclear fuel
International Nuclear Information System (INIS)
Burstall, R.F.; Dean, M.H.
1983-01-01
A number of ships have been constructed to carry irradiated fuel from Japan to the UK and France, for reprocessing. About twenty transport flasks may be carried on each voyage. Permanent shielding must be provided on the ships to ensure that no member of the crew receives an annual dose rate greater than a specified limit. As the fuel is of varying type and radiation history, and as flasks of differing designs are used, many calculations are needed. There are a number of difficulties in making shielding calculations for the ships. The geometry is complex, dimensions are large, and considerable air spaces are involved. The paper considers possible methods of calculation. The line-of-sight method is chosen for most of the calculations, for both gamma radiation and neutrons. The basic data which is used in the calculations is described. As the methods of calculation are somewhat approximate, it is necessary to provide confirmation that they are sufficiently accurate. Validation has been provided in two ways. First, measurements have been made on board the ships, and these have been checked against calculation. Second, a simplified model of the flasks and ship has been set up, and calculations checked against more sophisticated methods. Results of the validation checks are presented, and it is shown that adequate accuracy is achieved. (author)
Shielding calculations for the TFTR neutral beam injectors
International Nuclear Information System (INIS)
Santoro, R.T.; Lillie, R.A.; Alsmiller, R.G. Jr.; Barnes, J.M.
1979-07-01
Two-dimensional discrete ordinates calculations have been performed to determine the location and thickness of concrete shielding around the Tokamak Fusion Test Reactor (TFTR) neutral beam injectors. Two sets of calculations were performed: one to determine the dose equivalent rate on the roof and walls of the test cell building when no injectors are present, and one to determine the contribution to the dose equivalent rate at these locations from radiation streaming through the injection duct. Shielding the side and rear of the neutral beam injector with 0.305 and 0.61 m of concrete, respectively, and lining the inside of the test cell wall with an additional layer of concrete having a thickness of 0.305 m and a height above the axis of deuteron injection of 3.10 m are sufficient to maintain the biological dose equivalent rate outside the test cell to approx. 1 mrem/DT pulse
Problems in radiation shielding calculations with Monte Carlo methods
International Nuclear Information System (INIS)
Ueki, Kohtaro
1985-01-01
The Monte Carlo method is a very useful tool for solving a large class of radiation transport problem. In contrast with deterministic method, geometric complexity is a much less significant problem for Monte Carlo calculations. However, the accuracy of Monte Carlo calculations is of course, limited by statistical error of the quantities to be estimated. In this report, we point out some typical problems to solve a large shielding system including radiation streaming. The Monte Carlo coupling technique was developed to settle such a shielding problem accurately. However, the variance of the Monte Carlo results using the coupling technique of which detectors were located outside the radiation streaming, was still not enough. So as to bring on more accurate results for the detectors located outside the streaming and also for a multi-legged-duct streaming problem, a practicable way of ''Prism Scattering technique'' is proposed in the study. (author)
Bulk Shielding Calculation for 90 .deg. Bending Section of RISP
Energy Technology Data Exchange (ETDEWEB)
Oh, J. H.; Jung, N. S.; Lee, H. S. [Pohang Accelerator Laboratory, Pohang (Korea, Republic of); Oranj, L. Mokhtari [POSTECH, Pohang (Korea, Republic of); Ko, S. K. [Univ. of Ulsan, Ulsan (Korea, Republic of)
2014-10-15
The charge state of {sup 238}U beams with maximum intensity was 79+ among multi-charge states of 70+ to 89+, which were estimated by using LISE++ code. The bending section consists of twenty four quadrupoles, two dipoles, two two-cell type superconducting RF cavities and eleven slits. The complicated radiation environment is caused by the beam losses occurred normally during the stripping process and when the produced {sup 238}U beams are transported along the beam line. Secondary radiations generated by {sup 238}U beams irradiation are very important for predicting the prompt and residual doses and the radiation damage at the component. The production characteristics of neutron and photon from thin carbon and thick iron were studied to set up the shielding strategy. The dose estimation was done to the pre-designed the tunnel structure. In these calculations, major Monte Carlo codes, PHITS and FLUKA, were used. The present study provided information of shielding analysis for the 90 .deg. bending section of RISP facility. The source term was evaluated to determine fundamental parameter of the shielding analysis using PHITS and FLUKA codes. And the distribution of the dose rate at the outside of thick shielding wall was presented.
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.
Calculation analysis of the thickness of radiation shield for the RIA equipment IP10
International Nuclear Information System (INIS)
Benar Bukit; Kristiyanti; Hari Nurcahyadi
2011-01-01
Calculation Analysis has been performed on the thickness of radiation shield for the design of the Radioimmunoassay (RIA) IP10 counters using five detectors arranged in parallel. The calculation is intended to ensure that the radiation on each detector does not influence each other. The radiation shield is made of lead. The calculation of lead thickness was based on the principle of the lead plates absorptive power toward the gamma ray of a certain energy. which is the function of linear absorption coefficient and the material thickness. Assuming the use of Iodium-125(I-125) source with an activity 10 µCi, and expecting an absorptive power of 95%, calculations showed that the required lead thickness is equal to 0,013 cm. Since lead is soft and its availability in the market is limited, lead plate of 2 mm thickness are used instead, so that counting result for the detectors do not influence each other. (author)
Energy Technology Data Exchange (ETDEWEB)
Silva, Júlia R.A.S. da; Vieira, José W., E-mail: j.rafaela14@gmail.com, E-mail: jose.wilson@recife.ifpe.edu.br [Instituto Federal de Educação, Ciência e Tecnologia de Pernambuco (IFPE), Recife - PE (Brazil); Lima, Fernando R. A., E-mail: falima@cnen.gov.br [Centro Regional de Ciências Nucleares do Nordeste (CRCN-NE/CNEN-PE), Recife, PE (Brazil)
2017-07-01
The objective was to program a software to calculate the required thicknesses to attenuate X-rays in kilovoltage of 60 kV, 80 kV, 110 kV and 150 kV. The conventional methodological parameters for structural shield calculations established by the NCRP (National Council on Radiation Protection and Measurements) were presented. The descriptive and exploratory methods allowed the construction of the JULIA. In this sense and based on the result obtained, the tool presented is useful for professionals who wish to design structural shielding in radiodiagnostic and/or therapy. The development of calculations in the computational tool corresponds to the accessibility, optimization of time and estimation close to the real. Such heuristic exercise represents improvement of calculations for the estimation of primary barriers with barite.
Some experience of shielding calculations by combinatorial method
International Nuclear Information System (INIS)
Korobejnikov, V.V.; Oussanov, V.I.
1996-01-01
Some aspects of the compound systems shielding calculations by a combinatorial approach are discussed. The effectiveness of such an approach is based on the fundamental characteristic of a compound system: if some element of the system have in itself mathematical or physical properties favorable for calculation, these properties may be used in a combinatorial approach and are lost when the system is being calculated in the whole by a direct approach. The combinatorial technique applied is well known. A compound system are being splitting for two or more auxiliary subsystems (so that calculation each of them is a more simple problem than calculation of the original problem (or at last is a soluble problem if original one is not). Calculation of every subsystem are carried out by suitable method and code, the coupling being made through boundary conditions or boundary source. The special consideration in the paper is given to a fast reactor shielding combinatorial analysis and to the testing of the results received. (author)
International Nuclear Information System (INIS)
1980-01-01
X-ray screen unit consisting of a light transmissive carrier onto which scintillation material is deposited, which is able to generate light under the influence of incident X-ray irradiation, characterised in that the X-ray screen comprises a number of sectors, wherein the surface with respect to the incident X-radiation is maintained at an acute angle. (G.C.)
Effects of scattering anisotropy approximation in multigroup radiation shielding calculations
International Nuclear Information System (INIS)
Altiparmakov, D.
1983-01-01
Expansion of the scattering cross sections into Legendre series is the usual way of solving neutron transport problems. Because of the large space gradients of the neutron flux, the effects of that approximation become especially remarkable in the radiation shielding calculations. In this paper, a method taking into account the scattering anisotropy is presented. From the point od view of the accuracy and computing rate, the optimal approximation of the scattering anisotropy is established for the basic protective materials on the basis of simple problem calculations. (author)
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)
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)
Monteray Mark-I: Computer program (PC-version) for shielding calculation with Monte Carlo method
International Nuclear Information System (INIS)
Pudjijanto, M.S.; Akhmad, Y.R.
1998-01-01
A computer program for gamma ray shielding calculation using Monte Carlo method has been developed. The program is written in WATFOR77 language. The MONTERAY MARH-1 is originally developed by James Wood. The program was modified by the authors that the modified version is easily executed. Applying Monte Carlo method the program observe photon gamma transport in an infinity planar shielding with various thick. A photon gamma is observed till escape from the shielding or when its energy less than the cut off energy. Pair production process is treated as pure absorption process that annihilation photons generated in the process are neglected in the calculation. The out put data calculated by the program are total albedo, build-up factor, and photon spectra. The calculation result for build-up factor of a slab lead and water media with 6 MeV parallel beam gamma source shows that they are in agreement with published data. Hence the program is adequate as a shielding design tool for observing gamma radiation transport in various media
Unresolved resonance self shielding calculation: causes and importance of discrepancies
International Nuclear Information System (INIS)
Ribon, P.; Tellier, H.
1986-09-01
To compute the self shielding coefficient, it is necessary to know the point-wise cross-sections. In the unresolved resonance region, we do not know the parameters of each level but only the average parameters. Therefore we simulate the point-wise cross-section by random sampling of the energy levels and resonance parameters with respect to the Wigner law and the X 2 distributions, and by computing the cross-section in the same way as in the resolved regions. The result of this statistical calculation obviously depends on the initial parameters but also on the method of sampling, on the formalism which is used to compute the cross-section or on the weighting neutron flux. In this paper, we will survey the main phenomena which can induce discrepancies in self shielding computations. Results are given for typical dilutions which occur in nuclear reactors. 8 refs
Unresolved resonance self shielding calculation: causes and importance of discrepancies
International Nuclear Information System (INIS)
Ribon, P.; Tellier, H.
1986-01-01
To compute the self shielding coefficient, it is necessary to know the point-wise cross-sections. In the unresolved resonance region, the parameters of each level are not known; only the average parameters. Therefore the authors simulate the point-wise cross-section by random sampling of the energy levels and resonance parameters with respect to the Wigner law and the x 2 distributions, and by computing the cross-section in the same way as in the resolved regions. The result of this statistical calculation obviously depends on the initial parameters but also on the method of sampling, on the formalism which is used to compute the cross-section or on the weighting neutron flux. In this paper, the authors survey the main phenomena which can induce discrepancies in self shielding computations. Results are given for typical dilutions which occur in nuclear reactors
Research on Primary Shielding Calculation Source Generation Codes
Zheng, Zheng; Mei, Qiliang; Li, Hui; Shangguan, Danhua; Zhang, Guangchun
2017-09-01
Primary Shielding Calculation (PSC) plays an important role in reactor shielding design and analysis. In order to facilitate PSC, a source generation code is developed to generate cumulative distribution functions (CDF) for the source particle sample code of the J Monte Carlo Transport (JMCT) code, and a source particle sample code is deveoped to sample source particle directions, types, coordinates, energy and weights from the CDFs. A source generation code is developed to transform three dimensional (3D) power distributions in xyz geometry to source distributions in r θ z geometry for the J Discrete Ordinate Transport (JSNT) code. Validation on PSC model of Qinshan No.1 nuclear power plant (NPP), CAP1400 and CAP1700 reactors are performed. Numerical results show that the theoretical model and the codes are both correct.
Effects of Shielding on Gamma Rays
Energy Technology Data Exchange (ETDEWEB)
Karpius, Peter Joseph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
2017-03-13
The interaction of gamma rays with matter results in an effect we call attenuation (i.e. ‘shielding’). Attenuation can dramatically alter the appearance of a spectrum. Attenuating materials may actually create features in a spectrum via x-ray fluorescence
Methods for U.S. shielding calculations: applications to FFTF and CRBR designs
International Nuclear Information System (INIS)
Engle, W.W. Jr.; Mynatt, F.R.; Disney, R.K.
1978-01-01
The primary components of the U.S. reactor shielding methodology consist of: (1) computer code systems based on discrete ordinates or Monte Carlo radiation transport calculational methods; (2) a data base of neutron and gamma-ray interaction and gamma-ray-production cross sections used as input in the codes; (3) a capability for processing the cross sections into multigroup or point energy formats as required by the codes; (4) large-scale integral shielding experiments designed to test cross-section data or techniques utilized in the calculations; and (5) a ''sensitivity'' analysis capability that can identify the most important interactions in a transport calculation and assign uncertainties to the calculated result that are based on uncertainties in all of the input data. The required accuracy for the methodology is to within 5 to 10% for responses at locations near the core to within a factor of 2 for responses at distant locations. Under these criteria, the methodology has proved to be adequate for in-vessel LMFBR calculations of neutron transport through deep sodium and thick iron and stainless steel shields, of neutron streaming through lower axial coolant channels and primary pipe chaseways, and of the effects of fuel stored within the reactor vessel. For ex-vessel LMFBR problems, the methodology requires considerable improvement, the areas of concern including neutron streaming through heating and ventilation ducts, through the cavity surrounding the reactor vessel, and through gaps around rotating plugs in the reactor heat, as well as gamma-ray streaming through plant shield penetrations
New gadolinium based glasses for gamma-rays shielding materials
International Nuclear Information System (INIS)
Kaewjang, S.; Maghanemi, U.; Kothan, S.; Kim, H.J.; Limkitjaroenporn, P.; Kaewkhao, J.
2014-01-01
Highlights: • Gd 2 O 3 based glasses have been fabricated and investigated radiation shielding properties between 223 and 662 keV. • Density of the glass increases with increasing of Gd 2 O 3. • All the glasses of Gd 2 O 3 compositions studied had been shown lower HVL than X-rays shielding window. • Prepared glasses to be utilized as radiation shielding material with Pb-free advantage. • This work is the first to reports on radiation shielding properties of Gd 2 O 3 based glass matrices. - Abstract: In this work, Gd 2 O 3 based glasses in compositions (80−x)B 2 O 3 -10SiO 2 -10CaO-xGd 2 O 3 (where x = 15, 20, 25, 30 and 35 mol%) have been fabricated and investigated for their radiation shielding, physical and optical properties. The density of the glass was found to increase with the increasing of Gd 2 O 3 concentration. The experimental values of mass attenuation coefficients (μ m ), effective atomic number (Z eff ) and effective electron densities (N e ) of the glasses were found to increase with the increasing of Gd 2 O 3 concentration and also with the decreasing of photon energy from 223 to 662 keV. The glasses of all Gd 2 O 3 compositions studied have been shown with lower HVL values in comparison to an X-rays shielding window, ordinary concrete and commercial window; indicating their potential as radiation shielding materials with Pb-free advantage. Optical spectra of the glasses in the present study had been shown with light transparency; an advantage when used as radiation shielding materials
New gadolinium based glasses for gamma-rays shielding materials
Energy Technology Data Exchange (ETDEWEB)
Kaewjang, S.; Maghanemi, U.; Kothan, S. [Department of Radiologic Technology, Faculty of Associated Medical Sciences, Chang Mai University, Chang Mai 50200 (Thailand); Kim, H.J. [Department of Physics, Kyungpook National University, Daegu 702-701 (Korea, Republic of); Limkitjaroenporn, P. [Center of Excellence in Glass Technology and Materials Science (CEGM), Nakhon Pathom Rajabhat University, Nakhon Pathom 73000 (Thailand); Kaewkhao, J., E-mail: mink110@hotmail.com [Center of Excellence in Glass Technology and Materials Science (CEGM), Nakhon Pathom Rajabhat University, Nakhon Pathom 73000 (Thailand)
2014-12-15
Highlights: • Gd{sub 2}O{sub 3} based glasses have been fabricated and investigated radiation shielding properties between 223 and 662 keV. • Density of the glass increases with increasing of Gd{sub 2}O{sub 3.} • All the glasses of Gd{sub 2}O{sub 3} compositions studied had been shown lower HVL than X-rays shielding window. • Prepared glasses to be utilized as radiation shielding material with Pb-free advantage. • This work is the first to reports on radiation shielding properties of Gd{sub 2}O{sub 3} based glass matrices. - Abstract: In this work, Gd{sub 2}O{sub 3} based glasses in compositions (80−x)B{sub 2}O{sub 3}-10SiO{sub 2}-10CaO-xGd{sub 2}O{sub 3} (where x = 15, 20, 25, 30 and 35 mol%) have been fabricated and investigated for their radiation shielding, physical and optical properties. The density of the glass was found to increase with the increasing of Gd{sub 2}O{sub 3} concentration. The experimental values of mass attenuation coefficients (μ{sub m}), effective atomic number (Z{sub eff}) and effective electron densities (N{sub e}) of the glasses were found to increase with the increasing of Gd{sub 2}O{sub 3} concentration and also with the decreasing of photon energy from 223 to 662 keV. The glasses of all Gd{sub 2}O{sub 3} compositions studied have been shown with lower HVL values in comparison to an X-rays shielding window, ordinary concrete and commercial window; indicating their potential as radiation shielding materials with Pb-free advantage. Optical spectra of the glasses in the present study had been shown with light transparency; an advantage when used as radiation shielding materials.
International Nuclear Information System (INIS)
Falero, B.; Bueno, P.; Chaves, M. A.; Ordiales, J. M.; Villafana, O.; Gonzalez, M. J.
2013-01-01
The aim of this study was to develop a software application that performs calculation shields in radiology room depending on the type of equipment. The calculation will be done by selecting the user, the method proposed in the Guide 5.11, the Report 144 and 147 and also for the methodology given by the Portuguese Health Ministry. (Author)
Preparation of small group constants for calculation of shielding
International Nuclear Information System (INIS)
Khokhlov, V.F.; Shejno, I.N.; Tkachev, V.D.
1979-01-01
Studied is the effect of the shielding calculation error connected with neglect of the angular and spatial neutron flux dependences while determining the small-group constants on the basis of the many-group ones. The economical method allowing for dependences is proposed. The spatial dependence is substituted by the average value according to the zones singled out in the limits of the zones of the same content; the angular cross section dependence is substituted by the average values in the half-ranges of the angular variable. To solve the transfer equation the ALGOL-ROSA-M program using the method of characteristic interpolation and trial run method is developed. The program regards correctly for nonscattered and single scattered radiations. Compared are the calculation results of neutron transmission (10.5 MeV-0.01 eV) in the 21-group approximation with the 3-group calculations for water (the layer thickness is 30 cm) and 5-group calculations for heterogeneous shielding of alternating stainless steel layers (3 layers, each of the 16 cm thickness) and graphite layers (2 layers, each of the 20 cm thickness). The analysis shows that the method proposed permits to obtain rather accurate results in the course of preparation of the small-group cross sections, decreasing considerably the number of the groups (from 21 to 3-5) and saving the machine time
Resonance self-shielding calculation with regularized random ladders
Energy Technology Data Exchange (ETDEWEB)
Ribon, P.
1986-01-01
The straightforward method for calculation of resonance self-shielding is to generate one or several resonance ladders, and to process them as resolved resonances. The main drawback of Monte Carlo methods used to generate the ladders, is the difficulty of reducing the dispersion of data and results. Several methods are examined, and it is shown how one (a regularized sampling method) improves the accuracy. Analytical methods to compute the effective cross-section have recently appeared: they are basically exempt from dispersion, but are inevitably approximate. The accuracy of the most sophisticated one is checked. There is a neutron energy range which is improperly considered as statistical. An examination is presented of what happens when it is treated as statistical, and how it is possible to improve the accuracy of calculations in this range. To illustrate the results calculations have been performed in a simple case: nucleus /sup 238/U, at 300 K, between 4250 and 4750 eV.
The resonance self-shielding calculation with regularized random ladders
International Nuclear Information System (INIS)
Ribon, P.
1986-01-01
The straightforward method for calculation of resonance self-shielding is to generate one or several resonance ladders, and to process them as resolved resonances. The main drawback of Monte Carlo methods used to generate the ladders, is the difficulty of reducing the dispersion of data and results. Several methods are examined, and it is shown how one (a regularized sampling method) improves the accuracy. Analytical methods to compute the effective cross-section have recently appeared: they are basically exempt from dispersion, but are inevitably approximate. The accuracy of the most sophisticated one is checked. There is a neutron energy range which is improperly considered as statistical. An examination is presented of what happens when it is treated as statistical, and how it is possible to improve the accuracy of calculations in this range. To illustrate the results calculations have been performed in a simple case: nucleus 238 U, at 300 K, between 4250 and 4750 eV. (author)
Calculation of a concrete shielding for an ILU-8 D electron accelerator
International Nuclear Information System (INIS)
Helal, A.; Imam, A.
1996-01-01
A concrete shielding for an electron accelerator of 1 MeV is suggested to replace its structural steel shielding. The thickness of such a shield is calculated. The calculational model used is based on standard and transmission curves given in the literature. The calculated concrete shielding is generally adequate to attenuate the accelerator produced radiation to a level 1 μ Gy/h or less at any point outside of the vault enclosure. 5 figs
Calculation of a concrete shielding for an ILU-8 D electron accelerator
Energy Technology Data Exchange (ETDEWEB)
Helal, A [Nuclear Research Center, AEA, Cairo (Egypt); Imam, A [National Center for Nuclear Safety and Radiation Control, AEA, Cairo (Egypt)
1997-12-31
A concrete shielding for an electron accelerator of 1 MeV is suggested to replace its structural steel shielding. The thickness of such a shield is calculated. The calculational model used is based on standard and transmission curves given in the literature. The calculated concrete shielding is generally adequate to attenuate the accelerator produced radiation to a level 1 {mu} Gy/h or less at any point outside of the vault enclosure. 5 figs.
Evaluation of X-ray shielding performance of protective aprons
International Nuclear Information System (INIS)
Kumagai, Michitomo; Shintani, Mitsuo; Kuranishi, Makoto
1999-01-01
Lead equivalent, which offers protection against x-rays, is rated with a 100 kV tube voltage in Japanese Industrial Standard (JIS) Z 4501-1988, Testing method of lead equivalent for x-ray protective devices.'' However, the actual tube voltage in general diagnostic examinations (normal to special radiography; including computed tomography, CT) is 50 to 150 kV. Therefore, we measured whether the performance of current lead aprons (three products) and protective aprons using composite materials (two products) changes at 60 to 141 kV of tube voltage. Furthermore, we evaluated x-ray shielding performance by measuring the transmission ratio of scattered x-rays. The lead equivalent of two currently used lead aprons was almost the same at all voltages. However, in one currently used lead apron and both protective aprons made of composite materials, lead equivalent decreased rapidly when tube voltage exceeded 100 kV. The transmission ratio of scattered x-rays increased with increasing tube voltage in all of the protective aprons examined. Further, in all aprons examined, the transmission ratio of scattered x-rays declined with widening of the scatter angle. As mentioned above, the x-ray shielding performance of some x-ray protective aprons suddenly decreased at tube voltages over 100 kV. Thus the performance of x-ray protective aprons should be published, and JIS Z 4501 needs to be revised. (author)
Evaluation of X-ray shielding performance of protective aprons
Energy Technology Data Exchange (ETDEWEB)
Kumagai, Michitomo; Shintani, Mitsuo; Kuranishi, Makoto [Toyama Medical and Pharmaceutical Univ. (Japan). Hospital
1999-04-01
Lead equivalent, which offers protection against x-rays, is rated with a 100 kV tube voltage in Japanese Industrial Standard (JIS) Z 4501-1988, Testing method of lead equivalent for x-ray protective devices.`` However, the actual tube voltage in general diagnostic examinations (normal to special radiography; including computed tomography, CT) is 50 to 150 kV. Therefore, we measured whether the performance of current lead aprons (three products) and protective aprons using composite materials (two products) changes at 60 to 141 kV of tube voltage. Furthermore, we evaluated x-ray shielding performance by measuring the transmission ratio of scattered x-rays. The lead equivalent of two currently used lead aprons was almost the same at all voltages. However, in one currently used lead apron and both protective aprons made of composite materials, lead equivalent decreased rapidly when tube voltage exceeded 100 kV. The transmission ratio of scattered x-rays increased with increasing tube voltage in all of the protective aprons examined. Further, in all aprons examined, the transmission ratio of scattered x-rays declined with widening of the scatter angle. As mentioned above, the x-ray shielding performance of some x-ray protective aprons suddenly decreased at tube voltages over 100 kV. Thus the performance of x-ray protective aprons should be published, and JIS Z 4501 needs to be revised. (author)
A Shifting Shield Provides Protection Against Cosmic Rays
Kohler, Susanna
2017-12-01
The Sun plays an important role in protecting us from cosmic rays, energetic particles that pelt us from outside our solar system. But can we predict when and how it will provide the most protection, and use this to minimize the damage to both pilotedand roboticspace missions?The Challenge of Cosmic RaysSpacecraft outside of Earths atmosphere and magnetic field are at risk of damage from cosmic rays. [ESA]Galactic cosmic rays are high-energy, charged particles that originate from astrophysical processes like supernovae or even distant active galactic nuclei outside of our solar system.One reason to care about the cosmic rays arriving near Earth is because these particles can provide a significant challenge for space missions traveling above Earths protective atmosphere and magnetic field. Since impacts from cosmic rays can damage human DNA, this risk poses a major barrier to plans for interplanetary travel by crewed spacecraft. And roboticmissions arent safe either: cosmic rays can flip bits, wreaking havoc on spacecraft electronics as well.The magnetic field carried by the solar wind provides a protective shield, deflecting galactic cosmic rays from our solar system. [Walt Feimer/NASA GSFCs Conceptual Image Lab]Shielded by the SunConveniently, we do have some broader protection against galactic cosmic rays: a built-in shield provided by the Sun. The interplanetary magnetic field, which is embedded in the solar wind, deflects low-energy cosmic rays from us at the outer reaches of our solar system, decreasing the flux of these cosmic rays that reach us at Earth.This shield, however, isnt stationary; instead, it moves and changes as the strength and direction of the solar wind moves and changes. This results in a much lower cosmic-ray flux at Earth when solar activity is high i.e., at the peak of the 11-year solar cycle than when solar activity is low. This visible change in local cosmic-ray flux with solar activity is known as solar modulation of the cosmic ray flux
Validation of calculated self-shielding factors for Rh foils
Jaćimović, R.; Trkov, A.; Žerovnik, G.; Snoj, L.; Schillebeeckx, P.
2010-10-01
Rhodium foils of about 5 mm diameter were obtained from IRMM. One foil had thickness of 0.006 mm and three were 0.112 mm thick. They were irradiated in the pneumatic transfer system and in the carousel facility of the TRIGA reactor at the Jožef Stefan Institute. The foils were irradiated bare and enclosed in small cadmium boxes (about 2 g weight) of 1 mm thickness to minimise the perturbation of the local neutron flux. They were co-irradiated with 5 mm diameter and 0.2 mm thick Al-Au (0.1%) alloy monitor foils. The resonance self-shielding corrections for the 0.006 and 0.112 mm thick samples were calculated by the Monte Carlo simulation and amount to about 10% and 60%, respectively. The consistency of measurements confirmed the validity of self-shielding factors. Trial estimates of Q0 and k0 factors for the 555.8 keV gamma line of 104Rh were made and amount to 6.65±0.18 and (6.61±0.12)×10 -2, respectively.
International Nuclear Information System (INIS)
Lukaszek, W.; Kucypera, S.
1982-01-01
The basis of a semianalytic method for calculating attenuation of rays (neutron, gamma) in material medium is described. The method was applied in determining the neutrons' flux density in one dimensional Cartesian geometry of the reflector and the shield. (author)
Shielding calculations for a 30 MeV proton accelerator
International Nuclear Information System (INIS)
Nandy, Maitreyee; Sarkar, P.K.
2003-01-01
Full text: The thickness of the shield, made of ordinary concrete, required to reduce the equivalent dose rate below the maximum permissible limit and to ensure safe operation of a 30 MeV proton accelerator has been estimated using the Moyer model. Required double differential neutron yield from thick stopping targets has been calculated for several reactions to be used for production of 67 Ga, 111 In, 123 I and 201 Tl radioisotopes. The neutron emission at 0 deg and 90 deg angles with respect to the incident beam direction is estimated using the hybrid model code ALICE91 which considers preequilibrium and equilibrium emissions from the target+projectile composite system. From this neutron yield the equivalent neutron dose rate at unit distance is determined using the ICRP recommended flux-to-dose conversion factors
[About the use of gonad shielding during the taking of hip joint x-rays].
Lempicki, A
1993-01-01
X-rays of hip joints taken without gonadal shield can be found both in the literature and scientific presentations. The shield reduces irradiation of the ovaries 30 times. The shape and the size of the shield has been offered. Organisational measures to use gonadal shields have been suggested.
International Nuclear Information System (INIS)
Ofori, K.
2011-07-01
This research work was conducted to re-evaluate the shielding designs of the 6 MeV x-ray and the 1.253 MeV Co-60 gamma ray scanners used for cargo-containerized scanning at the Port of Tema. These scanners utilize ionizing radiation, therefore adequate shielding must be provided to reduce the radiation exposure of persons in and around the facilities to acceptable levels. The purpose of radiation shielding is to protect workers and the general public from the harmful effects of ionizing radiation. Investigations on the facilities indicated that after commissioning, no work had been carried out to re-evaluate the shielding designs. However, workloads have increased over time neccessitating review of the installed shielding. There has been introduction of scanner units with higher radiation energy (as in the case of the x-ray scanner) posibily increasing dose rates at various location requiring review of the shielding. New structures have been dotted around the facilities without particular attention to their distances and locations with respect to the radiation source. Measurements of distances from the source axes to the points of concern for primary and leakage barrier shielding; source to container and container to the points of concern for scattered radiation shielding were taken. The primary and secondary thicknesses required for both scanners were determined based on current operational parameters and compared with the thickness constituted during the construction of the facilities. Calculated and measured dose rate beyond the shielding barriers were used to established the adequacy or otherwise of the shielding employed by the shielding designers. Values obtained fell below the 20 µSv/hr specified by NCRP 151 (2005) which showed that the primary and secondary shields of both facilities were adequate requiring no additional shielding. (author)
Ab initio calculation of Ti NMR shieldings for titanium oxides and halides
Tossell, J. A.
Titanium NMR shielding constants have been calculated using ab initio coupled Hartree-Fock perturbation theory and polarized double-zeta basis sets for TiF 4, TiF 62-, TiCI 4, Ti(OH) 4, Ti(OH 2) 64+, Ti(OH) 4O, and Ti(OH) 3O -. In all cases the calculations were performed at Hartree-Fuck energy-optimized geometries. For Ti(OH) 4 a S4-symmetry geometry with nonlinear ∠ TiOH was employed. Relative shieldings are in reasonable agreement with experiment for TiF 62-, TiCI 4, and Ti(OR) 4, where R = H or alkyl. Ti(OH 2) 64+ is predicted to be more highly shielded than Ti(OH) 4 by about 340 ppm. The five-coordinate complex Ti(OH) 4O, whose calculated structure matches well that measured by extended X-ray absorption fine structure in K 2O · TiO 2 · SiO 2 glass, is actually deshielded compared to Ti(OH) 4 by about 40 ppm. X-ray absorption-near-edge spectral energies have also been calculated for TiF 4, TiCI 4, Ti(OH) 4, and Ti(OH) 4O using an equivalent ionic core virtual-orbital method and the observed reduction in term energy for the five-coordinate species compared to Ti(OH) 4 has been reproduced. Replacement of the H atoms in Ti(OH) 4 by point charges has only a slight effect upon σTi, suggesting a possible means of incorporating second-neighbor effects in NMR calculations for condensed phases.
Gamma rays shielding parameters for white metal alloys
Kaur, Taranjot; Sharma, Jeewan; Singh, Tejbir
2018-05-01
In the present study, an attempt has been made to check the feasibility of white metal alloys as gamma rays shielding materials. Different combinations of cadmium, lead, tin and zinc were used to prepare quaternary alloys Pb60Sn20ZnxCd20-x (where x = 5, 10, 15) using melt quench technique. These alloys were also known as white metal alloys because of its shining appearance. The density of prepared alloys has been measured using Archimedes Principle. Gamma rays shielding parameters viz. mass attenuation coefficient (µm), effective atomic number (Zeff), electron density (Nel), Mean free path (mfp), Half value layer (HVL) and Tenth value layer (TVL) has been evaluated for these alloys in the wide energy range from 1 keV to 100 GeV. The WinXCom software has been used for obtaining mass attenuation coefficient values for the prepared alloys in the given energy range. The effective atomic number (Zeff) has been assigned to prepared alloys using atomic to electronic cross section ratio method. Further, the variation of various shielding parameters with photon energy has been investigated for the prepared white metal alloys.
Evaluation of RayXpert® for shielding design of medical facilities
Derreumaux, Sylvie; Vecchiola, Sophie; Geoffray, Thomas; Etard, Cécile
2017-09-01
In a context of growing demands for expert evaluation concerning medical, industrial and research facilities, the French Institute for radiation protection and nuclear safety (IRSN) considered necessary to acquire new software for efficient dimensioning calculations. The selected software is RayXpert®. Before using this software in routine, exposure and transmission calculations for some basic configurations were validated. The validation was performed by the calculation of gamma dose constants and tenth value layers (TVL) for usual shielding materials and for radioisotopes most used in therapy (Ir-192, Co-60 and I-131). Calculated values were compared with results obtained using MCNPX as a reference code and with published values. The impact of different calculation parameters, such as the source emission rays considered for calculation and the use of biasing techniques, was evaluated.
International Nuclear Information System (INIS)
Yang, Bo; Qiu, Rui; Li, JunLi; Lu, Wei; Wu, Zhen; Li, Chunyan
2017-01-01
When a strong laser beam irradiates a solid target, a hot plasma is produced and high-energy electrons are usually generated (the so-called “hot electrons”). These energetic electrons subsequently generate hard X-rays in the solid target through the Bremsstrahlung process. To date, only limited studies have been conducted on this laser-induced radiological protection issue. In this study, extensive literature reviews on the physics and properties of hot electrons have been conducted. On the basis of these information, the photon dose generated by the interaction between hot electrons and a solid target was simulated with the Monte Carlo code FLUKA. With some reasonable assumptions, the calculated dose can be regarded as the upper boundary of the experimental results over the laser intensity ranging from 10 19 to 10 21 W/cm 2 . Furthermore, an equation to estimate the photon dose generated from ultraintense laser–solid interactions based on the normalized laser intensity is derived. The shielding effects of common materials including concrete and lead were also studied for the laser-driven X-ray source. The dose transmission curves and tenth-value layers (TVLs) in concrete and lead were calculated through Monte Carlo simulations. These results could be used to perform a preliminary and fast radiation safety assessment for the X-rays generated from ultraintense laser–solid interactions. - Highlights: • The laser–driven X-ray ionizing radiation source was analyzed in this study. • An equation to estimate the photon dose based on the laser intensity is given. • The shielding effects of concrete and lead were studied for this new X-ray source. • The aim of this study is to analyze and mitigate the laser–driven X-ray hazard.
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)
URR-PACK: Calculating Self-Shielding in the Unresolved Resonance Energy Range
International Nuclear Information System (INIS)
Cullen, Dermott E.; Trkov, Andrej
2016-07-01
This report describes HOW to calculate self-shielding in the unresolved resonance region (URR), in terms of the computer codes we provide to allow a user to do these calculations himself. Here we only describe HOW to calculate; a longer companion report describes in detail WHY it is necessary to include URR self-shielding.
Shield design and streaming calculations for the sodium cooled PEC reactor
International Nuclear Information System (INIS)
Prosperi, M.; Tavoni, R.; Travaglini, N.
1977-01-01
This paper summarises the shielding calculations carried out for the PEC reactor. A brief description of calculation methods and of the work carried out to set them up is given; the most representative calculations with the relative isoflux curves are also referred. A general outline is then given for the main shielding problems of the PEC reactor
International Nuclear Information System (INIS)
Borroto Valdes, M.; Saez, D.G.
1992-01-01
Shielding requirements for x-rays diagnostic installations are investigated. The description of an entirely analytical method for calculation of thickness, based in the papers of Simpkin and NCRP49, is presented. Considerations described in specialized method to solving this problem. A program for microcomputer IBM and compatibles ones is available for estimation of minimum shielding requirements in lead, concrete and steel. Similar results were obtained from comparing with others authors
Monte Carlo validation of self shielding and void effect calculations
International Nuclear Information System (INIS)
Tellier, H.; Coste, M.; Raepsaet, C.; Soldevila, M.; Van der Gucht, C.
1995-01-01
The self shielding validation and the void effect are studied with Monte Carlo method. The satisfactory comparison obtained between the APOLLO 2 results of the self shielding effect and the TRIPOLI and MCNP results allows us to be confident in the multigroup transport code. (K.A.)
Evaluation of the Prevalence and Utility of Gonad Shielding in Pediatrics Undergoing Pelvic X-Ray
Vahid Karami; Mansour Zabihzadeh; Nasim Shams; Saman Sarikhani
2016-01-01
Background: Gonad shielding has been recommended during pelvic x-rays since the 1950s. The popular method of gonad shielding is placement a lead shield in the midline of the pelvis. The aim of this study was to evaluate the prevalence and utility of gonad shielding in pediatrics undergoing pelvic x-rays.Materials and MethodsFollowing study approval, we retrospectively retrieved data from the digital image library of ten radiology depertments of Khuzestan provience-Iran to identify pediatric p...
International Nuclear Information System (INIS)
Tsai, Yi-Shan; Liu, Yi-Sheng; Chuang, Ming-Tsung; Wang, Chien-Kuo; Lai, Cheng-Shih; Tsai, Hong-Ming; Lu, Chia-Hsing; Lin, Chii-Jeng
2014-01-01
Patients with developmental dysplasia of the hip (DDH) generally undergo multiple x-ray examinations of both hip joints. During these examinations, the gonads are completely exposed to radiation, unless shielded. Although many types and sizes of gonad shields exist, they often do not provide adequate protection because of size and placement issues; additionally, these shields are frequently omitted for female patients. Our aim was to assess gonad protection during x-ray examination that is provided by gonad shields designed for individual female patients with DDH. We retrospectively retrieved data from the Picture Archiving and Communication System database; pelvic plain x-ray films from 766 females, 18 years old or younger, were included in our analysis. Based on x-ray measurements of the anterior superior iliac spine, we developed a system of gonad shield design that depended on the distance between anterior superior iliac spine markers. We custom-made shields and then examined shielding rates and shielding accuracy before and after these new shields became available. Standard (general-purpose) shields were used before our custom design project was implemented. The shielding rate and shielding accuracy were, respectively, 14.5% and 8.4% before the project was implemented and 72.7% and 32.2% after it was implemented. A shield that is more anatomically correct and available in several different sizes may increase the likelihood of gonad protection during pelvic x-ray examinations. (paper)
Tsai, Yi-Shan; Liu, Yi-Sheng; Chuang, Ming-Tsung; Wang, Chien-Kuo; Lai, Cheng-Shih; Tsai, Hong-Ming; Lin, Chii-Jeng; Lu, Chia-Hsing
2014-12-01
Patients with developmental dysplasia of the hip (DDH) generally undergo multiple x-ray examinations of both hip joints. During these examinations, the gonads are completely exposed to radiation, unless shielded. Although many types and sizes of gonad shields exist, they often do not provide adequate protection because of size and placement issues; additionally, these shields are frequently omitted for female patients. Our aim was to assess gonad protection during x-ray examination that is provided by gonad shields designed for individual female patients with DDH.We retrospectively retrieved data from the Picture Archiving and Communication System database; pelvic plain x-ray films from 766 females, 18 years old or younger, were included in our analysis. Based on x-ray measurements of the anterior superior iliac spine, we developed a system of gonad shield design that depended on the distance between anterior superior iliac spine markers. We custom-made shields and then examined shielding rates and shielding accuracy before and after these new shields became available. Standard (general-purpose) shields were used before our custom design project was implemented. The shielding rate and shielding accuracy were, respectively, 14.5% and 8.4% before the project was implemented and 72.7% and 32.2% after it was implemented. A shield that is more anatomically correct and available in several different sizes may increase the likelihood of gonad protection during pelvic x-ray examinations.
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.
GUI2QAD, Graphical Interface for QAD-CGPIC, Point Kernel for Shielding Calculations
International Nuclear Information System (INIS)
2001-01-01
1 - Description of program or function: GUI2QAD is an aid in preparation of input for the included QAD-CGPIC program, which is based on CCC-493/QAD-CGGP and PICTURE. QAD-CGPIC, which is included in this distribution, is a Fortran code for neutron and gamma-ray shielding calculations by the point kernel method. Provision is available to interactively view the geometry of the system. QAD-CG calculates fast-neutron and gamma-ray penetration through various shield configurations defined by combinatorial geometry specifications. The code can use the ANS-6.4.3 1990 buildup factor compilation (26 materials). 2 - Methods:The code QAD-CGPIC is based on point kernel method and has a provision to select either GP or Capo's build up factors. 3 - Restrictions on the complexity of the problem: Details on restrictions and limitations are available in the RSICC code manual CCC-493/QAD-CGGP. Because CCC-493 was obsoleted by CCC-645/QAD-CGGP-A, the CCC-493 documentation is not online but is included with this package. This package includes a Graphical User Interface to facilitate use
MO-D-213-07: RadShield: Semi- Automated Calculation of Air Kerma Rate and Barrier Thickness
International Nuclear Information System (INIS)
DeLorenzo, M; Wu, D; Rutel, I; Yang, K
2015-01-01
Purpose: To develop the first Java-based semi-automated calculation program intended to aid professional radiation shielding design. Air-kerma rate and barrier thickness calculations are performed by implementing NCRP Report 147 formalism into a Graphical User Interface (GUI). The ultimate aim of this newly created software package is to reduce errors and improve radiographic and fluoroscopic room designs over manual approaches. Methods: Floor plans are first imported as images into the RadShield software program. These plans serve as templates for drawing barriers, occupied regions and x-ray tube locations. We have implemented sub-GUIs that allow the specification in regions and equipment for occupancy factors, design goals, number of patients, primary beam directions, source-to-patient distances and workload distributions. Once the user enters the above parameters, the program automatically calculates air-kerma rate at sampled points beyond all barriers. For each sample point, a corresponding minimum barrier thickness is calculated to meet the design goal. RadShield allows control over preshielding, sample point location and material types. Results: A functional GUI package was developed and tested. Examination of sample walls and source distributions yields a maximum percent difference of less than 0.1% between hand-calculated air-kerma rates and RadShield. Conclusion: The initial results demonstrated that RadShield calculates air-kerma rates and required barrier thicknesses with reliable accuracy and can be used to make radiation shielding design more efficient and accurate. This newly developed approach differs from conventional calculation methods in that it finds air-kerma rates and thickness requirements for many points outside the barriers, stores the information and selects the largest value needed to comply with NCRP Report 147 design goals. Floor plans, parameters, designs and reports can be saved and accessed later for modification and recalculation
MO-D-213-07: RadShield: Semi- Automated Calculation of Air Kerma Rate and Barrier Thickness
Energy Technology Data Exchange (ETDEWEB)
DeLorenzo, M [Oklahoma University Health Sciences Center, Oklahoma City, OK (United States); Wu, D [University of Oklahoma Health Sciences Center, Oklahoma City, Ok (United States); Rutel, I [University of Oklahoma Health Science Center, Oklahoma City, OK (United States); Yang, K [Massachusetts General Hospital, Boston, MA (United States)
2015-06-15
Purpose: To develop the first Java-based semi-automated calculation program intended to aid professional radiation shielding design. Air-kerma rate and barrier thickness calculations are performed by implementing NCRP Report 147 formalism into a Graphical User Interface (GUI). The ultimate aim of this newly created software package is to reduce errors and improve radiographic and fluoroscopic room designs over manual approaches. Methods: Floor plans are first imported as images into the RadShield software program. These plans serve as templates for drawing barriers, occupied regions and x-ray tube locations. We have implemented sub-GUIs that allow the specification in regions and equipment for occupancy factors, design goals, number of patients, primary beam directions, source-to-patient distances and workload distributions. Once the user enters the above parameters, the program automatically calculates air-kerma rate at sampled points beyond all barriers. For each sample point, a corresponding minimum barrier thickness is calculated to meet the design goal. RadShield allows control over preshielding, sample point location and material types. Results: A functional GUI package was developed and tested. Examination of sample walls and source distributions yields a maximum percent difference of less than 0.1% between hand-calculated air-kerma rates and RadShield. Conclusion: The initial results demonstrated that RadShield calculates air-kerma rates and required barrier thicknesses with reliable accuracy and can be used to make radiation shielding design more efficient and accurate. This newly developed approach differs from conventional calculation methods in that it finds air-kerma rates and thickness requirements for many points outside the barriers, stores the information and selects the largest value needed to comply with NCRP Report 147 design goals. Floor plans, parameters, designs and reports can be saved and accessed later for modification and recalculation
Code system BCG for gamma-ray skyshine calculation
International Nuclear Information System (INIS)
Ryufuku, Hiroshi; Numakunai, Takao; Miyasaka, Shun-ichi; Minami, Kazuyoshi.
1979-03-01
A code system BCG has been developed for calculating conveniently and efficiently gamma-ray skyshine doses using the transport calculation codes ANISN and DOT and the point-kernel calculation codes G-33 and SPAN. To simplify the input forms to the system, the forms for these codes are unified, twelve geometric patterns are introduced to give material regions, and standard data are available as a library. To treat complex arrangements of source and shield, it is further possible to use successively the code such that the results from one code may be used as input data to the same or other code. (author)
Experimental investigation of clay fly ash bricks for gamma-ray shielding
Energy Technology Data Exchange (ETDEWEB)
Mann Harjinder Singh; Mudahar, Gumel Singh [Dept. of Physics, Punjabi University, Patiala (India); Brar, Gurdarshan Singh [Dept. of Higher Education, Additional Project Director, Chandigarh (India); Mann, Kulwinder Singh [Dept. of Applied Sciences, I.K. Gujral Punjab Technical University, Jalandhar (India)
2016-10-15
This study aims to determine the effect of fly ash with a high replacing ratio of clay on the radiation shielding properties of bricks. Some interaction parameters (mass attenuation coefficients, half value layer, effective atomic number, effective electron density, and absorption efficiency) of clay fly ash bricks were measured with a NaI(Tl) detector at 661.6 keV, 1,173.2 keV, and 1,332.5 keV. For the investigation of their shielding behavior, fly ash bricks were molded using an admixture to clay. A narrow beam transmission geometry condition was used for the measurements. The measured values of these parameters were found in good agreement with the theoretical calculations. The elemental compositions of the clay fly ash bricks were analyzed by using an energy dispersive X-ray fluorescence spectrometer. At selected energies the values of the effective atomic numbers and effective electron densities showed a very modest variation with the composition of the fly ash. This seems to be due to the similarity of their elemental compositions. The obtained results were also compared with concrete, in order to study the effect of fly ash content on the radiation shielding properties of clay fly ash bricks. The clay fly ash bricks showed good shielding properties for moderate energy gamma rays. Therefore, these bricks are feasible and eco-friendly compared with traditional clay bricks used for construction.
Calculation of shielding needed at the wall where cobalt therapy unit Alcyon II is installed
International Nuclear Information System (INIS)
Morales M, F.
1994-01-01
Calculations of shielding at the wall to avoid scattering radiation for the personnel and population were performed. The position of the shielding door was corrected because before it had been placed in front of the beams, thus producing excessive radiation to the operator and personnel and patients. The calculations were based on the German standard (DIN)
A study on the calculation of the shielding wall thickness in medical linear accelerator
Energy Technology Data Exchange (ETDEWEB)
Lee, Dong Yeon [Dept. of Radiation Oncology, Dongnam Ins. of Radiological and Medical Science, Busan (Korea, Republic of); Park, Eun Tae [Dept. of Radiation Oncology, Inje University Busan Paik Hospital, Busan (Korea, Republic of); Kim, Jung Hoon [Dept. of Radiological science, college of health sciences, Catholic University of Pusan, Busan (Korea, Republic of)
2017-06-15
The purpose of this study is to calculate the thickness of shielding for concrete which is mainly used for radiation shielding and study of the walls constructed to shield medical linear accelerator. The optimal shielding thickness was calculated using MCNPX(Ver.2.5.0) for 10 MV of photon beam energy generated by linear accelerator. As a result, the TVL for photon shielding was formed at 50⁓100 cm for pure concrete and concrete with Boron+polyethylene at 80⁓100 cm. The neutron shielding was calculated 100⁓140 cm for pure concrete and concrete with Boron+polyethylene at 90⁓100 cm. Based on this study, the concrete is considered to be most efficient method of using steel plates and adding Boron+polyethylene th the concrete.
The shielding calculation for the CN guide shielding assembly in HANARO
International Nuclear Information System (INIS)
Kim, H. S.; Lee, B. C.; Lee, K. H.; Kim, H.
2006-01-01
The cold neutron research facility in HANARO is under construction. The area including neutron guides and rotary shutter in the reactor hall should be shielded by the guide shielding assembly which is constructed of heavy concrete blocks and structure. The guide shielding assembly is divided into 2 parts, A and B. Part A is about 6.4 meters apart from the reactor biological shield and it is constructed of heavy concrete blocks whose density is above 4.0g/cm 3 . And part B is a fixed heavy concrete structure whose density is above 3.5g/cm 3 . The rotary shutter is also made with heavy concrete whose density is above 4.0g/cm 3 and includes 5 neutron guides inside. It can block the neutron beam by rotating when CNS is not operating. The dose criterion outside the guide shielding assembly is established as 12.5 μSv/hr which is also applied to reactor shielding in HANARO
Zughbi, A.; Kharita, M. H.; Shehada, A. M.
2017-07-01
A new method of recycling glass of Cathode Ray Tubes (CRTs) has been presented in this paper. The glass from CRTs suggested being used as raw materials for the production of radiation shielding glass. Cathode ray tubes glass contains considerable amounts of environmentally hazardous toxic wastes, namely heavy metal oxides such as lead oxide (PbO). This method makes CRTs glass a favorable choice to be used as raw material for Radiation Shielding Glass and concrete. The heavy metal oxides increase its density, which make this type of glass nearly equivalent to commercially available shielding glass. CRTs glass have been characterized to determine heavy oxides content, density, refractive index, and radiation shielding properties for different Gamma-Ray energies. Empirical methods have been used by using the Gamma-Ray source cobalt-60 and computational method by using the code XCOM. Measured and calculated values were in a good compatibility. The effects of irradiation by gamma rays of cobalt-60 on the optical transparency for each part of the CRTs glass have been studied. The Results had shown that some parts of CRTs glass have more resistant to Gamma radiation than others. The study had shown that the glass of cathode ray tubes could be recycled to be used as radiation shielding glass. This proposed use of CRT glass is only limited to the available quantity of CRT world-wide.
Mercuric Iodide Anticoincidence Shield for Gamma-Ray Spectrometer, Phase I
National Aeronautics and Space Administration — We propose to utilize a new detector material, polycrystalline mercuric iodide, for background suppression by active anticoincidence shielding in gamma-ray...
Mercuric Iodide Anticoincidence Shield for Gamma-Ray Spectrometer, Phase II
National Aeronautics and Space Administration — We utilize a new detector material, polycrystalline mercuric iodide, for background suppression by active anticoincidence shielding in gamma-ray spectrometers. Two...
The calculation of some gamma shielding parameters for semiconductor CsPbBr3
Oto, Berna; Gulebaglan, Sinem Erden; Kanberoglu, Gulsah Saydan
2017-02-01
Recently, researchers produced perovskites structures used in optoelectronic devices as substrates, sensors. CsPbBr3 crystal is found in the cubic perovskite structure and its space group is Pm-3m. CsPbBr3 is a developing material for detection of X- and γ-ray radiations and the knowledge of the attenuation parameters of CsPbBr3 crystal is important. In this study, some photon shielding parameters such as mass attenuation coefficient (μρ), effective atomic number (Zeff) and electron density (Nel) have been investigated for CsPbBr3 compound. The theoretical values of μρ have been calculated in the energy range from 1 keV to 100 GeV using WinXCom computer code and these values have been used in order to calculate the values of Zeff and Nel in the same energy range.
Deep-penetration calculations in concrete and iron for shielding of proton therapy accelerators
International Nuclear Information System (INIS)
Sheu, Rong-Jiun; Chen, Yen-Fu; Lin, Uei-Tyng; Jiang, Shiang-Huei
2012-01-01
Proton accelerators in the energy range of approximately 200 MeV have become increasingly popular for cancer treatment in recent years. These proton therapy facilities usually involve bulky concrete or iron in their shielding design or accelerator structure. Simple shielding data, such as source terms or attenuation lengths for various proton energies and materials are useful in designing accelerator shielding. Understanding the appropriateness or uncertainties associated with these data, which are largely generated from Monte Carlo simulations, is critical to the quality of a shielding design. This study demonstrated and investigated the problems of deep-penetration calculations on the estimation of shielding parameters through an extensive comparison between the FLUKA and MCNPX calculations for shielding against a 200-MeV proton beam hitting an iron target. Simulations of double-differential neutron production from proton bombardment were validated by comparison with experimental data. For the concrete shielding, the FLUKA calculated depth–dose distributions were consistent with the MCNPX results, except for some discrepancies in backward directions. However, for the iron shielding, if FLUKA is used inappropriately then overestimation of neutron attenuation can be expected as shown by this work because of the multigroup treatment for low-energy neutrons in FLUKA. Two neutron energy group structures, three degrees of self-shielding correction, and two iron compositions were considered in this study. Significant variation of the resulting attenuation lengths indicated the importance of problem-dependent multigroup cross sections and proper modeling of iron composition in deep-penetration calculations.
Semi-analytic flux formulas for shielding calculations
International Nuclear Information System (INIS)
Wallace, O.J.
1976-06-01
A special coordinate system based on the work of H. Ono and A. Tsuro has been used to derive exact semi-analytic formulas for the flux from cylindrical, spherical, toroidal, rectangular, annular and truncated cone volume sources; from cylindrical, spherical, truncated cone, disk and rectangular surface sources; and from curved and tilted line sources. In most of the cases where the source is curved, shields of the same curvature are allowed in addition to the standard slab shields; cylindrical shields are also allowed in the rectangular volume source flux formula. An especially complete treatment of a cylindrical volume source is given, in which dose points may be arbitrarily located both within and outside the source, and a finite cylindrical shield may be considered. Detector points may also be specified as lying within spherical and annular source volumes. The integral functions encountered in these formulas require at most two-dimensional numeric integration in order to evaluate the flux values. The classic flux formulas involving only slab shields and slab, disk, line, sphere and truncated cone sources become some of the many special cases which are given in addition to the more general formulas mentioned above
International Nuclear Information System (INIS)
Fukuda, Atsushi; Takahashi, Masaaki; Kitabayashi, Keitarou; Koshida, Kichiro; Matsubara, Kousuke; Noto, Kimiya; Nakagawa, Hiroto; Kawabata, Chikako
2004-01-01
Various pharmaceutical companies in Japan are making radioactive drugs available for positron emission tomography (PET) in hospitals without a cyclotron. With the distribution of these drugs to hospitals, medical check-ups and examinations using PET are expected to increase. However, the safety guidelines for radiation in the new deployment of PET have not been adequately improved. Therefore, we measured the shielding effect of a clinical X-ray protector and lead glass against annihilation radiation and gamma rays of 99m Tc. We then calculated the shielding effect of a 0.25 mm lead protector, 1 mm lead, and lead glass using the EGS4 (Electron Gamma Shower Version 4) code. The shielding effects of 22-mm lead glass against annihilation radiation and gamma rays of 99m Tc were approximately 31.5% and 93.3%, respectively. The clinical X-ray protector against annihilation radiation approximately doubled the skin-absorbed dose. (author)
Fukuda, Atsushi; Koshida, Kichiro; Yamaguchi, Ichiro; Takahashi, Masaaki; Kitabayashi, Keitarou; Matsubara, Kousuke; Noto, Kimiya; Kawabata, Chikako; Nakagawa, Hiroto
2004-12-01
Various pharmaceutical companies in Japan are making radioactive drugs available for positron emission tomography (PET) in hospitals without a cyclotron. With the distribution of these drugs to hospitals, medical check-ups and examinations using PET are expected to increase. However, the safety guidelines for radiation in the new deployment of PET have not been adequately improved. Therefore, we measured the shielding effect of a clinical X-ray protector and lead glass against annihilation radiation and gamma rays of (99m)Tc. We then calculated the shielding effect of a 0.25 mm lead protector, 1 mm lead, and lead glass using the EGS4 (Electron Gamma Shower Version 4) code. The shielding effects of 22-mm lead glass against annihilation radiation and gamma rays of (99m)Tc were approximately 31.5% and 93.3%, respectively. The clinical X-ray protector against annihilation radiation approximately doubled the skin-absorbed dose.
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.
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.
Evaluation of the Prevalence and Utility of Gonad Shielding in Pediatrics Undergoing Pelvic X-Ray
Directory of Open Access Journals (Sweden)
Vahid Karami
2016-11-01
Full Text Available Background: Gonad shielding has been recommended during pelvic x-rays since the 1950s. The popular method of gonad shielding is placement a lead shield in the midline of the pelvis. The aim of this study was to evaluate the prevalence and utility of gonad shielding in pediatrics undergoing pelvic x-rays.Materials and MethodsFollowing study approval, we retrospectively retrieved data from the digital image library of ten radiology depertments of Khuzestan provience-Iran to identify pediatric patients who underwent pelvic x-ray (anteriorposterior [AP] view. All the images were reviewed for the probable evidence of gonad shield. If there was evidence of shielding, the accuracy positioning of the shield was also investigated by a single assistant radiologist.ResultsIn all 1745 pelvic x-rays (942 girls and 803 boys were identified of which the shield was present in 51 (5.41% radiographs of girls and 132 (16.43% radiographs of boys. When a shield was present; the shields has adequate positioning only in 8 (15.68% radiographs in girls and 59 radiographs in boys. Inaccurate placement and absence of gonad shields were more common in girls than the boys. Due to the shield has concealed the anatomical criteria of the pelvis, retakes of the examination was required in 11 (21.56% radiographs of girls and 14 (10.6% radiographs of boys.ConclusionThe current methods of gonad shielding in girls pelvic x-ray was not effective nor is justifiable. We no longer advocate of gonad shielding during girls pelvic x-ray. However in boys it is controversial and depends on the skill and effort of radiographers.
Systems for neutronic, thermohydraulic and shielding calculation in personal computers
International Nuclear Information System (INIS)
Villarino, E.A.; Abbate, P.; Lovotti, O.; Santini, M.
1990-01-01
The MTR-PC (Materials Testing Reactors-Personal Computers) system has been developed by the Nuclear Engineering Division of INVAP S.E. with the aim of providing working conditions integrated with personal computers for design and neutronic, thermohydraulic and shielding analysis for reactors employing plate type fuel. (Author) [es
RASH D - A mercury programme for neutron shielding calculations
International Nuclear Information System (INIS)
Bendall, D.E.
1962-08-01
An improved version of an earlier neutron shielding programme (RASH B) is described. The new programme is also written in Mercury Autocode and solves a set of multigroup diffusion equations in one dimension. It differs from RASH B in that distributed source terms may be introduced into all the groups if required. Some other improvements are also included. (author)
The elimination of ray tracing in Monte Carlo shielding programs
International Nuclear Information System (INIS)
Bendall, D.E.
1988-01-01
The MONK6 code has clearly demonstrated the advantages of hole tracking, which was devised by Woodcock et at. for use in criticality codes from earlier work by Von Neumann. Hole tracking eliminates ray tracing by introducing, for all materials present in the problem, a pseudo scattering reaction that forward scatters without energy loss. The cross section for this reaction is chosen so that the total cross sections for all the materials are equal at a given energy. By this means, tracking takes place with a constant total cross section everywhere, so there is now no need to ray trace. The present work extends hole tracking to shielding codes, where it functions in tandem with Russian roulette and splitting. An algorithm has been evolved and its performance is compared with the ray-tracking code McBEND. A disadvantage with hole tracking occurs when there is a wide variation in total cross section for materials present. As the tracking uses the total cross section of the material that has the maximum cross section, there can be a large number of pseudo collisions in the materials with low total cross sections. In extreme cases, the advantages of hole tracking can be lost by the by the extra time taken in servicing these pseudo collisions; however, techniques for eliminating this problem are under consideration
The problem of resonance self-shielding effect in neutron multigroup calculations
International Nuclear Information System (INIS)
Wang Qingming; Huang Jinghua
1991-01-01
It is not allowed to neglect the resonance self-shielding effect in hybrid blanket and fast reactor neutron designs. The authors discussed the importance as well as the method of considering the resonance self-shielding effect in hybrid blanket and fast reactor neutron multigroup calculations
Shielding Calculations for Industrial 5/7.5MeV Electron Accelerators Using the MCNP Monte Carlo Code
International Nuclear Information System (INIS)
Peri, E.; Orion, I.
2014-01-01
High energy X-rays from accelerators are used to irradiate food ingredients to prevent growth and development of unwanted biological organisms in food, in order to extend the shelf life of products. High energy photons can cause food activation due to (D 3 ,n) reactions. Until 2004, to eliminate the possibility of food activation, the electron energy was limited to 5 MeV X-rays for food irradiation. In 2004, the FDA approved the usage of up to 7.5 MeV, but only with tantalum and gold targets (1). Higher X-ray energy results an increased flux of X-rays in the forward direction, increased penetration, and higher photon dose rate due to better electron-to-photon conversion. These improvements could decrease the irradiation time and allow irradiation of larger packages, thereby providing higher production rates with lower treatment cost. Medical accelerators usually work with 6-18 MV electron energy with tungsten target to convert the electron beam to X-rays. In order to protect the patients, the accelerator head is protected with a heavy lead shielding; therefore, the bremsstrahlung is emitted only in the forward direction. There are many publications and standards that guide how to design optimal shielding for medical accelerator rooms. The shielding data for medical accelerators is not applicable for industrial accelerators, since the data is for different conversion targets, different X-Ray energies, and only for the forward direction. Collimators are not always in use in industrial accelerators, and therefore bremsstrahlung photons can be emitted in all directions. The bremsstrahlung spectrum and dose rate change as a function of the emission angle. The dose rate decreases from maximum in the forward direction (0°) to minimum at 180° by 1-2 orders of magnitude. In order to design and calculate optimal shielding for food accelerator rooms, there is a need to have the bremsstrahlung spectrum data, dose rates and concrete attenuation data in all emission directions
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.)
Energy Technology Data Exchange (ETDEWEB)
Gonzalez Portilla, M. I.; Marquez, J.
2011-07-01
Radiological protection aims to limit the ionizing radiation received by people and equipment, which in numerous occasions requires of protection shields. Although, for certain configurations, there are analytical formulas, to characterize these shields, the design setup may be very intensive in numerical calculations, therefore the most efficient from to design the shields is by means of computer programs to calculate dose and dose rates. In the present article we review the codes most frequently used to perform these calculations, and the techniques used by such codes. (Author) 13 refs.
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)
A versatile program for the calculation of linear accelerator room shielding.
Hassan, Zeinab El-Taher; Farag, Nehad M; Elshemey, Wael M
2018-03-22
This work aims at designing a computer program to calculate the necessary amount of shielding for a given or proposed linear accelerator room design in radiotherapy. The program (Shield Calculation in Radiotherapy, SCR) has been developed using Microsoft Visual Basic. It applies the treatment room shielding calculations of NCRP report no. 151 to calculate proper shielding thicknesses for a given linear accelerator treatment room design. The program is composed of six main user-friendly interfaces. The first enables the user to upload their choice of treatment room design and to measure the distances required for shielding calculations. The second interface enables the user to calculate the primary barrier thickness in case of three-dimensional conventional radiotherapy (3D-CRT), intensity modulated radiotherapy (IMRT) and total body irradiation (TBI). The third interface calculates the required secondary barrier thickness due to both scattered and leakage radiation. The fourth and fifth interfaces provide a means to calculate the photon dose equivalent for low and high energy radiation, respectively, in door and maze areas. The sixth interface enables the user to calculate the skyshine radiation for photons and neutrons. The SCR program has been successfully validated, precisely reproducing all of the calculated examples presented in NCRP report no. 151 in a simple and fast manner. Moreover, it easily performed the same calculations for a test design that was also calculated manually, and produced the same results. The program includes a new and important feature that is the ability to calculate required treatment room thickness in case of IMRT and TBI. It is characterised by simplicity, precision, data saving, printing and retrieval, in addition to providing a means for uploading and testing any proposed treatment room shielding design. The SCR program provides comprehensive, simple, fast and accurate room shielding calculations in radiotherapy.
International Nuclear Information System (INIS)
Zughbi, A.; Kharita, M.H.; Shehada, A.M.
2017-01-01
A new method of recycling glass of Cathode Ray Tubes (CRTs) has been presented in this paper. The glass from CRTs suggested being used as raw materials for the production of radiation shielding glass. Cathode ray tubes glass contains considerable amounts of environmentally hazardous toxic wastes, namely heavy metal oxides such as lead oxide (PbO). This method makes CRTs glass a favorable choice to be used as raw material for Radiation Shielding Glass and concrete. The heavy metal oxides increase its density, which make this type of glass nearly equivalent to commercially available shielding glass. CRTs glass have been characterized to determine heavy oxides content, density, refractive index, and radiation shielding properties for different Gamma-Ray energies. Empirical methods have been used by using the Gamma-Ray source cobalt-60 and computational method by using the code XCOM. Measured and calculated values were in a good compatibility. The effects of irradiation by gamma rays of cobalt-60 on the optical transparency for each part of the CRTs glass have been studied. The Results had shown that some parts of CRTs glass have more resistant to Gamma radiation than others. The study had shown that the glass of cathode ray tubes could be recycled to be used as radiation shielding glass. This proposed use of CRT glass is only limited to the available quantity of CRT world-wide. - Highlights: • A new method of recycling glass of Cathode Ray Tubes (CRTs) has been presented. • The glass from CRTs used as raw materials for radiation shielding glass. • The resulted glass have good optical properties and stability against radiations.
Shielding Calculations for Positron Emission Tomography - Computed Tomography Facilities
Energy Technology Data Exchange (ETDEWEB)
Baasandorj, Khashbayar [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Yang, Jeongseon [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)
2015-10-15
Integrated PET-CT has been shown to be more accurate for lesion localization and characterization than PET or CT alone, and the results obtained from PET and CT separately and interpreted side by side or following software based fusion of the PET and CT datasets. At the same time, PET-CT scans can result in high patient and staff doses; therefore, careful site planning and shielding of this imaging modality have become challenging issues in the field. In Mongolia, the introduction of PET-CT facilities is currently being considered in many hospitals. Thus, additional regulatory legislation for nuclear and radiation applications is necessary, for example, in regulating licensee processes and ensuring radiation safety during the operations. This paper aims to determine appropriate PET-CT shielding designs using numerical formulas and computer code. Since presently there are no PET-CT facilities in Mongolia, contact was made with radiological staff at the Nuclear Medicine Center of the National Cancer Center of Mongolia (NCCM) to get information about facilities where the introduction of PET-CT is being considered. Well-designed facilities do not require additional shielding, which should help cut down overall costs related to PET-CT installation. According to the results of this study, building barrier thicknesses of the NCCM building is not sufficient to keep radiation dose within the limits.
Calculation of the neutrons shielding in cyclotron accelerator
International Nuclear Information System (INIS)
Ribeiro, Martha S.; Sanches, Matias P.; Rodrigues, Demerval L.
2000-01-01
The objective of radioprotection in cyclotron facilities is to reduce the dose levels in the workplaces to classify them like supervised areas. In this way, the radiation dose rates in areas occupied by workers during cyclotron operations should not exceed 7,5 μSv/h. In controlled areas these levels are not observed and some rigorous controls must be exerted by administrative procedures or protection mechanisms. The Cyclotron Laboratory at IPEN-CNEN/SP has a cyclotron model Cyclone 30, 30 MeV, used for research and it is also used for radioisotopes production for medical diagnosis and therapeutical applications. Among them, 123 I, 67 Ga and 18 F can be pointed. When accelerator is operating, failures in perforations and paths that conduce to room accelerator can be occur and thus, the dose levels are higher than that established by law. For this reason, a review for shielding structure was necessary in order to optimize radiation dose. The purpose of this work was to determine the shielding thickness and adequate material to diminish the dose rates in workplaces to a value below 7,5 μSv/h. It was used a method to employ the equivalent dose value in the facility areas for neutrons fluency rate for the principal reactions in target irradiation processes. The purposed shielding for the vault doors ensures dose levels lower than established limits to supervised areas. (author)
A method for assessing the structural shielding in diagnostic x-ray facilities
International Nuclear Information System (INIS)
Almeida, F.; Martinez de la Fuente, O.; Perez, C.
1992-01-01
The design of each X-Ray medical facility involves, in order to guarantee the optimun levels of Radiologic safety for everybody who could be exposed during the performance of the examinations the need of assessing the required shieldings for the room which contains the emiter tubes. In such sense, this paper gives a number of criteria to calculate the structural requirements for the diverse configurations which exist in Health Centres using the method proposed by NCRP in its Report 49 as a reference. (author)
International Nuclear Information System (INIS)
Hill, Robin; Healy, Brendan; Holloway, Lois; Baldock, Clive
2007-01-01
Kilovoltage x-ray beams are used to treat cancer on or close to the skin surface. Many clinical cases use high atomic number materials as shielding to reduce dose to underlying healthy tissues. In this work, we have investigated the effect on both the surface dose and depth doses in a water phantom with lead shielding at depth in the phantom. The EGSnrc Monte Carlo code was used to simulate the water phantom and to calculate the surface doses and depth doses using primary x-ray beam spectra derived from an analytical model. The x-ray beams were in the energy range of 75-135 kVp with field sizes of 2, 5 and 8 cm diameter. The lead sheet was located beneath the water surface at depths ranging from 0.5-7.5 cm. The surface dose decreased as the lead was positioned closer to the water surface and as the field size was increased. The variation in surface dose as a function of x-ray beam energy was only small but the maximum reduction occurred for the 100 kVp x-ray beam. For the 8 cm diameter field with the lead at 1 cm depth and using the 100 kVp x-ray beam, the surface dose was reduced to 0.898 of the surface dose in the water phantom only. Measured surface dose changes, using a Farmer-type ionization chamber, agreed with the Monte Carlo calculated doses. Calculated depth doses in water with a lead sheet positioned below the surface showed that the dose fall-off increased as the lead was positioned closer to the water surface as compared to the depth dose in the water phantom only. Monte Carlo calculations of the total x-ray beam spectrum at the water surface showed that the total fluence decreased due to a reduction in backscatter from within the water and very little backscatter from the lead. The mean energy of the x-ray spectrum varied less than 1 keV, with the lead at 1 cm beneath the water phantom surface. As the Monte Carlo calculations showed good agreement with the measured results, this method can be used to verify surface dose changes in clinical situations
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)
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.
Studying the shielding properties of lead glass composites using neutrons and gamma rays
International Nuclear Information System (INIS)
Osman, A.M.; El-Sarraf, M.A.; Abdel-Monem, A.M.; El-Sayed Abdo, A.
2015-01-01
Highlights: • Samples of sodalime silica glass loaded with different ratios of PbO were prepared. • Leaded glass composites were investigated for radiation shielding. • Experimental and theoretical attenuation parameters were studied. • Experimental and theoretical (MCNP5) results were in good agreement. - Abstract: The present work deals with the shielding properties of lead glass composites to find out its integrity for practical shielding applications and radiological safety. Composites of different lead oxide ratios (x = 0, 5, 10, 15 and 25 wt.%) have been prepared by the Nasser Glass and Crystal Company (Egypt). Attenuation measurements have been carried out using a collimated emitted beam from a fission 252 Cf (100 μg) neutron source, and the neutron–gamma spectrometer with stilbene scintillator. The pulse shape discriminating (P.S.D.) technique based on the zero cross-over method was used to discriminate between neutron and gamma-ray pulses. Thermal neutron fluxes were measured using the BF3 detector and thermal neutron detection system. The attenuation relations were used to evaluate fast neutron macroscopic effective removal cross-section Σ R-Meas (cm −1 ), gamma rays total attenuation coefficient μ (cm −1 ) and thermal neutron macroscopic cross-section Σ Meas (cm −1 ). Theoretical calculations have been achieved using MCNP5 code to calculate the same two parameters. Also, MERCSF-N program was used to calculate fast neutron macroscopic removal cross-section Σ R-MER (cm −1 ). Measured and MCNP5 calculated results have been compared and were found to be in reasonable agreement
Assessment of shielding integrity of Co-60 gamma irradiation-ray scanner at Aflao Border, Ghana
International Nuclear Information System (INIS)
Agbemafo, Edwin Capacity
2016-07-01
This study examines the current state of the shielding integrity of the 38.7 TBq Co-60 gamma ray scanner with an average energy of 1.25 MeV operated by NICK TC Scan Limited, which has been in use for destination inspection at Aflao Border of Ghana, for the past six years, (2010-2016). The facility uses a high energy ionizing radiation in its operation; therefore continuous adequacy of the installed biological shielding is critical to the protection and safety of the workers and the general public. The workload of the facility has increased since its commissioning, requiring the review of the status of the installed shielding. Theoretical calculations for dose rates and barrier thicknesses based on tenth – value layer (TVL) concept and NCRP 151, 2005 recommendations, were done around the scanning facility using the current operational data. The results were then compared with the measured dose rates and the shielding thickness constituted during the commissioning stage, and international standards. Calculated dose rate at commissioning state ranges from 0.6μSv/hr to 2.4μSv/hr with an average dose rate of 1.43μSv/hr and that of the current operational state ranges from 1.1μSv/hr to 2.6μSv/hr with an average dose rate of 1.54μSv/hr, indicating an increase of 7.9%. Even though the dose rates were all below the recommended dose limit of 20μSvh"-"1 by NCRP, there has been an increase in dose to the staff and the general public. It has been observed that, the workload has increased three-fold from the commissioning stage to current operational state over the past six years. The assessment done on the installed shielding using the current operational data indicates that the shielding is inadequate in providing protection for the general public and the workers against X-ray radiation source of energy of at least 6MeV, and therefore the facility in its current state cannot be used to house a linear accelerator of energy up to 10MeV. (au)
Shielding calculations for the design of neutron radiography facility around PARR
International Nuclear Information System (INIS)
Ashraf, M.M.; Khan, A.R.
1989-06-01
Shielding calculations for neutron radiography facility, proposed to be established around PARR have been carried out using two group diffusion theory and shielding formulae. Gamma radiation penetration calculations have been carried out using simple attenuation methods. The fabrication and installation of the neutron radiography facility would provide the basis for designing a better collimating system and would help establish under water radiography facility for the inspection of highly radioactive materials and components etc. (orig./A.B.)
DEFF Research Database (Denmark)
Ruud, Kenneth; Helgaker, Trygve; Kobayashi, Rika
1994-01-01
to corresponding individual gauges for localized orbitals (IGLO) results. The London results show better basis set convergence than IGLO, especially for heavier atoms. It is shown that the choice of active space is crucial for determination of accurate nuclear shielding constants.......Nuclear shielding calculations are presented for multiconfigurational self-consistent field wave functions using London atomic orbitals (gauge invariant atomic orbitals). Calculations of nuclear shieldings for eight molecules (H2O, H2S, CH4, N2, CO, HF, F2, and SO2) are presented and compared...
Development of Neutron and Photon Shielding Calculation System for Workstation (NPSS-W)
International Nuclear Information System (INIS)
Shimizu, Yoshio; Nojiri, Ichiro; Odajima, Akira; Sasaki, Toshihisa; Kurosawa, Naohiro
1998-01-01
In plant designs and safety evaluations of nuclear fuel cycle facilities, it is important to evaluate the direct radiation and the skyshine (air-scattered photon radiation) from facilities reasonably. The Neutron and Photon Shielding Calculation System for Workstation (NPSS-W) was developed. The NPSS-W can carry out the shielding calculations of the photon and the neutron easily and rapidly. The NPSS-W can easily calculate the radiation source intensity by ORIGEN-S and the dose equivalent rate by SN transport calculational codes, which are ANISN and DOT3.5. The NPSS-W consists of five modules, which named CAL1, CAL2, CAL3, CAL4, CAL5). Some kinds of shielding calculational systems are calculated. The user's manual of NPSS-W, the examples of calculations for each module and the output data are appended. (author)
International Nuclear Information System (INIS)
Johnson, Jeffrey O.; Gallmeier, Franz X.; Popova, Irina
2002-01-01
Determining the bulk shielding requirements for accelerator environments is generally an easy task compared to analyzing the radiation transport through the complex shield configurations and penetrations typically associated with the detailed Title II design efforts of a facility. Shielding calculations for penetrations in the SNS accelerator environment are presented based on hybrid Monte Carlo and discrete ordinates particle transport methods. This methodology relies on coupling tools that map boundary surface leakage information from the Monte Carlo calculations to boundary sources for one-, two-, and three-dimensional discrete ordinates calculations. The paper will briefly introduce the coupling tools for coupling MCNPX to the one-, two-, and three-dimensional discrete ordinates codes in the DOORS code suite. The paper will briefly present typical applications of these tools in the design of complex shield configurations and penetrations in the SNS proton beam transport system
Shield device for controlling the dose of x-rays applied in an x-ray machine
International Nuclear Information System (INIS)
Charrier, P.
1983-01-01
This invention provides an improved shield for use with an x-ray machine. The shield can control the dose of x-rays applied by the machine in different areas without affecting the power of the x-rays. This is achieved with a shield especially designed and positioned to intercept with x-rays for longer or shorter periods in different areas during the taking of the picture, but not for the whole period of time necessary for taking this picture. Each area of the subject being x-rayed is exposed to full power x-rays. However, owing to the shield, the areas that require smaller dose receive these full power x-rays for a shorter portion of the time required to take the picture while the other areas that require larger dose of x-rays, receive the full power x-rays for a longer portion of the full period of time required to take the picture. To ensure this differential exposure, the shield is placed through the path of the x-rays and rotated about an axis which is generally transverse to the direction of travel of the x-rays to cut out some of said x-rays for different portions of the period of time necessary for taking the picture. The shield is preferably shaped to intercept x-rays for a longer period in some areas than in others depending on the required doses. A plurality of differently shaped shields can be provided to suit different picture taking situations
MARS14 deep-penetration calculation for the ISIS target station shielding
International Nuclear Information System (INIS)
Nakao, Noriaki; Nunomiya, Tomoya; Iwase, Hiroshi; Nakamura, Takashi
2004-01-01
The calculation of neutron penetration through a thick shield was performed with a three-dimensional multi-layer technique using the MARS14(02) Monte Carlo code to compare with the experimental shielding data in 1998 at the ISIS spallation neutron source facility of Rutherford Appleton Laboratory. In this calculation, secondary particles from a tantalum target bombarded by 800-MeV protons were transmitted through a bulk shield of approximately 3-m-thick iron and 1-m-thick concrete. To accomplish this deep-penetration calculation, a three-dimensional multi-layer technique and energy cut-off method were used considering a spatial statistical balance. Finally, the energy spectra of neutrons behind the very thick shield could be calculated down to the thermal energy with good statistics, and the calculated results typically agree well within a factor of two with the experimental data over a broad energy range. The 12 C(n,2n) 11 C reaction rates behind the bulk shield were also calculated, which agree with the experimental data typically within 60%. These results are quite impressive in calculation accuracy for deep-penetration problem
Structural shielding of medical X-ray rooms for diagnostic installations
International Nuclear Information System (INIS)
Rabitsch, H.
1979-06-01
In Part I (RIG 8), the various design procedures for shielding against X-rays are discussed and compared. In particular, this comparison is carried out between the shielding obtained conforming to the Austrian Regulations for Radiation Protection and that obtained from the DIN-standard DIN 6812; this latter includes the various operating conditions of diagnostic installations up to 150 kV. Several examples for particular structural shielding components in medical radiation rooms are given. (author)
Effect of low-Z absorber's thickness on gamma-ray shielding parameters
Energy Technology Data Exchange (ETDEWEB)
Mann, Kulwinder Singh, E-mail: ksmann6268@gmail.com [Department of Applied Sciences, Punjab Technical University, Kapurthala 144601 (India); Department of Physics, D.A.V. College, Bathinda 151001, Punjab (India); Heer, Manmohan Singh [Department of Physics, Kanya Maha Vidyalaya, Jalandhar 144001 (India); Rani, Asha [Department of Applied Sciences, Ferozpur College of Engineering and Technology, Ferozshah, Ferozpur 142052 (India)
2015-10-11
Gamma ray shielding behaviour of any material can be studied by various interaction parameters such as total mass attenuation coefficient (μ{sub m}); half value layer (HVL); tenth value layer (TVL); effective atomic number (Z{sub eff}), electron density (N{sub el}), effective atomic weight (A{sub eff}) and buildup factor. For gamma rays, the accurate measurements of μ{sub m} (cm{sup 2} g{sup −1}) theoretically require perfect narrow beam irradiation geometry. However, the practical geometries used for the experimental investigations deviate from perfect-narrowness thereby the multiple scattered photons cause systematic errors in the measured values of μ{sub m}. Present investigation is an attempt to find the optimum value of absorber thickness (low-Z) for which these errors are insignificant and acceptable. Both experimental and theoretical calculations have been performed to investigate the effect of absorber's thickness on μ{sub m} of six low-Z (10
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
Due to the necessity of the use of accelerators with voltages above 10 MV in medical facilities, becomes necessary to evaluate the efficiency of the thickness of shielding materials used in rooms that contain these devices. This work presents the development of an algorithm able to provide data in a practical way, regarding the thickness of materials that can be used for an effective shielding against primary beams from these equipment. The use of the computer language C ++ allowed developing a practical tool for determining the thickness of materials required to protect the public and Individuals Occupationally Exposed (IOEs) against major powers beams. Furthermore, it was considered by calculations Intensity Modulated Radiotherapy Technique (IMRT). The construction of this tool was based to ensure the dose limits established in the CNEN-NN-3.01. The dose limiting is done through the use of loops able to validate the efficiency of thickness determined by the algorithm itself, and ensure if the radiation dose exceeds the limits set by the standard, it will be the inclusion of sufficient Half-Reducer Layers in so that the dose is within the parameters established by the Brazilian National Nuclear Energy Commission (CNEN). The code validation is performed by comparing results obtained in the examples p recalculated in the NCRP Report-151 (Structural Shielding Design and Evaluation for megavoltage X and Gamma-Ray Radiotherapy Facilities) with the results generated by the code. The results are satisfactory and consistent with that report
Fabrication and replacement work of beryllium frame and gamma-ray shield
International Nuclear Information System (INIS)
Watahiki, Shunsuke; Hanawa, Yoshio; Asano, Norikazu; Hiyama, Kazuhisa; Ito, Sachito; Tsuboi, Kazuaki; Fukasaku, Akitomi
2012-03-01
This replacement work was carried out under refurbishment plan of JMTR for beryllium distortion draw to acceptable limit. And gamma-ray shield refurbishment was carried out the view point of prevention maintenance in consideration of operation plan. Fabrication of beryllium frame and gamma-ray shield was spent for two years it was finished in February, 2010. It took five months to replacement work from January 2010. In this report is presented fabrication and replacement work of beryllium frame and gamma-ray shield. (author)
International Nuclear Information System (INIS)
Nasrabadi, M.N.; Mohammadi, A.; Jalali, M.
2009-01-01
In this paper bulk sample prompt gamma neutron activation analysis (BSPGNAA) was applied to aqueous sample analysis using a relative method. For elemental analysis of an unknown bulk sample, gamma self-shielding coefficient was required. Gamma self-shielding coefficient of unknown samples was estimated by an experimental method and also by MCNP code calculation. The proposed methodology can be used for the determination of the elemental concentration of unknown aqueous samples by BSPGNAA where knowledge of the gamma self-shielding within the sample volume is required.
FCXSEC: multigroup cross-section libraries for nuclear fuel cycle shielding calculations
International Nuclear Information System (INIS)
Ford, W.E. III; Webster, C.C.; Diggs, B.R.; Pevey, R.E.; Croff, A.G.
1980-05-01
Starting with the pseudo-composition-independent VITAMIN-C cross-sectin library, composition-dependent fine-(171n-36γ) and broad-group (22n-21γ) self-shielded AMPX master, broad-group microscopic ANISN-formatted, and broad-group macroscopic ANISN-formatted cross-section libraries were generated to be used for nuclear fuel cycle shielding calculations. The specifications for the data and the procedure used to prepare the libraries are described
Shielding calculation of slow extracted beam facility at KEK proton synchrotron
International Nuclear Information System (INIS)
Hirabayashi, Hiromi; Katoh, Kazuaki
1978-01-01
The KEK proton synchrotron has two external beam lines, i.e. a fast extracted beam line for a bubble chamber and a slow extracted beam line for counter experiments. The maximum total intensity of the slow beam is estimated as 5 x 10 12 protons per sec. For beam losses along the line, shielding calculation was made, and on the basis of these results, adequacy of the current shielding construction plans was discussed. (Mori, K.)
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
Simbol-X Mirror Module Thermal Shields: I-Design and X-Ray Transmission
Collura, A.; Barbera, M.; Varisco, S.; Basso, S.; Pareschi, G.; Tagliaferri, G.; Ayers, T.
2009-05-01
The Simbol-X mission is designed to fly in formation flight configuration. As a consequence, the telescope has both ends open to space, and thermal shielding at telescope entrance and exit is required to maintain temperature uniformity throughout the mirrors. Both mesh and meshless solutions are presently under study for the shields. We discuss the design and the X-ray transmission.
Simbol-X Mirror Module Thermal Shields: I - Design and X-Ray Transmission
International Nuclear Information System (INIS)
Collura, A.; Varisco, S.; Barbera, M.; Basso, S.; Pareschi, G.; Tagliaferri, G.; Ayers, T.
2009-01-01
The Simbol-X mission is designed to fly in formation flight configuration. As a consequence, the telescope has both ends open to space, and thermal shielding at telescope entrance and exit is required to maintain temperature uniformity throughout the mirrors. Both mesh and meshless solutions are presently under study for the shields. We discuss the design and the X-ray transmission.
Calculation of the electron trajectory for 200 kV self-shielded electron accelerator
International Nuclear Information System (INIS)
Wang Shuiqing
2000-01-01
In order to calculate the electron trajectory of 200 kV self-shielded electron accelerator, the electric field is calculated with a TRAJ program. In this program, following electron track mash points one by one, the electron beam trajectories are calculated. Knowing the effect of grid voltage on electron optics and gaining grid voltage focusing effect in the various energy grades, the authors have gained scientific basis for adjusting grid voltage, and also accumulated a wealth of experience for designing self-shielded electron accelerator or electron curtain in future
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
Directory of Open Access Journals (Sweden)
M.I. Sayyed
Full Text Available The present investigation has been performed on different bricks for the purpose of gamma-ray shielding. The values of the mass attenuation coefficient (µ/ρ, energy absorption buildup factor (EABF and exposure buildup factor (EBF were determined and utilized to assess the shielding effectiveness of the bricks under investigation. The mass attenuation coefficients of the selected bricks were calculated theoretically using WinXcom program and compared with MCNPX code. Good agreement between WinXcom and MCNPX results was observed. Furthermore, the EABF and EBF have been discussed as functions of the incident photon energy and penetration depth. It has been found that the EABF and EBF values are very large in the intermediate energy region. The steel slag showed good shielding properties, consequently, this brick is eco-friendly and feasible compared with other types of bricks used for construction. The results in this work should be useful in the construction of effectual shielding against hazardous gamma-rays. Keywords: Brick, Mass attenuation coefficient, Buildup factor, G-P fitting, Radiation shielding
International Nuclear Information System (INIS)
Kim, Jae Cheon; Kim, Soon Young; Lee, Hwan Soo; Ha, Pham Nhu Viet; Kim, Jong Kyung
2008-01-01
EASYQAD version β was developed by MATLAB GUI (Graphical User Interface) as a visualization code system based on QAD-CGGP-A point-kernel code for convenient shielding calculations of gammas and neutrons. It consists of four graphic interface modules including GEOMETRY, INPUT, OUTPUT, and SHIELD. These modules were compiled in C++ programming language by using the MATLAB Compiler Toolbox to form a stand-along code system that can be run on the Windows XP operating system without MATLAB installation. In addition, EASYQAD version β has user-friendly graphical interfaces and, additionally, many useful functions in comparison with QAD- CGGP-A such as common material library, line and grid detectors, and multi-group energy calculations so as to increase its applicability in the field of radiation shielding analysis. It is a powerful tool for non-experts to analyze easily the shielding problems without special training. Therefore, EASYOAD version β is expected to contribute effectively to the development of radiation shielding analysis by providing users in medical and industrial fields with an efficient radiation shielding code. (author)
DEMONR, Monte-Carlo Shielding Calculation for Neutron Flux and Neutron Spectra, Teaching Program
International Nuclear Information System (INIS)
Courtney, J. C.
1987-01-01
1 - Description of problem or function: DEMONR treats the behavior of neutrons in a slab shield. It is frequently used as a teaching tool. 2 - Method of solution: An unbiased Monte Carlo code calculates the number, energy, and direction of neutrons that penetrate or are reflected from a shield. 3 - Restrictions on the complexity of the problem: Only one shield may be used in each problem. The shield material may be a single element or a homogeneous mixture of elements with a single effective atomic weight. Only elastic scattering and neutron capture processes are allowed. The source is a point located on one face of the slab. It provides a cosine distribution of current. Monoenergetic or fission spectrum neutrons may be selected
DEFF Research Database (Denmark)
Hinrichsen, Yvonne; Finck, Robert; Östlund, Karl
2018-01-01
building used was a standard prefabricated structure obtained from a commercial manufacturer. Four reference positions for the gamma radiation detectors were used inside the building. Theoretical dose rate calculations were performed using the Monte Carlo code MCNP6, and additional calculations were......Experimentally and theoretically determined shielding factors for a common light construction dwelling type were obtained and compared. Sources of the gamma-emitting radionuclides 60Co and 137Cs were positioned around and on top of a modular building to represent homogeneous fallout. The modular...... performed that compared the shielding factor for 137Cs and 134Cs. This work demonstrated the applicability of using MCNP6 for theoretical calculations of radioactive fallout scenarios. Furthermore, the work showed that the shielding effect for modular buildings is almost the same for 134Cs as for 137Cs....
Simbol-X Mirror Module Thermal Shields: II-Small Angle X-Ray Scattering Measurements
Barbera, M.; Ayers, T.; Collura, A.; Nasillo, G.; Pareschi, G.; Tagliaferri, G.
2009-05-01
The formation flight configuration of the Simbol-X mission implies that the X-ray mirror module will be open to Space on both ends. In order to reduce the power required to maintain the thermal stability and, therefore, the high angular resolution of the shell optics, a thin foil thermal shield will cover the mirror module. Different options are presently being studied for the foil material of these shields. We report results of an experimental investigation conducted to verify that the scattering of X-rays, by interaction with the thin foil material of the thermal shield, will not significantly affect the performances of the telescope.
Simbol-X Mirror Module Thermal Shields: II-Small Angle X-Ray Scattering Measurements
International Nuclear Information System (INIS)
Barbera, M.; Ayers, T.; Collura, A.; Nasillo, G.; Pareschi, G.; Tagliaferri, G.
2009-01-01
The formation flight configuration of the Simbol-X mission implies that the X-ray mirror module will be open to Space on both ends. In order to reduce the power required to maintain the thermal stability and, therefore, the high angular resolution of the shell optics, a thin foil thermal shield will cover the mirror module. Different options are presently being studied for the foil material of these shields. We report results of an experimental investigation conducted to verify that the scattering of X-rays, by interaction with the thin foil material of the thermal shield, will not significantly affect the performances of the telescope.
SUBGR: A Program to Generate Subgroup Data for the Subgroup Resonance Self-Shielding Calculation
Energy Technology Data Exchange (ETDEWEB)
Kim, Kang Seog [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
2016-06-06
The Subgroup Data Generation (SUBGR) program generates subgroup data, including levels and weights from the resonance self-shielded cross section table as a function of background cross section. Depending on the nuclide and the energy range, these subgroup data can be generated by (a) narrow resonance approximation, (b) pointwise flux calculations for homogeneous media; and (c) pointwise flux calculations for heterogeneous lattice cells. The latter two options are performed by the AMPX module IRFFACTOR. These subgroup data are to be used in the Consortium for Advanced Simulation of Light Water Reactors (CASL) neutronic simulator MPACT, for which the primary resonance self-shielding method is the subgroup method.
Shielding calculation techniques used in the design of fuel storage systems
International Nuclear Information System (INIS)
Wang, S.S.; Massey, J.V.
1986-01-01
This paper addresses the shielding design and analysis of a concrete modular spent fuel storage system. Particular attention is given to comparing various computation techniques in determining bulk shielding thickness, and also in dealing with the radiation streaming effect through the air exit penetration openings in the module. Three computer codes QADMOD, ANISN, and DOT-IV were used to solve the same problem. In addition, hand albedo calculation were done to augment the result of the QADMOD calculation to properly deal with the surface scattering
Laboratory tests on neutron shields for gamma-ray detectors in space
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...
Design and shielding calculation for a PET/CT facility
International Nuclear Information System (INIS)
Martin Escuela, J. M.; Palau San Pedro, A.; Lopez Diaz, A.
2013-01-01
Following the AAPM Task Group Report No. 108, the NCRP Report No. 147 recommendations and the Cuban's local regulations for nuclear medicine practice were carried out the safety planning and design of a new PET/CT facility for the Nuclear Medicine Department of 'Hermanos Ameijeiras' Hospital. It should be installed in the top floor of the NM building (3th floor), occupied by offices, classrooms and ancillaries areas, meanwhile in the second floor is working the conventional nuclear medicine department. The radiation doses were evaluated in areas of the second, third and quarter floor taking into account the pet isotope, the workload, the occupancy factors of each place, the use factors of different sources and the dose reduction factors, warranty the accomplish of the Cuban dose restrictions associated to the nuclear medicine practice. In each point of calculation was considered the contribution from each source to the total dose, as well as the contribution of the CT in the adjacent room to the imaging room. For the proper facility design was considered the transmission factors of the existing barriers, and calculated the new ones to be added between each source and the estimation point, keeping in mind the space limitations. The PET/CT design plan meet all the needs, the development of the project is consistent with the mission of the facility and the radiation protection regulations of nuclear medicine. (Author)
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.
FENDL neutronics benchmark: Specifications for the calculational neutronics and shielding benchmark
International Nuclear Information System (INIS)
Sawan, M.E.
1994-12-01
During the IAEA Advisory Group Meeting on ''Improved Evaluations and Integral Data Testing for FENDL'' held in Garching near Munich, Germany in the period 12-16 September 1994, the Working Group II on ''Experimental and Calculational Benchmarks on Fusion Neutronics for ITER'' recommended that a calculational benchmark representative of the ITER design should be developed. This report describes the neutronics and shielding calculational benchmark available for scientists interested in performing analysis for this benchmark. (author)
International Nuclear Information System (INIS)
Makarious, A.S.; Swilem, Y.I.; Awwad, Z.; Bayomy, T.
1993-01-01
Slow neutrons and secondary gamma ray distributions in concrete shields with and without a reflecting layer behind layer behind the concrete shield have been investigated first in case of using a bare reactor beam and then on using a B-4 C filtered beam. The total and capture secondary gamma ray coefficient (B gamma and B gamma C ), the ratio of the reflected thermal neutron (gamma) the ratio of the secondary gamma rays caused by reflected neutrons to those caused transmitted neutrons (Th I gamma/F I gamma) and the effect of inserting a blocking layer (a B-4 C layer) between the concrete shield and the reflector on the suppression of the produced secondary gamma rays have been investigated. It was found that the presence of the reflector layer behind the concrete shield reflects some thermal neutrons back to the concrete shields and so it increases the number of thermal neutrons at the interface between the concrete shield and the reflector. Also the capture secondary gamma rays was increased at the interface between the two medii due to the capture of the reflected thermal neutrons in the concrete shields. It was shown that B-gamma is higher than and that B g amma B gamma C and I gamma T h/ I gamma i f for the different concrete types is higher in case of using the graphite reflector than that in using either water or paraffin reflectors. Putting a blocking layer (B 4 C layer) between the concrete shield and the reflector decreases the produced secondary gamma rays due to the absorption of the reflected thermal neutrons. 17 figs
CREST : a computer program for the calculation of composition dependent self-shielded cross-sections
International Nuclear Information System (INIS)
Kapil, S.K.
1977-01-01
A computer program CREST for the calculation of the composition and temperature dependent self-shielded cross-sections using the shielding factor approach has been described. The code includes the editing and formation of the data library, calculation of the effective shielding factors and cross-sections, a fundamental mode calculation to generate the neutron spectrum for the system which is further used to calculate the effective elastic removal cross-sections. Studies to explore the sensitivity of reactor parameters to changes in group cross-sections can also be carried out by using the facility available in the code to temporarily change the desired constants. The final self-shielded and transport corrected group cross-sections can be dumped on cards or magnetic tape in a suitable form for their direct use in a transport or diffusion theory code for detailed reactor calculations. The program is written in FORTRAN and can be accommodated in a computer with 32 K work memory. The input preparation details, sample problem and the listing of the program are given. (author)
Shielded helix traveling wave cathode ray tube deflection structure
Norris, N.J.; Hudson, C.L.
1992-12-15
Various embodiments of a helical coil deflection structure of a CRT are described and illustrated which provide shielding between adjacent turns of the coil on either three or four sides of each turn in the coil. Threaded members formed with either male or female threads and having the same pitch as the deflection coil are utilized for shielding the deflection coil with each turn of the helical coil placed between adjacent threads which act to shield each coil turn from adjacent turns and to confine the field generated by the coil to prevent or inhibit cross-coupling between adjacent turns of the coil to thereby prevent generation of fast fields which might otherwise deflect the beam out of time synchronization with the electron beam pulse. 13 figs.
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)
SU-E-T-270: Optimized Shielding Calculations for Medical Linear Accelerators (LINACs).
Muhammad, W; Lee, S; Hussain, A
2012-06-01
The purpose of radiation shielding is to reduce the effective equivalent dose from a medical linear accelerator (LINAC) to a point outside the room to a level determined by individual state/international regulations. The study was performed to design LINAC's room for newly planned radiotherapy centers. Optimized shielding calculations were performed for LINACs having maximum photon energy of 20 MV based on NCRP 151. The maximum permissible dose limits were kept 0.04 mSv/week and 0.002 mSv/week for controlled and uncontrolled areas respectively by following ALARA principle. The planned LINAC's room was compared to the already constructed (non-optimized) LINAC's room to evaluate the shielding costs and the other facilities those are directly related to the room design. In the evaluation process it was noted that the non-optimized room size (i.e., 610 × 610 cm 2 or 20 feet × 20 feet) is not suitable for total body irradiation (TBI) although the machine installed inside was having not only the facility of TBI but the license was acquired. By keeping this point in view, the optimized INAC's room size was kept 762 × 762 cm 2. Although, the area of the optimized rooms was greater than the non-planned room (i.e., 762 × 762 cm 2 instead of 610 × 610 cm 2), the shielding cost for the optimized LINAC's rooms was reduced by 15%. When optimized shielding calculations were re-performed for non-optimized shielding room (i.e., keeping room size, occupancy factors, workload etc. same), it was found that the shielding cost may be lower to 41 %. In conclusion, non- optimized LINAC's room can not only put extra financial burden on the hospital but also can cause of some serious issues related to providing health care facilities for patients. © 2012 American Association of Physicists in Medicine.
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.
Study on the use of gonadal protection shields during paediatric pelvic X-rays.
Sikand, Manoj; Stinchcombe, S; Livesley, P J
2003-11-01
There are national guidelines issued by all hospital radiology departments concerning the use of gonadal protection shields for taking X-rays of the pelvis. It is important to follow these guidelines especially when paediatric X-rays are taken. Gonads are very susceptible to radiation as they fall directly in the line of radiation exposure when pelvic X-rays are taken. To examine whether these guidelines were being followed. This audit considered 355 radiographs taken in a 6-month period on 149 patients, under the age of 16 years, attending the orthopaedic department at King's Mill Hospital. In only 23% of the cases studied, the correct use of gonad protection shields had been performed. In 67% of the unprotected patients, the shields were not used at all. In the remainder, the shield was incorrectly applied. Out of all the patients, 45% had more than one X-ray thus exposing the gonads to unnecessary radiation. In addition, 8% of patients had a CT scan, 6% had fluoroscopy and 42% had radiographs of other regions of the body. Guidelines should be adhered to as far as possible and efforts always be made to decrease radiation exposure. Application of the current guidelines excludes the first X-ray exposure of the female pelvis and of the pelvis of trauma patients from the use of shields, thus adding to the number of the X-rays done without protection.
Process of cross section generation for radiation shielding calculations, using the NJOY code
International Nuclear Information System (INIS)
Ono, S.; Corcuera, R.P.
1986-10-01
The process of multigroup cross sections generation for radiation shielding calculations, using the NJOY code, is explained. Photon production cross sections, processed by the GROUPR module, and photon interaction cross sections processed by the GAMINR are given. These data are compared with the data produced by the AMPX system and published data. (author) [pt
Radiation shielding analysis of a special linear accelerator for electron beam and X-ray.
Kang, W G; Pyo, S H; Alkhuraiji, T S; Han, B S; Kang, C M
2017-01-01
The King AbdulAziz City for Science & Technology in the Kingdom of Saudi Arabia plans to build a 10 MeV, 15 kW linear accelerator (LINAC) for electron beam and X-ray. The accelerator will be supplied by EB Tech, Republic of Korea, and the design and construction of the accelerator building will be conducted in the cooperation with EB Tech. This report presents the shielding analysis of the accelerator building using the Monte Carlo N-Particle Transport Code (MCNP). In order to improve the accuracy in estimating deep radiation penetration and to reduce computation time, various variance reduction techniques, including the weight window (WW) method, the deterministic transport (DXTRAN) spheres were considered. Radiation levels were estimated at selected locations in the shielding facility running MCNP6 for particle histories up to 1.0×10+8. The final results indicated that the calculated doses at all selected detector locations met the dose requirement of 50 mSv/yr, which is the United State Nuclear Regulatory Commission (U.S. NRC) requirement.
Comparision of γ -ray shielding properties of some borate glasses
International Nuclear Information System (INIS)
Thind, K.S.
2003-01-01
Several new glasses have been prepared in recent years to suit their increasing number of applications. Some of the glass compositions have distinct properties which make them the most preferred materials for certain applications such as shielding, optical fibers, electronics displays etc. The information of composition, processing and effect of environment on the glass properties is of great importance for their design and application. The shielding ability of pure elements and some mixtures have already been studied but limited attempts have been made on glasses. A good shielding glass should have high absorption cross - section for radiation and at the same time irradiation effects on its mechanical and optical properties should be small. By keeping in view of the importance of shielding ability of borate glasses, we have studied two series of different glass type: x PbO - (1-x) B 2 O 3 and x ZnO - 2xPbO - (1-3x) B 2 O 3 (where x is the mole fraction) by using narrow beam transmission method. A 2' x 2' NaI(Tl) crystal with an energy resolution of 12.5% at 662 keV of 137 Cs was used for the determination of attenuation coefficients and hence interaction cross-sections. Glass samples were prepared by using melt-quenching technique. Thickness measurement was carried out by micrometer and density was measured by Archimede's Principle using benzene as the immersion liquid. The densities of the glasses were found to increase linearly with the increase in the chemical composition of heavy metal oxide. Variations in mass attenuation coefficients and interaction cross ' sections were observed with the change in chemical composition and photon energy. It is found that these glasses have potential applications to be used as radiation shielding materials
Calculation of neutron fluxes in biological shield of the TRIGA Mark II reactor
International Nuclear Information System (INIS)
Bozic, M.; Zagar, T.; Ravnik, M.
2001-01-01
The complete calculation of neutron fluxes in biological shield and verification with experimental results is presented. Calculated results are obtained with TORT code (TORT-Three Dimensional Oak Ridge Discrete Ordinates Neutron/Photon Transport Code). Experimental results used for comparison are available from irradiation experiment with selected type of concrete and other materials in irradiation channel 4 in TRIGA Mark II reactor. These experimental results were used as a benchmark. Homogeneous type of problem (without inserted irradiation channel) and problem with asymmetry (inserted beam port 4, filled with different materials) were of interest for neutron flux calculation. Deviation from material data set up as original parameters is also considered (first of all presence of water in concrete and density of concrete) for type of concrete in biological shield and for selected type of concrete in irradiation channel. BUGLE-96 (47 neutron energy groups) library is used. Excellent agreement between calculated and experimental results for reaction rate is received.(author)
International Nuclear Information System (INIS)
Ling, Tung-Chai; Poon, Chi-Sun; Lam, Wai-Shung; Chan, Tai-Po; Fung, Karl Ka-Lok
2012-01-01
Highlights: ► It is feasible to use recycled CRT glass in mortar as shield against X-ray radiation. ► Shielding properties of CRT mortar is strongly depended on CRT content. ► Linear attenuation coefficient was reduced by 142% upon 100% CRT glass in mortar. ► Effect of mortar thickness and irradiation energies on shielding was investigated. - Abstract: Recycled glass derived from cathode ray tubes (CRT) glass with a specific gravity of approximately 3.0 g/cm 3 can be potentially suitable to be used as fine aggregate for preparing cement mortars for X-ray radiation-shielding applications. In this work, the effects of using crushed glass derived from crushed CRT funnel glass (both acid washed and unwashed) and crushed ordinary beverage container glass at different replacement levels (0%, 25%, 50%, 75% and 100% by volume) of sand on the mechanical properties (strength and density) and radiation-shielding performance of the cement–sand mortars were studied. The results show that all the prepared mortars had compressive strength values greater than 30 MPa which are suitable for most building applications based on ASTM C 270. The density and shielding performance of the mortar prepared with ordinary crushed (lead-free) glass was similar to the control mortar. However, a significant enhancement of radiation-shielding was achieved when the CRT glasses were used due to the presence of lead in the glass. In addition, the radiation shielding contribution of CRT glasses was more pronounced when the mortar was subject to a higher level of X-ray energy.
International Nuclear Information System (INIS)
Sakamoto, Yukio
2001-01-01
The information about neutrons at the surrounding of JCO site in the critical accident is limited to survey results by neutron Rem counter in the period of accident and activation data very near the test facility measured after the shut down of accident. This caused the big uncertainty in the dose estimation by detailed shielding calculation codes. On the other hand, environmental activity data measured by radiochemical researchers included the information about fast neutrons inside of JCO site and thermal neutrons up to 1 km from test facility. It is important to grasp the actual circumstance and examine the executed evaluation of the critical accident as scientifically as possible. Therefore, it is meaningful for different field researchers to corporate and exchange the information. In the Technical Divisions of Radiation Science and Technology in Atomic Energy Society of Japan, the information about neutron spectra are released from their home page and three groups of JAERI/CRC, Sumitomo Atomic Energy Industry and Nuclear Power Engineering Corp. (NUPEC)/Mitsubishi Research Institute Inc. (MRI), tried the shielding calculation by Monte Carlo Code MCNP-4B. The procedures and main results of shielding calculations were reviewed in this report. The main difference of shielding calculation by three groups was density and water content of autoclaved light-weight concrete (ALC) as the wall and ceiling. From the result by NUPEC/MRI, it was estimated that the water content in ALC was from 0.05 g/cm 3 to 0.10 g/cm 3 . The behavior of dose equivalent attenuation obtained by shielding calculation was very similar with the measured data from 250 m to 1,700 m obtained by survey meter, TLD and monitoring post. For more exact dose estimation, more detail examination of density and water content of ALC will be needed. (author)
Attenuation of neutrons and gamma-rays in homogeneous and multilayered shields
International Nuclear Information System (INIS)
Abdo, A.E.; Megahid, R.M.
1997-01-01
Measurements were carried-out to compare the attenuation properties of homogeneous shields and shields of two layers and three layers for fast neutrons and total gamma-rays. These were performed by measuring the fast neutron and total gamma-ray spectra behind homogeneous shields of magnetite-limonite, ilmenite-ilmenite and magnetite-magnetite concretes. The two layers assembly consists of iron and one of the above mentioned concretes, while the three layers shield consists of water, iron and one of the previously mentioned concretes. All measurements were carried-out using a neutron-gamma spectrometer with stilbene scintillator coupled to a fast photo multi player tube. Separation between pulses of recoil protons and recoil electrons was achieved by a pulse shape discrimination technique. 3 tabs., 10 figs., 13 refs
DEFF Research Database (Denmark)
Ilias, Miroslav; Saue, Trond; Enevoldsen, Thomas
2009-01-01
The use of perturbation-dependent London atomic orbitals, also called gauge including atomic orbitals, has proven efficient for calculations of NMR shielding constants and other magnetic properties in the nonrelativistic framework. In this paper, the theory of London atomic orbitals for NMR...... calculates the diamagnetic contribution as an expectation value, leads to significant errors and is not recommended. (C) 2009 American Institute of Physics. [doi:10.1063/1.3240198]...
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
International Nuclear Information System (INIS)
Bakos, G.C.
2001-01-01
This paper deals with the application of Monte Carlo method for the calculation of dose build up factor of, mixed 1.37 and 2.75 MeV, a-rays penetration through stratified shielding slabs. Six double layer shielding slabs namely, 12 A l+Fe, 12 A l+Pb, 6 F e+Al, 6 F e+Pb, 4 P b+Al, 4 P b+Fe were examined. Furthermore, experimental and theoretical results are also presented. The experimental results were taken from the experimental facility installed at the Universities Research reactor Center (Risley, UK). Activated Na2SO3 solution provided a uniform Na-24 disc source of a-rays at both energies (1.37 and 2.75 MeV) with equal intensity. The theoretical results were calculated using the Bowman and Trubey formula. This formula takes into account an exponentially decaying function of the shield thickness (in mfp) to the end point of the multi-layer slab. The experimental and theoretical results were used to evaluate the simulation results produced from a Monte Carlo program (DUTMONCA code) which was developed in Democritus University of Thrace (Xanthi, Greece). The DUTMONCA code was written in Pascal language and run on an Intel PIII-800 microprocessor. The developed code (which is an improved version of an existing Monte Carlo program) has the ability to produce good results for thick shielding slabs overcoming the problems encountered in older version program. The simulation results are compared with experimental and theoretical results. Good agreement can be observed, even for thick layer shielding slabs, although there are some wayward experimental values which are due to sources of error associated with the experimental procedure
International Nuclear Information System (INIS)
Japiassu, Fernando Parois
2013-01-01
When designing radiotherapy treatment rooms, the dimensions of barriers are established on the basis of American calculation methodologies specifically; NCRP Report N° 49, NCRP Report N° 51, and more recently, NCRP Report N° 151. Such barrier calculations are based on parameters reflecting predictions of treatments to be performed within the room; which, in tum, reftect a specific reality found in a country. There exists, however, a variety of modern radiotherapy techniques, such as Intensity Modulated Radiation Therapy (IMRT); Total Body Irradiation (TBl) and radiosurgery (SRS); where patierits are treated in a much different way than during more conventional treatrnents, which are not taken into account the traditional shielding calculation methodology. This may lead to a faulty design of treattnent rooms. In order to establish a comparison between the methodology used to calculate shielding design and the reality of treatments performed in Brazil, two radiotherapy facilitie were selected, both of them offering traditional and modern treatment techniqued as described above. Data in relation with reatments perfotmed over a period of six (6)months of operations in both institutions were collected. Based on tlis informaton, a new set of realistic parameters required for shielding design was estãblished, whicb in turn allowed for a nwe caculation of barrier thickness for both facilities. The barrier thickness resultaing from this calculation was then compared with the barrier thickness propose as part of the original shielding design, approved by the regulatory authority. First, concerning the public facility, the thickness of all primary barriers proposed in the shielding design was actually larger than the thickness resulting from calculations based on realistic parameters. Second, concerning the private facility, the new data show that the thickness of three out of the four primary barriers described in the project is larger than the thickness oresulting from
Energy Technology Data Exchange (ETDEWEB)
Brizuela, Martin; Albornoz, Felipe [INVAP SE, Av. Cmte. Piedrabuena, Bariloche (Argentina)
2012-03-15
A comparison of OPAL shielding calculations against measurements carried out during Commissioning, is presented for relevant structures such as the reactor block, primary shutters, neutron guide bunker, etc. All the results obtained agree very well with the measured values and contribute to establish the confidence on the calculation tools (MCNP4, DORT, etc.) and methodology used for shielding design. (author)
International Nuclear Information System (INIS)
Downar, T.
2009-01-01
The overall objective of the work here has been to eliminate the approximations used in current resonance treatments by developing continuous energy multi-dimensional transport calculations for problem dependent self-shielding calculations. The work here builds on the existing resonance treatment capabilities in the ORNL SCALE code system. The overall objective of the work here has been to eliminate the approximations used in current resonance treatments by developing continuous energy multidimensional transport calculations for problem dependent self-shielding calculations. The work here builds on the existing resonance treatment capabilities in the ORNL SCALE code system. Specifically, the methods here utilize the existing continuous energy SCALE5 module, CENTRM, and the multi-dimensional discrete ordinates solver, NEWT to develop a new code, CENTRM( ) NEWT. The work here addresses specific theoretical limitations in existing CENTRM resonance treatment, as well as investigates advanced numerical and parallel computing algorithms for CENTRM and NEWT in order to reduce the computational burden. The result of the work here will be a new computer code capable of performing problem dependent self-shielding analysis for both existing and proposed GENIV fuel designs. The objective of the work was to have an immediate impact on the safety analysis of existing reactors through improvements in the calculation of fuel temperature effects, as well as on the analysis of more sophisticated GENIV/NGNP systems through improvements in the depletion/transmutation of actinides for Advanced Fuel Cycle Initiatives.
Preliminary shielding calculation for the system of CyberKnife robotic radiosurgery
International Nuclear Information System (INIS)
Toreti, Dalila; Xavier, Clarice; Moura, Fabio
2011-01-01
The CyberKnife robotic system uses a manipulator with six grade of freedom for positioning a 6 MV Linac accelerator for treatment of lesions. This paper presents calculations for a standard room, with 200 cm of thickness walls primary, build for a CyberKnife system, and calculations for a room originally designed for a Linac conventional (with gantry), with secondary barriers of 107 cm thickness. After the realization of shielding for both rooms, the results shown that walls of standard room with 200 cm thickness are adequate for the secondary shield, and for a room with a conventional Linac, from all six evaluated points, two would require additional shielding of nine cm and four cm of concrete with 2.4 g/cubic cm. This shows that the CyberKnife system can be installed in a originally designed room for a conventional Linac with neither restrict nor any shielding, since no incidence of beams on the secondary barriers is existent
Calculation of shielding and radiation doses for PET/CT nuclear medicine facility
International Nuclear Information System (INIS)
Mollah, A.S.; Muraduzzaman, S.M.
2011-01-01
Positron emission tomography (PET) is a new modality that is gaining use in nuclear medicine. The use of PET and computed tomography (CT) has grown dramatically. Because of the high energy of the annihilation radiation (511 keV), shielding requirements are an important consideration in the design of a PET or PET/CT imaging facility. The goal of nuclear medicine and PET facility shielding design is to keep doses to workers and the public as low as reasonably achievable (ALARA). Design involves: 1. Calculation of doses to occupants of the facility and adjacent regions based on projected layouts, protocols and workflows, and 2. Reduction of doses to ALARA through adjustment of the aforementioned parameters. The radiological evaluation of a PET/CT facility consists of the assessment of the annual effective dose both to workers occupationally exposed, and to members of the public. This assessment takes into account the radionuclides involved, the facility features, the working procedures, the expected number of patients per year, and so on. The objective of the study was to evaluate shielding requirements for a PET/CT to be installed in the department of nuclear medicine of Bangladesh Atomic Energy Commission (BAEC). Minimizing shielding would result in a possible reduction of structural as well as financial burden. Formulas and attenuation coefficients following the basic AAPM guidelines were used to calculate un-attenuated radiation through shielding materials. Doses to all points on the floor plan are calculated based primarily on the AAPM guidelines and include consideration of broad beam attenuation and radionuclide energy and decay. The analysis presented is useful for both, facility designers and regulators. (author)
Calculation of the 13C NMR shieldings of the C0 2 complexes of aluminosilicates
Tossell, J. A.
1995-04-01
13C NMR shieldings have been calculated using the random-phase-approximation, localized-orbital local-origins version of ab initio coupled Hartree-Fuck perturbation theory for CO 2 and and for several complexes formed by the reaction of CO 2 with molecular models for aluminosilicate glasses, H 3TOT'H3 3-n, T,T' = Si,Al. Two isomeric forms of the CO 2-aluminosilicate complexes have been considered: (1) "CO 2-like" complexes, in which the CO 2 group is bound through carbon to a bridging oxygen and (2) "CO 3-like" complexes, in which two oxygens of a central CO 3 group form bridging bonds to the two TH 3 groups. The CO 2-like isomer of CO 2-H 3SiOSiH 3 is quite weakly bonded and its 13C isotropic NMR shielding is almost identical to that in free CO 2. As Si is progressively replaced by Al in the - H terminated aluminosilicate model, the CO 2-like isomers show increasing distortion from the free CO 2 geometry and their 13C NMR shieldings decrease uniformly. The calculated 13C shielding value for H 3AlO(CO 2)AlH 3-2 is only about 6 ppm larger than that calculated for point charge stabilized CO 3-2. However, for a geometry of H 3SiO(CO 2) AlH 3-1, in which the bridging oxygen to C bond length has been artificially increased to that found in the - OH terminated cluster (OH) 3SiO(CO 2)Al(OH) 3-1, the calculated 13C shielding is almost identical to that for free CO 2. The CO 3-like isomers of the CO 2-aluminosili-cate complexes show carbonate like geometries and 13C NMR shieldings about 4-9 ppm larger than those of carbonate for all T,T' pairs. For the Si,Si tetrahedral atom pair the CO 2-like isomer is more stable energetically, while for the Si,Al and Al,Al cases the CO 3-like isomer is more stable. Addition of Na + ions to the CO 3-2 or H 3AlO(CO 2)AlH 3-2 complexes reduces the 13C NMR shieldings by about 10 ppm. Complexation with either Na + or CO 2 also reduces the 29Si NMR shieldings of the aluminosilicate models, while the changes in 27Al shielding with Na + or CO 2
Reassessment of shielding calculations for a room housing a Cesium-137 irradiator
International Nuclear Information System (INIS)
Oliveira, Leticia S.; Barbosa, Rugles C.; Rezende, Ana C.B.
2017-01-01
This aim of this work is to reassess the shielding calculations for a room that houses an irradiator with cesium-137 ( 137 Cs) source with activity of 444GBq (12Ci). Shielding or barriers have the function of reducing the intensity of the radiation emitted by a radioactive source, are constituted by materials of high atomic number and guarantee the radiological protection in areas occupied by occupationally exposed individuals or by individuals of the public. The barriers located in the direction of the direct beam of radiation are called primary barriers and are thicker. Already the barriers that attenuate the radiation scattered by the radiated surface are called secondary barriers. In the new calculations, the thickness of the primary barrier was determined by model of the point nucleus model and for the secondary barriers, the differential albedo dose model was used. The results obtained show that all secondary barriers were constructed with overestimated thicknesses and that the radiological protection of individuals from the public and occupationally exposed individuals in the areas outside these barriers is guaranteed. The primary barrier was constructed with a thickness 8% smaller than the thickness obtained in the new calculations. In addition to shielding calculations, classification and signaling of adjacent areas were performed, including necessary emergency procedures. The necessary instrumentation for monitoring these areas was also determined. (author)
Reassessment of shielding calculations for a room housing a Cesium-137 irradiator
Energy Technology Data Exchange (ETDEWEB)
Oliveira, Leticia S.; Barbosa, Rugles C., E-mail: leticia.fmufg@gmail.com, E-mail: rbarbosa@cnen.gov.br [Centro Regional de Ciências Nucleares do Centro Oeste (CRCN-CO/CNEN-GO), Abadia de Goiás, GO (Brazil); Rezende, Ana C.B., E-mail: anacbrz@gmail.com [Universidade Federal de Goiás (UFG), Goiânia, GO (Brazil). Escola de Engenharia
2017-07-01
This aim of this work is to reassess the shielding calculations for a room that houses an irradiator with cesium-137 ({sup 137}Cs) source with activity of 444GBq (12Ci). Shielding or barriers have the function of reducing the intensity of the radiation emitted by a radioactive source, are constituted by materials of high atomic number and guarantee the radiological protection in areas occupied by occupationally exposed individuals or by individuals of the public. The barriers located in the direction of the direct beam of radiation are called primary barriers and are thicker. Already the barriers that attenuate the radiation scattered by the radiated surface are called secondary barriers. In the new calculations, the thickness of the primary barrier was determined by model of the point nucleus model and for the secondary barriers, the differential albedo dose model was used. The results obtained show that all secondary barriers were constructed with overestimated thicknesses and that the radiological protection of individuals from the public and occupationally exposed individuals in the areas outside these barriers is guaranteed. The primary barrier was constructed with a thickness 8% smaller than the thickness obtained in the new calculations. In addition to shielding calculations, classification and signaling of adjacent areas were performed, including necessary emergency procedures. The necessary instrumentation for monitoring these areas was also determined. (author)
Development of a computational code for calculations of shielding in dental facilities
International Nuclear Information System (INIS)
Lava, Deise D.; Borges, Diogo da S.; Affonso, Renato R.W.; Guimaraes, Antonio C.F.; Moreira, Maria de L.
2014-01-01
This paper is prepared in order to address calculations of shielding to minimize the interaction of patients with ionizing radiation and / or personnel. The work includes the use of protection report Radiation in Dental Medicine (NCRP-145 or Radiation Protection in Dentistry), which establishes calculations and standards to be adopted to ensure safety to those who may be exposed to ionizing radiation in dental facilities, according to the dose limits established by CNEN-NN-3.1 standard published in September / 2011. The methodology comprises the use of computer language for processing data provided by that report, and a commercial application used for creating residential projects and decoration. The FORTRAN language was adopted as a method for application to a real case. The result is a programming capable of returning data related to the thickness of material, such as steel, lead, wood, glass, plaster, acrylic, acrylic and leaded glass, which can be used for effective shielding against single or continuous pulse beams. Several variables are used to calculate the thickness of the shield, as: number of films used in the week, film load, use factor, occupational factor, distance between the wall and the source, transmission factor, workload, area definition, beam intensity, intraoral and panoramic exam. Before the application of the methodology is made a validation of results with examples provided by NCRP-145. The calculations redone from the examples provide answers consistent with the report
International Nuclear Information System (INIS)
Sin, M. W.; Kim, M. H.
2002-01-01
To calculate total dose effect on semi-conductor devices in satellite for a period of space mission effectively, two approximate calculation models for a comic radiation shielding were proposed. They are a sectoring method and a chord-length distribution method. When an approximate method was applied in this study, complex structure of satellite was described into multiple 1-dimensional slabs, structural materials were converted to reference material(aluminum), and the pre-calculated dose-depth conversion function was introduced to simplify the calculation process. Verification calculation was performed for orbit location and structure geometry of KITSAT-1 and compared with detailed 3-dimensional calculation results and experimental values. The calculation results from approximate method were estimated conservatively with acceptable error. However, results for satellite mission simulation were underestimated in total dose rate compared with experimental values
Energy Technology Data Exchange (ETDEWEB)
Sin, M. W.; Kim, M. H. [Kyunghee Univ., Yongin (Korea, Republic of)
2002-10-01
To calculate total dose effect on semi-conductor devices in satellite for a period of space mission effectively, two approximate calculation models for a comic radiation shielding were proposed. They are a sectoring method and a chord-length distribution method. When an approximate method was applied in this study, complex structure of satellite was described into multiple 1-dimensional slabs, structural materials were converted to reference material(aluminum), and the pre-calculated dose-depth conversion function was introduced to simplify the calculation process. Verification calculation was performed for orbit location and structure geometry of KITSAT-1 and compared with detailed 3-dimensional calculation results and experimental values. The calculation results from approximate method were estimated conservatively with acceptable error. However, results for satellite mission simulation were underestimated in total dose rate compared with experimental values.
Calculation of neutron shielding for a real loaded C-30 cask by code DORT
International Nuclear Information System (INIS)
Lacina, J.
1999-01-01
Measured neutron dose rates of real loaded C-30 casks for WWER spent fuel assemblies are compared with calculated values in the frame of benchmark calculation task. The part of this benchmark task concerning neutron shielding was calculated. Neutron sources values were taken from data presented by V. Chrapciak during the eighth symposium Atomic Energy Research, Bystrice pod Perstejnem in 1998 and the data about cask from the article of the same author from the Atomic Energy Research working group E meeting at Stolpen in 1998. (Author)
Nomogram for Determining Shield Thickness for Point and Line Sources of Gamma Rays
International Nuclear Information System (INIS)
Joenemalm, C.; Malen, K
1966-10-01
A set of nomograms is given for the determination of the required shield thickness against gamma radiation. The sources handled are point and infinite line sources with shields of Pb, Fe, magnetite concrete (p = 3.6), ordinary concrete (p = 2.3) or water. The gamma energy range covered is 0.5 - 10 MeV. The nomograms are directly applicable for source and dose points on the surfaces of the shield. They can easily be extended to source and dose points in other positions by applying a geometrical correction. Also included are data for calculation of the source strength for the most common materials and for fission product sources
Nomogram for Determining Shield Thickness for Point and Line Sources of Gamma Rays
Energy Technology Data Exchange (ETDEWEB)
Joenemalm, C; Malen, K
1966-10-15
A set of nomograms is given for the determination of the required shield thickness against gamma radiation. The sources handled are point and infinite line sources with shields of Pb, Fe, magnetite concrete (p = 3.6), ordinary concrete (p = 2.3) or water. The gamma energy range covered is 0.5 - 10 MeV. The nomograms are directly applicable for source and dose points on the surfaces of the shield. They can easily be extended to source and dose points in other positions by applying a geometrical correction. Also included are data for calculation of the source strength for the most common materials and for fission product sources.
Ling, Tung-Chai; Poon, Chi-Sun; Lam, Wai-Shung; Chan, Tai-Po; Fung, Karl Ka-Lok
2012-01-15
Recycled glass derived from cathode ray tubes (CRT) glass with a specific gravity of approximately 3.0 g/cm(3) can be potentially suitable to be used as fine aggregate for preparing cement mortars for X-ray radiation-shielding applications. In this work, the effects of using crushed glass derived from crushed CRT funnel glass (both acid washed and unwashed) and crushed ordinary beverage container glass at different replacement levels (0%, 25%, 50%, 75% and 100% by volume) of sand on the mechanical properties (strength and density) and radiation-shielding performance of the cement-sand mortars were studied. The results show that all the prepared mortars had compressive strength values greater than 30 MPa which are suitable for most building applications based on ASTM C 270. The density and shielding performance of the mortar prepared with ordinary crushed (lead-free) glass was similar to the control mortar. However, a significant enhancement of radiation-shielding was achieved when the CRT glasses were used due to the presence of lead in the glass. In addition, the radiation shielding contribution of CRT glasses was more pronounced when the mortar was subject to a higher level of X-ray energy. Copyright © 2011 Elsevier B.V. All rights reserved.
Barium and calcium borate glasses as shielding materials for x rays and gamma rays
DEFF Research Database (Denmark)
Singh, H.; Singh, K.; Sharma, G.
2003-01-01
Values of the gamma-ray, mass attenuation coefficient and the effective atomic number have been determined experimentally for xBaO.(1-x) B2O3 (x=0.24, 0.30, 0.34,0.40 and 0.44) and xCaO. (I-x)B2O3 (x=0.30 and 0.40) glasses at photon energies 356, 511, 662, 1173, and 1332 keV It is pointed out tha...... that these glasses are potential radiation shielding materials. The specific volume of the glasses has been derived from density measurements and studied as a function of composition.......Values of the gamma-ray, mass attenuation coefficient and the effective atomic number have been determined experimentally for xBaO.(1-x) B2O3 (x=0.24, 0.30, 0.34,0.40 and 0.44) and xCaO. (I-x)B2O3 (x=0.30 and 0.40) glasses at photon energies 356, 511, 662, 1173, and 1332 keV It is pointed out...
Shielded radiography with a laser-driven MeV-energy X-ray source
Energy Technology Data Exchange (ETDEWEB)
Chen, Shouyuan; Golovin, Grigory [Department of Physics and Astronomy, University of Nebraska-Lincoln, Lincoln, NE 68588 (United States); Miller, Cameron [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Haden, Daniel; Banerjee, Sudeep; Zhang, Ping; Liu, Cheng; Zhang, Jun; Zhao, Baozhen [Department of Physics and Astronomy, University of Nebraska-Lincoln, Lincoln, NE 68588 (United States); Clarke, Shaun; Pozzi, Sara [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109 (United States); Umstadter, Donald, E-mail: donald.umstadter@unl.edu [Department of Physics and Astronomy, University of Nebraska-Lincoln, Lincoln, NE 68588 (United States)
2016-01-01
We report the results of experimental and numerical-simulation studies of shielded radiography using narrowband MeV-energy X-rays from a compact all-laser-driven inverse-Compton-scattering X-ray light source. This recently developed X-ray light source is based on a laser-wakefield accelerator with ultra-high-field gradient (GeV/cm). We demonstrate experimentally high-quality radiographic imaging (image contrast of 0.4 and signal-to-noise ratio of 2:1) of a target composed of 8-mm thick depleted uranium shielded by 80-mm thick steel, using a 6-MeV X-ray beam with a spread of 45% (FWHM) and 10{sup 7} photons in a single shot. The corresponding dose of the X-ray pulse measured in front of the target is ∼100 nGy/pulse. Simulations performed using the Monte-Carlo code MCNPX accurately reproduce the experimental results. These simulations also demonstrate that the narrow bandwidth of the Compton X-ray source operating at 6 and 9 MeV leads to a reduction of deposited dose as compared to broadband bremsstrahlung sources with the same end-point energy. The X-ray beam’s inherently low-divergence angle (∼mrad) is advantageous and effective for interrogation at standoff distance. These results demonstrate significant benefits of all-laser driven Compton X-rays for shielded radiography.
Quality control, mean glandular dose estimate and room shielding calculation in mammography
International Nuclear Information System (INIS)
Rakotomalala, H.M.
2014-01-01
This study focuses in the importance of Radiation Protection in mammography. A good control of the radiological risk depends on the dose optimization, room shielding calculation and the quality of equipment. The work was carried out in the three private medical centers called A, B, and C. Dosimetry estimates were made on the equipment of the three centers. Values has been compared with the Diagnostic Reference Levels established by the International Atomic Energy Agency (IAEA). Conformity control of the radiological devices has also been done with the Mammographic Quality Control Kit of the INSTN-Madagascar. Verifications of shields of the room containing the mammography equipment were done by theoretical calculations using the method provided by NCRP 147. [fr
Coupled Hartree-Fock calculation of {sup 13} C shielding tensors in acetylene clusters
Energy Technology Data Exchange (ETDEWEB)
Craw, John Simon; Nascimento, Marco Antonio Chaer [Universidade Federal, Rio de Janeiro, RJ (Brazil). Inst. de Quimica
1992-12-31
The coupled Hartree Fock method has been used to calculate ab-initio carbon magnetic shielding tensors for small clusters of acetylene molecules. The chemical shift increases from the monomer to the dimer and trimer. This is mainly due increased diamagnetism, which is imperfectly cancelled by increased paramagnetism due to loss of axial symmetry. Anisotropic effects are shown to be small in both the dimer the and trimer. (author) 21 refs., 2 tabs.
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
International Nuclear Information System (INIS)
Kharrati, Hedi; Agrebi, Amel; Karaoui, Mohamed-Karim
2007-01-01
X-ray buildup factors of lead in broad beam geometry for energies from 15 to 150 keV are determined using the general purpose Monte Carlo N-particle radiation transport computer code (MCNP4C). The obtained buildup factors data are fitted to a modified three parameter Archer et al. model for ease in calculating the broad beam transmission with computer at any tube potentials/filters combinations in diagnostic energies range. An example for their use to compute the broad beam transmission at 70, 100, 120, and 140 kVp is given. The calculated broad beam transmission is compared to data derived from literature, presenting good agreement. Therefore, the combination of the buildup factors data as determined and a mathematical model to generate x-ray spectra provide a computationally based solution to broad beam transmission for lead barriers in shielding x-ray facilities
International Nuclear Information System (INIS)
Borges, Diogo da S.; Lava, Deise D.; Affonso, Renato R.W.; Moreira, Maria de L.; Guimaraes, Antonio C.F.
2014-01-01
This paper addresses the development of a algorithm capable of analyzing the thickness of the secondary shielding due to the production of secondary beams. The production of this beam requires consideration of scattering angle, as well as factors normally used for screening of medical facilities using radiographic techniques. Besides the beam emanated from scattering radiation, is is necessary to evaluate the contribution of leakage radiation, originating from equipment used for the production of the primary beam. A view of the mutual contribution of these radiation to the formation of the secondary beam has shown the need of using shieldings in adjacent walls of the room. The code was validated by comparison with an example case provided by NCRP-151 Report. In this report calculations for determining the secondary barrier for small angles are presented, that deserves greater attention for shielding and statements related to radiotherapy procedures of Modulated intensity. The results are consistent with those provided in the report, which makes the code can be used as a practical tool for the determination of effective shielding beams of megavoltage X-rays
A comparison of shielding calculation methods for multi-slice computed tomography (CT) systems
International Nuclear Information System (INIS)
Cole, J A; Platten, D J
2008-01-01
Currently in the UK, shielding calculations for computed tomography (CT) systems are based on the BIR-IPEM (British Institute of Radiology and Institute of Physics in Engineering in Medicine) working group publication from 2000. Concerns have been raised internationally regarding the accuracy of the dose plots on which this method depends and the effect that new scanner technologies may have. Additionally, more recent shielding methods have been proposed by the NCRP (National Council on Radiation Protection) from the USA. Thermoluminescent detectors (TLDs) were placed in three CT scanner rooms at different positions for several weeks before being processed. Patient workload and dose data (DLP: the dose length product and mAs: the tube current-time product) were collected for this period. Individual dose data were available for more than 95% of patients scanned and the remainder were estimated. The patient workload data were used to calculate expected scatter radiation for each TLD location by both the NCRP and BIR-IPEM methods. The results were then compared to the measured scattered radiation. Calculated scattered air kerma and the minimum required lead shielding were found to be frequently overestimated compared to the measured air kerma, on average almost five times the measured scattered air kerma.
Improved shielding of the gonads of infants during comparative X-ray examination of the hip joints
International Nuclear Information System (INIS)
Petrik, R.; Reumuth, H.J.
1979-01-01
Improved shielding of the gonads of infants during comparative X-ray examination of the hip joints. The presented gonad shielding device for male and female infants aged 3 to 12 months has two small windows for the incident X-ray beam and allows to keep the gonad dose very low. The regions of the skeleton concerned are adequately displayed. The shield can be applied in routine examinations and does not require personnel to hold the patient. (author)
X-ray shielding behaviour of kaolin derived mullite-barites ceramic
Ripin, A.; Mohamed, F.; Choo, T. F.; Yusof, M. R.; Hashim, S.; Ghoshal, S. K.
2018-03-01
Mullite-barite ceramic (MBC) is an emergent material for effective shielding of redundant ionizing radiation exposure. The composition dependent mechanical, thermal, and microstructure properties of MBC that makes MBC a high performing novel radiation shielding candidate remained unexplored. This paper examines the possibility of exploiting Malaysian kaolin (AKIM-35) and barite (BaSO4) derived ceramic (MBC) system for X-ray shielding operation. Using conventional pressing and sintering method six ceramic samples are prepared by mixing AKIM-35 with barite at varying contents (0, 10, 20, 30, 40 and 50 wt%). Synthesized pressed mixtures are calcined at 400 °C for 30 min and then sintered to 1300 °C for 120 min at a heating rate of 10 °C/min. Sintered samples are characterized via X-ray Diffraction (XRD), Field Emission Scanning Electron Microscope (FESEM), lead equivalent (LE), uniformity and dose reduction analyses. XRD pattern of prepared ceramics revealed the presence of monoclinic barium alumino-silicate (BAS) and orthorhombic mullite as major shielding phases together with other minor phase of barite and hexagonal quartz (SiO2) structures. Furthermore, FESEM images of ceramics (between 0 and 30 wt%) displayed the existence of compacted monoclinic plate of BAS and acicular mullite morphology (ceramics at 40 and 50 wt%). Radiation tests displayed the capacity of ceramics (at 0 and 10 wt%) to shield the X-ray radiation emanated at tube potential range of 50-120 kV. The highest radiation attenuation is ascertained at 70 kV where the dose is reduced remarkably between 99.11% and 97.42%. Ceramics at 0 and 10 wt% demonstrated the highest lead (Pb) equivalent thickness (LE) of 0.44 mm and 0.34 mm, respectively. It is established that such MBC may contribute towards the development of shielding material against ionizing radiation in diagnostic radiology (X-ray) dose range.
Monzen, Hajime; Kanno, Ikuo; Fujimoto, Takahiro; Hiraoka, Masahiro
2017-09-01
Tungsten functional paper (TFP) is a novel paper-based radiation-shielding material. We measured the shielding ability of TFP against x-rays and gamma rays. The TFP was supplied in 0.3-mm-thick sheets that contained 80% tungsten powder and 20% cellulose (C 6 H 10 O 5 ) by mass. In dose measurements for x-rays (60, 80, 100, and 120 kVp), we measured doses after through 1, 2, 3, 5, 10, and 12 TFP sheets, as well as 0.3 and 0.5 mm of lead. In lead equivalence measurements, we measured doses after through 2 and 10 TFP sheets for x-rays (100 and 150 kVp), and 0, 7, 10, 20, and 30 TFP sheets for gamma rays from cesium-137 source (662 keV). And then, the lead equivalent thicknesses of TFP were determined by comparison with doses after through standard lead plates (purity >99.9%). Additionally, we evaluated uniformity of the transmitted dose by TFP with a computed radiography image plate for 50 kVp x-rays. A single TFP sheet was found to have a shielding ability of 65%, 53%, 48%, and 46% for x-rays (60, 80, 100, and 120 kVp), respectively. The lead equivalent thicknesses of two TFP sheets were 0.10 ± 0.02, 0.09 ± 0.02 mmPb, and of ten TFP sheets were 0.48 ± 0.02 and 0.51 ± 0.02 mmPb for 100 and 150 kVp x-rays, respectively. The lead equivalent thicknesses of 7, 10, 20, and 30 sheets of TFP for gamma rays from cesium-137 source were estimated as 0.28, 0.43, 0.91, and 1.50 mmPb with an error of ± 0.01 mm. One TFP sheet had nonuniformity, however, seven TFP sheets provided complete shielding for 50 kVp x-rays. TFP has adequate radiation shielding ability for x-rays and gamma rays within the energy range used in diagnostic imaging field. © 2017 The Authors. Journal of Applied Clinical Medical Physics published by Wiley Periodicals, Inc. on behalf of American Association of Physicists in Medicine.
Energy Technology Data Exchange (ETDEWEB)
Polf, J C; Peterson, S; Beddar, S [M D Anderson Cancer Center, Univeristy of Texas, Houston, TX 77030 (United States); McCleskey, M; Roeder, B T; Spiridon, A; Trache, L [Cyclotron Institute, Texas A and M University, College Station, TX 77843 (United States)], E-mail: jcpolf@mdanderson.org
2009-11-21
In this paper, we present results of initial measurements and calculations of prompt gamma ray spectra (produced by proton-nucleus interactions) emitted from tissue equivalent phantoms during irradiations with proton beams. Measurements of prompt gamma ray spectra were made using a high-purity germanium detector shielded either with lead (passive shielding), or a Compton suppression system (active shielding). Calculations of the spectra were performed using a model of both the passive and active shielding experimental setups developed using the Geant4 Monte Carlo toolkit. From the measured spectra it was shown that it is possible to distinguish the characteristic emission lines from the major elemental constituent atoms (C, O, Ca) in the irradiated phantoms during delivery of proton doses similar to those delivered during patient treatment. Also, the Monte Carlo spectra were found to be in very good agreement with the measured spectra providing an initial validation of our model for use in further studies of prompt gamma ray emission during proton therapy. (note)
Shielding system for the detection of radioisotopes gamma-rays emitters
International Nuclear Information System (INIS)
Nascimento Filho, V.F. do
1983-01-01
A shielding system for detection of radioisotopes gamma-rays emitters in samples of big volumes (plants, animals, soils) is presented. The detection between the beaker Marinelli and the glass tube (inside of the scintillator crystal well) is compared. The beaker Marinelli method allows a drastic reduction in the time detection of the sample. (M.A.C.) [pt
Structural Shielding Design and Evaluation for Megavoltage X- and Gamma-Ray Radiotherapy Facilities.
Glasgow, Glenn P
2006-09-01
Structural Shielding Design and Evaluation for Megavoltage X- and Gamma-Ray Radiotherapy Facilities NCRP Report No. 151, 2005, 246 pp. (Hardcover $100). National Council on Radiation Protection and Measurements, 7910 Woodmont Avenue, Suite 400, Bethesda, MD 20814-3095. ISBN-10 0-0929600-87-8; http://www.NCRPonline.org. © 2006 American Association of Physicists in Medicine.
International Nuclear Information System (INIS)
Burson, Z.G.; Profio, A.E.
1975-12-01
Radiation shielding provided by transportation vehicles and structures typical of where people live and work were estimated for cloud and fallout gamma-ray sources resulting from a hypothetical reactor accident. Dose reduction factors are recommended for a variety of situations for realistically assessing the consequences of reactor accidents
Gamma-ray shielding effect of Gd3+ doped lead barium borate glasses
Kummathi, Harshitha; Naveen Kumar, P.; Vedavathi T., C.; Abhiram, J.; Rajaramakrishna, R.
2018-05-01
The glasses of the batch xPbO: 10BaO: (90-x)B2O3: 0.2Gd2O3 (x = 40,45,50 mol %) were prepared by melt-quench technique. The work emphasizes on gamma ray shielding effect on doped lead glasses. The role of Boron is significant as it acts as better neutron attenuator as compared with any other materials, as the thermal neutron cross-sections are high for Gadolinium, 0.2 mol% is chosen as the optimum concentration for this matrix, as higher the concentration may lead to further increase as it produces secondary γ rays due to inelastic neutron scattering. Shielding effects were studied using Sodium Iodide (NaI) - Scintillation Gamma ray spectrometer. It was found that at higher concentration of lead oxide (PbO) in the matrix, higher the attenuation which can be co-related with density. Infra-red (I.R.) spectra reveals that the conversion of Lose triangles to tight tetrahedral structure results in enhancement of shielding properties. The Differential Scanning Calorimeter (D.S.C.) study also reveals that the increase in glass forming range increases the stability which in-turn results in inter-conversion of BO3 to BO4 units such that the density of glass increases with increase in PbO content, resulting in much stable and efficient gamma ray shielding glasses.
Effects of the scattering anisotropy approximation in multigroup radiation shielding calculations
International Nuclear Information System (INIS)
Altiparmarkov, D.
1983-01-01
Expansion of the scattering cross-sections into Legendre series is the usual way of solving the neutron transport problem. Because of the large space gradients of the neutron flux, the effects of that approximations become especially remarkable in the radiation shielding calculations. In this paper, a method taking into account scattering anisotropy is presented. From the point of view of the accuracy and computing speed, the optimal approximation of the scattering anisotropy is established for the basic protective materials on the basis of simple problem calculations (author) [sr
Energy Technology Data Exchange (ETDEWEB)
Barros, Gustavo A.S.J.; Caldas, Linda V.E., E-mail: gustavaobarros@gmail.co, E-mail: lcaldas@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)
2009-07-01
This work performed the shield calculation of the future rooms walls of the five X-ray equipment of the Instrument Calibration Laboratory of the IPEN, Sao Paulo, Brazil, which will be constructed in project of laboratory enlargement. The obtained results by application of a calculation methodology from an international regulation have shown that the largest thickness of shielding (25.7 cm of concrete or 7.1 mm of lead) will be of the wall which will receive the primary beam of the equipment with a 320 kV voltage. The cost/benefit analysis indicated the concrete as the best material option for the shielding
International Nuclear Information System (INIS)
Chen, Y.; Fischer, U.
2005-01-01
Shielding calculations of advanced nuclear facilities such as accelerator based neutron sources or fusion devices of the tokamak type are complicated due to their complex geometries and their large dimensions, including bulk shields of several meters thickness. While the complexity of the geometry in the shielding calculation can be hardly handled by the discrete ordinates method, the deep penetration of radiation through bulk shields is a severe challenge for the Monte Carlo particle transport technique. This work proposes a dedicated computational scheme for coupled Monte Carlo-Discrete Ordinates transport calculations to handle this kind of shielding problems. The Monte Carlo technique is used to simulate the particle generation and transport in the target region with both complex geometry and reaction physics, and the discrete ordinates method is used to treat the deep penetration problem in the bulk shield. The coupling scheme has been implemented in a program system by loosely integrating the Monte Carlo transport code MCNP, the three-dimensional discrete ordinates code TORT and a newly developed coupling interface program for mapping process. Test calculations were performed with comparison to MCNP solutions. Satisfactory agreements were obtained between these two approaches. The program system has been chosen to treat the complicated shielding problem of the accelerator-based IFMIF neutron source. The successful application demonstrates that coupling scheme with the program system is a useful computational tool for the shielding analysis of complex and large nuclear facilities. (authors)
Radiation-shielded double crystal X-ray monochromator for JET
International Nuclear Information System (INIS)
Barnsley, R.; Morsi, H.W.; Rupprecht, G.; Kaellne, E.
1989-01-01
A double crystal X-ray monochromator for absolute wavelength and intensity measurements with very effective shielding of its detector against neutrons and hard X-rays was brought into operation at JET. Fast wavelength scans were taken of impurity line radiation in the wavelength region from about 0.1 nm to 2.3 nm, and monochromatic as well as spectral line scans, for different operational modes of JET. (author) 5 refs., 4 figs
Evaluation of the shielding of dental X-rays units
International Nuclear Information System (INIS)
Medrano, E.; Vega C, H. R.; Letechipia de L, C.; Hernandez D, V. M.; Salas L, M. A.
2014-08-01
The capacity of the walls of the dental radio-diagnostic rooms has been determined, to diminish the dose levels during the use of the X-rays equipment s. The study was carried out in the Dentistry Academic Unit of the campus Siglo X XI of the Universidad Autonoma de Zacatecas. The X-rays equipment s are a learning tool for the dentistry students and they are also used for offering health services to the population; for this reason is important to verify that the dose levels outside of the room walls are safe. During the evaluation process were used conservative approaches without prejudice of the thickness necessary in benefit of the radiological protection. Of the evaluation was found that all the walls satisfy their function thoroughly like barriers against the X-rays. (Author)
International Nuclear Information System (INIS)
Wagner, J.C.; Haghighat, A.
1998-01-01
Although the Monte Carlo method is considered to be the most accurate method available for solving radiation transport problems, its applicability is limited by its computational expense. Thus, biasing techniques, which require intuition, guesswork, and iterations involving manual adjustments, are employed to make reactor shielding calculations feasible. To overcome this difficulty, the authors have developed a method for using the S N adjoint function for automated variance reduction of Monte Carlo calculations through source biasing and consistent transport biasing with the weight window technique. They describe the implementation of this method into the standard production Monte Carlo code MCNP and its application to a realistic calculation, namely, the reactor cavity dosimetry calculation. The computational effectiveness of the method, as demonstrated through the increase in calculational efficiency, is demonstrated and quantified. Important issues associated with this method and its efficient use are addressed and analyzed. Additional benefits in terms of the reduction in time and effort required of the user are difficult to quantify but are possibly as important as the computational efficiency. In general, the automated variance reduction method presented is capable of increases in computational performance on the order of thousands, while at the same time significantly reducing the current requirements for user experience, time, and effort. Therefore, this method can substantially increase the applicability and reliability of Monte Carlo for large, real-world shielding applications
Calculations for Extra Well Shielding for 15 MV Clinical Linear accelerator
International Nuclear Information System (INIS)
Mahmoud, M.A.; Emran, M.M.; Ahmad, A.S.
2000-01-01
A radiological survey was conducted around the walls of a clinical linear accelerator (Siemens Mevatron) in South Egypt Cancer Institute, Assiut University. Neutron measurements showed adequate results for all beam orientations. Photon measurements showed adequate results for all beam orientations except for beam orientation 270 degree, facing the control room. During operation, photon measurements were taken in order to calculate the additional shield thickness required to reduce measurements to accepted values. For convenience, lead was the material of choice for extra shielding. A value for the build up factor needed in the calculations of broad beam attenuation was estimated. Measurements inside the control room after adding the calculated lead thickness are much lower than the annual effective equivalent dose limits recommended by the ICRP-60 (International Commission on Radiation Protection) for occupational exposure. Also, measurements taken in the patients waiting hall recorded levels consistent with the six-hour daily occupancy for members of the public. The value of the build up factor was verified by calculations. Also the variation of build up factor distance from the field centre was calculated. Important and useful recommendations were reached from this experience which should be discussed to avoid facing similar situations in radiotherapy departments in Egypt
International Nuclear Information System (INIS)
Maerker, R.E.; Muckenthaler, F.J.
1976-01-01
Measurements, using a 4-in. Bonner Ball, have been made of the neutron fluxes penetrating a simulation of CRBR upper axial biological shielding at the Tower Shielding Facility. The simulation consisted of a 45.7 cm thick slab of SS-304 followed by a series of sodium tanks having a total thickness of 457 cm followed by slabs of carbon steel up to 61.0 cm thick. Measurements were made behind the stainless steel, behind intermediate thicknesses of 152 cm, 305 cm, and 457 cm of sodium (with the stainless steel in place), and behind various thicknesses of the carbon steel following both 305 cm and 457 cm of sodium (also with the stainless steel in place). Calculated and measured data are presented and compared
International Nuclear Information System (INIS)
Fuente P, A. de la; Dumenigo G, C.; Quevedo G, J.R.; Lopez F, Y.
2006-01-01
The evaluation of applications of construction licenses for the new services of radiotherapy has occupied a significant space in the activity developed by the National Center of Nuclear Safety (CNSN) in the last 2 years. Presently work the experiences of the authors in the evaluation of the required shield for the local where cobalt therapy equipment and lineal accelerators of medical use are used its are exposed, the practical problems detected are approached during the application of the methodologies recommended in both cases and its are discussed which have been the suppositions of items accepted by the Regulatory Authority for the realization of these shield calculations. The accumulated experience allows to assure that the realistic application of the item data and the rational use of the engineering logic makes possible to design local for radiotherapy equipment that fulfill the established dose restrictions in the in use legislation in Cuba, without it implies an excessive expense of construction materials. (Author)
Calculation of shielding thickness by combining the LTSN and Decomposition methods
International Nuclear Information System (INIS)
Borges, Volnei; Vilhena, Marco T. de
1997-01-01
A combination of the LTS N and Decomposition methods is reported to shielding thickness calculation. The angular flux is evaluated solving a transport problem in planar geometry considering the S N approximation, anisotropic scattering and one-group of energy. The Laplace transform is applied in the set of S N equations. The transformed angular flux is then obtained solving a transcendental equation and the angular flux is restored by the Heaviside expansion technique. The scalar flux is attained integrating the angular flux by Gaussian quadrature scheme. On the other hand, the scalar flux is linearly related to the dose rate through the mass and energy absorption coefficient. The shielding thickness is obtained solving a transcendental equation resulting from the application of the LTS N approach by the Decomposition methods. Numerical simulations are reported. (author). 6 refs., 3 tabs
Radiation Resistance Test of Wireless Sensor Node and the Radiation Shielding Calculation
Energy Technology Data Exchange (ETDEWEB)
Li, Liqan; Sur, Bhaskar [Atomic Energy of Canada Limited, Ontario (Canada); Wang, Quan [University of Western Ontario, Ontario (Canada); Deng, Changjian [The University of Electronic Science and Technology, Chengdu (China); Chen, Dongyi; Jiang, Jin [Applied Physics Branch, Ontario (Korea, Republic of)
2014-08-15
A wireless sensor network (WSN) is being developed for nuclear power plants. Amongst others, ionizing radiation resistance is one essential requirement for WSN to be successful. This paper documents the work done in Chalk River Laboratories of Atomic Energy of Canada Limited (AECL) to test the resistance to neutron and gamma radiation of some WSN nodes. The recorded dose limit that the nodes can withstand before being damaged by the radiation is compared with the radiation environment inside a typical CANDU (CANada Deuterium Uranium) power plant reactor building. Shielding effects of polyethylene, cadmium and lead to neutron and gamma radiations are also analyzed using MCNP simulation. The shielding calculation can be a reference for the node case design when high dose rate or accidental condition (like Fukushima) is to be considered.
International Nuclear Information System (INIS)
Matijevic, M.; Grgic, D.; Jecmenica, R.
2016-01-01
This paper presents comparison of the Krsko Power Plant simplified Spent Fuel Pool (SFP) dose rates using different computational shielding methodologies. The analysis was performed to estimate limiting gamma dose rates on wall mounted level instrumentation in case of significant loss of cooling water. The SFP was represented with simple homogenized cylinders (point kernel and Monte Carlo (MC)) or cuboids (MC) using uranium, iron, water, and dry-air as bulk region materials. The pool is divided on the old and new section where the old one has three additional subsections representing fuel assemblies (FAs) with different burnup/cooling time (60 days, 1 year and 5 years). The new section represents the FAs with the cooling time of 10 years. The time dependent fuel assembly isotopic composition was calculated using ORIGEN2 code applied to the depletion of one of the fuel assemblies present in the pool (AC-29). The source used in Microshield calculation is based on imported isotopic activities. The time dependent photon spectra with total source intensity from Microshield multigroup point kernel calculations was then prepared for two hybrid deterministic-stochastic sequences. One is based on SCALE/MAVRIC (Monaco and Denovo) methodology and another uses Monte Carlo code MCNP6.1.1b and ADVANTG3.0.1. code. Even though this model is a fairly simple one, the layers of shielding materials are thick enough to pose a significant shielding problem for MC method without the use of effective variance reduction (VR) technique. For that purpose the ADVANTG code was used to generate VR parameters (SB cards in SDEF and WWINP file) for MCNP fixed-source calculation using continuous energy transport. ADVATNG employs a deterministic forward-adjoint transport solver Denovo which implements CADIS/FW-CADIS methodology. Denovo implements a structured, Cartesian-grid SN solver based on the Koch-Baker-Alcouffe parallel transport sweep algorithm across x-y domain blocks. This was first
Shielding calculations for industrial 5/7.5MeV electron accelerators using the MCNP Monte Carlo Code
Peri, Eyal; Orion, Itzhak
2017-09-01
High energy X-rays from accelerators are used to irradiate food ingredients to prevent growth and development of unwanted biological organisms in food, and by that extend the shelf life of the products. The production of X-rays is done by accelerating 5 MeV electrons and bombarding them into a heavy target (high Z). Since 2004, the FDA has approved using 7.5 MeV energy, providing higher production rates with lower treatments costs. In this study we calculated all the essential data needed for a straightforward concrete shielding design of typical food accelerator rooms. The following evaluation is done using the MCNP Monte Carlo code system: (1) Angular dependence (0-180°) of photon dose rate for 5 MeV and 7.5 MeV electron beams bombarding iron, aluminum, gold, tantalum, and tungsten targets. (2) Angular dependence (0-180°) spectral distribution simulations of bremsstrahlung for gold, tantalum, and tungsten bombarded by 5 MeV and 7.5 MeV electron beams. (3) Concrete attenuation calculations in several photon emission angles for the 5 MeV and 7.5 MeV electron beams bombarding a tantalum target. Based on the simulation, we calculated the expected increase in dose rate for facilities intending to increase the energy from 5 MeV to 7.5 MeV, and the concrete width needed to be added in order to keep the existing dose rate unchanged.
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.
International Nuclear Information System (INIS)
Stoker, C.C.; Ball, G.
2000-01-01
The ever increasing expansion of the irradiation product portfolio of the SAFARI-1 reactor leads to the need to routinely calculate the radio-isotope concentrations and source terms for the materials irradiated in the reactor accurately. In addition to this, the required shielding for the transportation and processing of these irradiation products needs to be determined. In this paper the calculational methodology applied is described with special attention given to the spectrum dependence of the one-group cross sections of selected SAFARI-1 irradiation materials and the consequent effect on the determination of the isotope concentrations and source terms. Comparisons of the calculated isotopic concentrations and dose rates with experimental analysis and measurements provide confidence in the calculational methodologies and data used. (author)
Study of x-ray medical mitigation with lead and aluminium shield
International Nuclear Information System (INIS)
Malheiros, Emiliane A.; Ramos, Roberto Paulo B.; Oliveira, Ezequias Fernandes
2016-01-01
In this work, lead and aluminum as shielding materials and their variations in the spectra emitted by the X-ray equipment through the use of a computer program that determines the photon fluence . The study of the primary beam for power spectra used in the practice of diagnostic radiology allows you to analyze data representative of the average transmission and fluency for the studied materials. So we seek to analyze the transmission curves of lead and aluminum, as well as its influence on the thickness of shielding and changing the radiation spectrum characteristics X in the transmission of photons. (author)
Komorovsky, Stanislav; Repisky, Michal; Ruud, Kenneth; Malkina, Olga L; Malkin, Vladimir G
2013-12-27
A four-component relativistic method for the calculation of NMR shielding constants of paramagnetic doublet systems has been developed and implemented in the ReSpect program package. The method uses a Kramer unrestricted noncollinear formulation of density functional theory (DFT), providing the best DFT framework for property calculations of open-shell species. The evaluation of paramagnetic nuclear magnetic resonance (pNMR) tensors reduces to the calculation of electronic g tensors, hyperfine coupling tensors, and NMR shielding tensors. For all properties, modern four-component formulations were adopted. The use of both restricted kinetically and magnetically balanced basis sets along with gauge-including atomic orbitals ensures rapid basis-set convergence. These approaches are exact in the framework of the Dirac-Coulomb Hamiltonian, thus providing useful reference data for more approximate methods. Benchmark calculations on Ru(III) complexes demonstrate good performance of the method in reproducing experimental data and also its applicability to chemically relevant medium-sized systems. Decomposition of the temperature-dependent part of the pNMR tensor into the traditional contact and pseudocontact terms is proposed.
Solar Proton Transport Within an ICRU Sphere Surrounded by a Complex Shield: Ray-trace Geometry
Slaba, Tony C.; Wilson, John W.; Badavi, Francis F.; Reddell, Brandon D.; Bahadori, Amir A.
2015-01-01
A computationally efficient 3DHZETRN code with enhanced neutron and light ion (Z is less than or equal to 2) propagation was recently developed for complex, inhomogeneous shield geometry described by combinatorial objects. Comparisons were made between 3DHZETRN results and Monte Carlo (MC) simulations at locations within the combinatorial geometry, and it was shown that 3DHZETRN agrees with the MC codes to the extent they agree with each other. In the present report, the 3DHZETRN code is extended to enable analysis in ray-trace geometry. This latest extension enables the code to be used within current engineering design practices utilizing fully detailed vehicle and habitat geometries. Through convergence testing, it is shown that fidelity in an actual shield geometry can be maintained in the discrete ray-trace description by systematically increasing the number of discrete rays used. It is also shown that this fidelity is carried into transport procedures and resulting exposure quantities without sacrificing computational efficiency.
International Nuclear Information System (INIS)
Park, Hong Sik; Kim, Min; Park, Seong Chan; Seo, Jong Tae; Kim, Eun Kee
2005-01-01
The SHIELD code has been used to calculate the source terms of NSSS Auxiliary System (comprising CVCS, SIS, and SCS) components of the OPR1000. Because the code had been developed based upon the SYSTEM80 design and the APR1400 NSSS Auxiliary System design is considerably changed from that of SYSTEM80 or OPR1000, the SHIELD code cannot be used directly for APR1400 radiation design. Thus the hand-calculation is needed for the portion of design changes using the results of the SHIELD code calculation. In this study, the SHIELD code is modified to incorporate the APR1400 design changes and the source term calculation is performed for the APR1400 NSSS Auxiliary System components
Neutron Buildup Factors Calculation for Support Vector Regression Application in Shielding Analysis
International Nuclear Information System (INIS)
Duckic, P.; Matijevic, M.; Grgic, D.
2016-01-01
In this paper initial set of data for neutron buildup factors determination using Support Vector Regression (SVR) method is prepared. The performance of SVR technique strongly depends on the quality of information used for model training. Thus it is very important to provide representable data to the SVR. SVR is a supervised type of learning so it demands data in the input/output form. In the case of neutron buildup factors estimation, the input parameters are the incident neutron energy, shielding thickness and shielding material and the output parameter is the neutron buildup factor value. So far the initial sets of data for different shielding configurations have been obtained using SCALE4.4 sequence SAS3. However, this results were obtained using group constants, thus the incident neutron energy was determined as the average value for each energy group. Obtained this way, the data provided to the SVR are fewer and therefore insufficient. More valuable information is obtained using SCALE6.2beta5 sequence MAVRIC which can perform calculations for the explicit incident neutron energy, which leads to greater maneuvering possibilities when active learning measures are employed, and consequently improves the quality of the developed SVR model.(author).
AUTOSECOL: an automatic calculation of the self-shielding of heavy isotope resonances
International Nuclear Information System (INIS)
Grandotto-Biettoli, Marc.
The formalism is based on separating both types of resonance effects: local energy effects creating a fine structure in the flux, and bulk effects resulting in a slow variation in the flux. Effective reaction rates are defined that, used as tables in a multigroup calculation of cells with a large pitch in regard to resonance widths, allow an exact account of the dependence of the effective integral upon fast variations in the flux. These tables are used to introduce this phenomenon of resonance self-shielding in the multigroup Apollo program for solving the neutron transport equation, they are derived from nuclear data with using some parameters relating to the physical state of the resonant isotope inside the fuel medium. The AUTOSECOL system provides a library of effective reaction rates for taking account of the resonance self-shielding effect on the neutron flux in nuclear reactor cells. Its versatility in regard to the methods previously used for solving the same problem allows a rapid testing of the consequences of considering the self-shielding effect of new isotope resonances, a following up of the evolution in nuclear data evaluation, and rapidly studying the interest lying in new data. Results obtained with AUTOSECOL are compared with those obtained when using the SECOL code for computing the effective reaction rates of 235 U, 239 Pu, 107 Ag, 109 Ag, and 241 Pu [fr
Efficient heterogeneous execution of Monte Carlo shielding calculations on a Beowulf cluster
International Nuclear Information System (INIS)
Dewar, D.; Hulse, P.; Cooper, A.; Smith, N.
2005-01-01
Recent work has been done in using a high-performance 'Beowulf' cluster computer system for the efficient distribution of Monte Carlo shielding calculations. This has enabled the rapid solution of complex shielding problems at low cost and with greater modularity and scalability than traditional platforms. The work has shown that a simple approach to distributing the workload is as efficient as using more traditional techniques such as PVM (Parallel Virtual Machine). In addition, when used in an operational setting this technique is fairer with the use of resources than traditional methods, in that it does not tie up a single computing resource but instead shares the capacity with other tasks. These developments in computing technology have enabled shielding problems to be solved that would have taken an unacceptably long time to run on traditional platforms. This paper discusses the BNFL Beowulf cluster and a number of tests that have recently been run to demonstrate the efficiency of the asynchronous technique in running the MCBEND program. The BNFL Beowulf currently consists of 84 standard PCs running RedHat Linux. Current performance of the machine has been estimated to be between 40 and 100 Gflop s -1 . When the whole system is employed on one problem up to four million particles can be tracked per second. There are plans to review its size in line with future business needs. (authors)
A new technique for radiation shielding in superficial X-ray therapy
International Nuclear Information System (INIS)
Baily, B.; Coe, M.A.; Hearnden, T.M.
1981-01-01
A new technique is described for making face shields for superficial X-ray therapy of small lesions located near sharp face contour changes. A ''Darvic'' shell is first vacuum-formed on a plaster cast. A low-melting-point alloy is then finely sprayed on to the shell. The area to be treated is cut out of the alloy mask and Darvic shell. Four alloys of varying melting point and varying lead and/or tin content were tested: MCP 69, MCP 70 and MCP 124 alloys afforded better shielding over the whole energy range than MCP 137. This technique of spraying masks produces a perfectly fitting shield, it is quick to make and is cheap because the mask can be recycled once treatment is finished. (U.K.)
DEFF Research Database (Denmark)
Steinmann, Casper; Olsen, Jógvan Magnus Haugaard; Kongsted, Jacob
2014-01-01
We present NMR shielding constants obtained through quantum mechanical/molecular mechanical (QM/MM) embedding calculations. Contrary to previous reports, we show that a relatively small QM region is sufficient, provided that a high-quality embedding potential is used. The calculated averaged NMR...... shielding constants of both acrolein and acetone solvated in water are based on a number of snapshots extracted from classical molecular dynamics simulations. We focus on the carbonyl chromophore in both molecules, which shows large solvation effects, and we study the convergence of shielding constants...
International Nuclear Information System (INIS)
Serikov, A.; Fischer, U.; Grosse, D.; Leichtle, D.; Majerle, M.
2011-01-01
The Monte Carlo (MC) method is the most suitable computational technique of radiation transport for shielding applications in fusion neutronics. This paper is intended for sharing the results of long term experience of the fusion neutronics group at Karlsruhe Institute of Technology (KIT) in radiation shielding calculations with the MCNP5 code for the ITER fusion reactor with emphasizing on the use of several ITER project-driven computer programs developed at KIT. Two of them, McCad and R2S, seem to be the most useful in radiation shielding analyses. The McCad computer graphical tool allows to perform automatic conversion of the MCNP models from the underlying CAD (CATIA) data files, while the R2S activation interface couples the MCNP radiation transport with the FISPACT activation allowing to estimate nuclear responses such as dose rate and nuclear heating after the ITER reactor shutdown. The cell-based R2S scheme was applied in shutdown photon dose analysis for the designing of the In-Vessel Viewing System (IVVS) and the Glow Discharge Cleaning (GDC) unit in ITER. Newly developed at KIT mesh-based R2S feature was successfully tested on the shutdown dose rate calculations for the upper port in the Neutral Beam (NB) cell of ITER. The merits of McCad graphical program were broadly acknowledged by the neutronic analysts and its continuous improvement at KIT has introduced its stable and more convenient run with its Graphical User Interface. Detailed 3D ITER neutronic modeling with the MCNP Monte Carlo method requires a lot of computation resources, inevitably leading to parallel calculations on clusters. Performance assessments of the MCNP5 parallel runs on the JUROPA/HPC-FF supercomputer cluster permitted to find the optimal number of processors for ITER-type runs. (author)
Calculations of the photon dose behind concrete shielding of high energy proton accelerators
International Nuclear Information System (INIS)
Dworak, D.; Tesch, K.; Zazula, J.M.
1992-02-01
The photon dose per primary beam proton behind lateral concrete shieldings was calculated by using an extension of the Monte Carlo particle shower code FLUKA. The following photon-producing processes were taken into account: capture of thermal neutrons, deexcitation of nuclei after nuclear evaporation, inelastic neutron scattering and nuclear reactions below 140 MeV, as well as photons from electromagnetic cascades. The obtained ratio of the photon dose to the neutron dose equivalent varies from 8% to 20% and it well compares with measurements performed recently at DESY giving a mean ratio of 14%. (orig.)
International Nuclear Information System (INIS)
Meireles, Ramiro Conceicao
2016-01-01
The shielding calculation methodology for radiotherapy services adopted in Brazil and in several countries is that described in publication 151 of the National Council on Radiation Protection and Measurements (NCRP 151). This methodology however, markedly employs several approaches that can impact both in the construction cost and in the radiological safety of the facility. Although this methodology is currently well established by the high level of use, some parameters employed in the calculation methodology did not undergo to a detailed assessment to evaluate the impact of the various approaches considered. In this work the MCNP5 Monte Carlo code was used with the purpose of evaluating the above mentioned approaches. TVLs values were obtained for photons in conventional concrete (2.35g / cm 3 ), considering the energies of 6, 10 and 25 MeV, respectively, first considering an isotropic radiation source impinging perpendicular to the barriers, and subsequently a lead head shielding emitting a shaped beam, in the format of a pyramid trunk. Primary barriers safety margins, taking in account the head shielding emitting photon beam pyramid-shaped in the energies of 6, 10, 15 and 18 MeV were assessed. A study was conducted considering the attenuation provided by the patient's body in the energies of 6,10, 15 and 18 MeV, leading to new attenuation factors. Experimental measurements were performed in a real radiotherapy room, in order to map the leakage radiation emitted by the accelerator head shielding and the results obtained were employed in the Monte Carlo simulation, as well as to validate the entire study. The study results indicate that the TVLs values provided by (NCRP, 2005) show discrepancies in comparison with the values obtained by simulation and that there may be some barriers that are calculated with insufficient thickness. Furthermore, the simulation results show that the additional safety margins considered when calculating the width of the primary
International Nuclear Information System (INIS)
Kim, Yong IL
2014-01-01
Activation calculation requires flux information at desired location and reaction cross sections for the constituent elements to obtain production rate of activation products. Generally it is not an easy task to obtain fluxes or reaction rates with low uncertainties in a reasonable time for deep penetration problems by using standard Monte Carlo methods. The MAVRIC (Monaco with Automated Variance Reduction using Importance Calculations) sequence in SCALE 6.1 code package is intended to perform radiation transport on problems that are too challenging for standard, unbiased Monte Carlo methods. And the SCALE code system provides plenty of ENDF reaction types enough to consider almost all activation reactions in the nuclear reactor materials. To evaluate the activation of the important isotopes in primary shield, SCALE 6.1 MAVRIC sequence has been utilized for the KSNP reactor model and the calculated results are compared to the isotopic activity concentration of related standard. Related to the planning for decommission, the activation products in concrete primary shield such as Fe-55, Co-60, Ba-133, Eu-152, and Eu-154 are identified as important elements according to the comparisons with related standard for exemption. In this study, reference data are used for the concrete compositions in the activation calculation to see the applicability of MAVRIC code to the evaluation of activation inventory in the concrete primary shield. The composition data of trace elements as shown in Table 1 are obtained from various US power plant sites and accordingly they have large variations in quantity due to the characteristics of concrete composition. In practical estimation of activation radioactivity for a specific plant related to decommissioning, rigorous chemical analysis of concrete samples of the plant would first have to be performed to get exact information for compositions of concrete. Considering the capability of solving deep penetration transport problems and richness
Böhm, Karl-Heinz; Banert, Klaus; Auer, Alexander A
2014-04-23
We present ab-initio calculations of secondary isotope effects on NMR chemical shieldings. The change of the NMR chemical shift of a certain nucleus that is observed if another nucleus is replaced by a different isotope can be calculated by computing vibrational corrections on the NMR parameters using electronic structure methods. We demonstrate that the accuracy of the computational results is sufficient to even distinguish different conformers. For this purpose, benchmark calculations for fluoro(2-2H)ethane in gauche and antiperiplanar conformation are carried out at the HF, MP2 and CCSD(T) level of theory using basis sets ranging from double- to quadruple-zeta quality. The methodology is applied to the secondary isotope shifts for 2-fluoronorbornane in order to resolve an ambiguity in the literature on the assignment of endo- and exo-2-fluoronorbornanes with deuterium substituents in endo-3 and exo-3 positions, also yielding insight into mechanistic details of the corresponding synthesis.
Directory of Open Access Journals (Sweden)
Karl-Heinz Böhm
2014-04-01
Full Text Available We present ab-initio calculations of secondary isotope effects on NMR chemical shieldings. The change of the NMR chemical shift of a certain nucleus that is observed if another nucleus is replaced by a different isotope can be calculated by computing vibrational corrections on the NMR parameters using electronic structure methods. We demonstrate that the accuracy of the computational results is sufficient to even distinguish different conformers. For this purpose, benchmark calculations for fluoro(2-2Hethane in gauche and antiperiplanar conformation are carried out at the HF, MP2 and CCSD(T level of theory using basis sets ranging from double- to quadruple-zeta quality. The methodology is applied to the secondary isotope shifts for 2-fluoronorbornane in order to resolve an ambiguity in the literature on the assignment of endo- and exo-2-fluoronorbornanes with deuterium substituents in endo-3 and exo-3 positions, also yielding insight into mechanistic details of the corresponding synthesis.
International Nuclear Information System (INIS)
Elmorabit, A.; Labrim, H.
2010-01-01
Full text: this work is part of developing an analytical method for solving the neutrons transport equation in improving the treatment of the anisotropy of neutron scattering through heterogeneous shielding. We also develop the tools necessary for the formation of multigroup libraries (cross section) with the best choice of the weighting function. Among the radioprotection problems of radioisotopes production experiments in the research reactor core is mainly the photons gamma generation produced by radiative capture: activation of samples and their capsules. So, in order to review the safety of operating personnel and the public is essential to quantify the neutrons flux and gamma photons produced. In this study a numerical methods is used in two different Fortran program to solve the neutron transport problem and to determine the neutron and photon flux. This program based on the Monte Carlo method: the neutron is born with a unit statistical weight, this corrected after each imposed scattering event during its whole history within the shield. The final neutron statistical weight is used in an appropriate estimator to determine the searched response. The generated gamma rays by neutron capture are calculated of different isotopes, and then the equivalent dose rate is evaluated in biological tissue for different neutron source energies. We have identified and studied the choice of the best weighting function to calculate a library of multigroup cross sections self protected by using the energy weighting function. A Fortran program is used as a mathematical tool to solve the neutron slowing down equation in infinite homogeneous medium for different dilutions. We determined the energetic flux distribution and the effective integrals. The results of both calculations are in a good agreement; the relative error is less than 0.5%.
Derradji, Mehdi; Zegaoui, Abdeldjalil; Xu, Yi-Le; Wang, An-ran; Dayo, Abdul Qadeer; Wang, Jun; Liu, Wen-bin; Liu, Yu-Guang; Khiari, Karim
2018-04-01
The phthalonitrile resins have claimed the leading place in the field of high performance polymers thanks to their combination of outstanding properties. The present work explores for the first time the gamma rays radiation resistance and shielding efficiency of the phthalonitrile resins and its related tungsten-reinforced nanocomposites. The primary goal of this research is to define the basic behavior of the phthalonitrile resins under highly ionizing gamma rays. The obtained results confirmed that the neat phthalonitrile resins can resist absorbed doses as high as 200 kGy. Meanwhile, the remarkable shielding efficiency of the phthalonitrile polymers was confirmed to be easily improved by preparing lead-free nanocomposites. In fact, the gamma rays screening ratio reached the exceptional value of 42% for the nanocomposites of 50 wt% of nano-tungsten loading. Thus, this study confirms that the remarkable performances of the phthalonitrile resins are not limited to the thermal and mechanical properties and can be extended to the gamma rays radiation and shielding resistances.
Radiation exposure to foetus and breasts from dental X-ray examinations: effect of lead shields.
Kelaranta, Anna; Ekholm, Marja; Toroi, Paula; Kortesniemi, Mika
2016-01-01
Dental radiography may involve situations where the patient is known to be pregnant or the pregnancy is noticed after the X-ray procedure. In such cases, the radiation dose to the foetus, though low, needs to be estimated. Uniform and widely used guidance on dental X-ray procedures during pregnancy are presently lacking, the usefulness of lead shields is unclear and practices vary. Upper estimates of radiation doses to the foetus and breasts of the pregnant patient were estimated with an anthropomorphic female phantom in intraoral, panoramic, cephalometric and CBCT dental modalities with and without lead shields. The upper estimates of foetal doses varied from 0.009 to 6.9 μGy, and doses at the breast level varied from 0.602 to 75.4 μGy. With lead shields, the foetal doses varied from 0.005 to 2.1 μGy, and breast doses varied from 0.002 to 10.4 μGy. The foetal dose levels without lead shielding were dental radiographic examination.
Low background gamma ray spectrometer using the anticoincidence shield technique at KAERI
International Nuclear Information System (INIS)
Byun, Jong In; Choi, Yun Ho; Kwak, Seung Im; Hwang, Han Yull; Chung, Kun Ho; Choi, Geun Sik; Park, Doo Won; Lee, Chang Woo
2002-01-01
We develop a ultra-low background gamma ray spectrometer, using active and passive shielding technique at the same time. Cosmic ray induced background is suppressed by means of active shield devices consisting of plastic scintillating plates of 50 mm thick and anti-coincidence electronic system. The shield is made of 150 mm thick walls of very low activity lead, especially 20 mm with activity of -1 and 0.36 s -1 with and without active shield, respectively, on the regions from 50 keV to 3 MeV. The detection efficiency curve has been precisely measured for regions from 80 keV to 2 MeV with a 10 3 ml marinelli beaker sample, made with calibrated mixed-sources consists of 109 Cd, 57 Co, 139 Ce, 203 Hg, 113 Sn, 85 Sr, 137 Cs, 60 Co and 88 Y. The virtues of the method are demonstrated by applying on experiment that requires the lowest detection limit
The magnetic shield design and simulation of an X-ray spectrometer for Chang'E mission
International Nuclear Information System (INIS)
Zhang Jiayu; Wang Huanyu; Zhang Chengmo; Yang Jiawei; Liang Xiaohua; Wang Jinzhou; Cao Xuelei; Gao Min; Cui Xingzhu; Peng Wenxi
2008-01-01
Basic design methods about the magnetic shield of an X-ray spectrometer for Chang'E Mission were introduced in this paper. The real magnetic field distribution was obtained through the measure experiment, and according to the measure results, the simulation to evaluate the magnetic shield effect was carded on. The results showed that the collimator can play a good role in magnetic shield to the electron. (authors)
Energy Technology Data Exchange (ETDEWEB)
Bagheri, Reza; Yousefinia, Hassan [Nuclear Fuel Cycle Research School (NFCRS), Nuclear Science and Technology Research Institute (NSTRI), Atomic Energy Organization of Iran, Tehran (Iran, Islamic Republic of); Moghaddam, Alireza Khorrami [Radiology Department, Paramedical Faculty, Mazandaran University of Medical Sciences, Sari (Iran, Islamic Republic of)
2017-02-15
In this work, linear and mass attenuation coefficients, effective atomic number and electron density, mean free paths, and half value layer and 10th value layer values of barium-bismuth-borosilicate glasses were obtained for 662 keV, 1,173 keV, and 1,332 keV gamma ray energies using MCNP-4C code and XCOM program. Then obtained data were compared with available experimental data. The MCNP-4C code and XCOM program results were in good agreement with the experimental data. Barium-bismuth-borosilicate glasses have good gamma ray shielding properties from the shielding point of view.
Directory of Open Access Journals (Sweden)
Reza Bagheri
2017-02-01
Full Text Available In this work, linear and mass attenuation coefficients, effective atomic number and electron density, mean free paths, and half value layer and 10th value layer values of barium–bismuth–borosilicate glasses were obtained for 662 keV, 1,173 keV, and 1,332 keV gamma ray energies using MCNP-4C code and XCOM program. Then obtained data were compared with available experimental data. The MCNP-4C code and XCOM program results were in good agreement with the experimental data. Barium–bismuth–borosilicate glasses have good gamma ray shielding properties from the shielding point of view.
Optimized shielding calculation to the transport of 131I employed in nuclear medicine
International Nuclear Information System (INIS)
Sahyun, A.; Sordi, G.M.; Rodrigues, D.; Sanches, M.P.; Romero F, C.R.
1996-01-01
The objective of this paper is to present the basis for shielding calculation used in different situations that could occur during the transport of 131 I utilized in nuclear medicine for diagnostic and therapeutic purposes. The aim of these calculation is to optimize the shielding in order to satisfy the transport of radioactive material. These calculations were proposed for estimated activities around 1,85 GBq (50mCi), 3,7 GBq(100mCi) and 7,4 GBq(200mCi), considering the driver of the cargo company and his assistant as the critical group and the general people considered as effect of collective dose. The population density considered in the models is the one related to Sao Paulo city, because the transport is done by the highway across the city and the radioactive material is distributed from west to north and south, where the airports are located. This area ranges a perimeter of 40 km. For the collective dose calculation, it was considered a population dose of less than 1/100 of the annual limit dose for the public. Our main concern is related to the large volume of radioactive material that is transported per week, specially because 1/3 of this material has activities around 3,7 GBq (100mCi). During the calculations, we have figured out that the activities at the moment of transport are nearly 40% greater than the one related to the calibration date. As for the discrepancy of official alpha value of US$10000/man-Sv and the real value for our country of US$3000/man-Sv,a comparative study was performed. (authors). 3 refs., 2 figs., 2 tabs
International Nuclear Information System (INIS)
Sato, Satoshi
2003-09-01
In tokamak-type DT nuclear fusion reactor, there are various type slits and ducts in the blanket and the vacuum vessel. The helium production in the rewelding location of the blanket and the vacuum vessel, the nuclear properties in the super-conductive TF coil, e.g. the nuclear heating rate in the coil winding pack, are enhanced by the radiation streaming through the slits and ducts, and they are critical concern in the shielding design. The decay gamma ray dose rate around the duct penetrating the blanket and the vacuum vessel is also enhanced by the radiation streaming through the duct, and they are also critical concern from the view point of the human access to the cryostat during maintenance. In order to evaluate these nuclear properties with good accuracy, three dimensional Monte Carlo calculation is required but requires long calculation time. Therefore, the development of the effective simple design evaluation method for radiation streaming is substantially important. This study aims to establish the systematic evaluation method for the nuclear properties of the blanket, the vacuum vessel and the Toroidal Field (TF) coil taking into account the radiation streaming through various types of slits and ducts, based on three dimensional Monte Carlo calculation using the MNCP code, and for the decay gamma ray dose rates penetrated around the ducts. The present thesis describes three topics in five chapters as follows; 1) In Chapter 2, the results calculated by the Monte Carlo code, MCNP, are compared with those by the Sn code, DOT3.5, for the radiation streaming in the tokamak-type nuclear fusion reactor, for validating the results of the Sn calculation. From this comparison, the uncertainties of the Sn calculation results coming from the ray-effect and the effect due to approximation of the geometry are investigated whether the two dimensional Sn calculation can be applied instead of the Monte Carlo calculation. Through the study, it can be concluded that the
Update of KASHIL-E6 library for shielding analysis and benchmark calculations
International Nuclear Information System (INIS)
Kim, D. H.; Kil, C. S.; Jang, J. H.
2004-01-01
For various shielding and reactor pressure vessel dosimetry applications, a pseudo-problem-independent neutron-photon coupled MATXS-format library based on the last release of ENDF/B-VI has been generated as a part of the update program for KASHIL-E6, which was based on ENDF/B-VI.5. It has VITAMIN-B6 neutron and photon energy group structures, i.e., 199 groups for neutron and 42 groups for photon. The neutron and photon weighting functions and the Legendre order of scattering are same as KASHIL-E6. The library has been validated through some benchmarks: the PCA-REPLICA and NESDIP-2 experiments for LWR pressure vessel facility benchmark, the Winfrith Iron88 experiment for validation of iron data, and the Winfrith Graphite experiment for validation of graphite data. These calculations were performed by the TRANSXlDANTSYS code system. In addition, the substitutions of the JENDL-3.3 and JEFF-3.0 data for Fe, Cr, Cu and Ni, which are very important nuclides for shielding analyses, were investigated to estimate the effects on the benchmark calculation results
Optimized design of shields for diagnostic X rays with NCRP 147 technique
International Nuclear Information System (INIS)
Gama T, G.
2006-01-01
A comparison among the design techniques of shielding for X-ray diagnostic rooms with the NCRP 49 (1976) report technique, AAPM 39 (1993) Y the one of the NCRP 147 (2005) technique. The designs correspond to a room of conventional X-rays, one of fluoroscopy, one of tomography Y one of mammography. In all the cases it demonstrates that the NCRP 49 technique overestimate the shieldings. The causes of the overestimation of the NCRP 49 can be attributed to: a) high values of the work charge that don't consider the spectral fluence of the photons that are present in each room, b) to the differences in the values of the kerma in air without attenuation for the dispersed primary radiation Y of leakage among both reports. (Author)
Guideline of Monte Carlo calculation. Neutron/gamma ray transport simulation by Monte Carlo method
2002-01-01
This report condenses basic theories and advanced applications of neutron/gamma ray transport calculations in many fields of nuclear energy research. Chapters 1 through 5 treat historical progress of Monte Carlo methods, general issues of variance reduction technique, cross section libraries used in continuous energy Monte Carlo codes. In chapter 6, the following issues are discussed: fusion benchmark experiments, design of ITER, experiment analyses of fast critical assembly, core analyses of JMTR, simulation of pulsed neutron experiment, core analyses of HTTR, duct streaming calculations, bulk shielding calculations, neutron/gamma ray transport calculations of the Hiroshima atomic bomb. Chapters 8 and 9 treat function enhancements of MCNP and MVP codes, and a parallel processing of Monte Carlo calculation, respectively. An important references are attached at the end of this report.
X-Ray Micro-Tomography Applied to Nasa's Materials Research: Heat Shields, Parachutes and Asteroids
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.
Geant4 calculations for space radiation shielding material Al2O3
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.
International Nuclear Information System (INIS)
Torres Pozas, S.; Monja Rey, P. de la; Sanchez Carrasca, M.; Yanez Lopez, D.; Macias Verde, D.; Martin Oliva, R.
2011-01-01
In recent years, the progress experienced in cancer treatment with ionizing radiation can deliver higher doses to smaller volumes and better shaped, making it necessary to take into account new aspects in the calculation of structural barriers. Furthermore, given that forecasts suggest that in the near future will install a large number of accelerators, or existing ones modified, we believe a useful tool to estimate the thickness of the structural barriers of treatment rooms. The shielding calculation methods are based on standard DIN 6847-2 and the recommendations given by the NCRP 151. In our experience we found only estimates originated from the DIN. Therefore, we considered interesting to develop an application that incorporates the formulation suggested by the NCRP, together with previous work based on the rules DIN allow us to establish a comparison between the results of both methods. (Author)
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.
Protective shielding parameters of diagnostic x-ray rooms in some hospitals in Benue State
International Nuclear Information System (INIS)
Agba, E.H.; Gemanam, S.; Sombo, T.
2011-01-01
Protective shielding parameters of diagnostic x-ray units at Federal Medical Centre, Makurdi, Baki Hospital, Gboko and Mkar Christian Hospital, Gboko have been determined using a radiation meter, (Inspector, Exp.S.E). The parameters determined include: Operating potential, Workload and Use factors of each diagnostic x-ray room. These parameters were used to estimate the primary and secondary protective barriers for the hospitals. The primary and secondary protective barrier values at Mkar Christian Hospital, Baki Hospital, Gboko and Federal Medical Centre, Makurdi are found to be: 11.0±0.11 x10 -1 mm and 9.0±9x10 -2 mm; 6.0±6.0x10 -1 mm and 6.0±6.0x10 -2 mm; and 7.0±7.0x10 -1 mm and 6.0±6.0x10 -2 mm respectively. The wall thicknesses around the x-ray rooms of the respective hospitals are 300±3.0x1 0 -1 mm for Mkar Christian Hospital and Federal Medical Centre, Makurdi, while that of Baki Hospital, Gboko is 270±2.7x10 -1 mm. The measured wall thicknesses are seen to be adequate protective structural shields on the basis of International NCRP Standards on Structural Shielding.
Schreckenbach, Georg
2002-12-16
In this and a previous article (J. Phys. Chem. A 2000, 104, 8244), the range of application for relativistic density functional theory (DFT) is extended to the calculation of nuclear magnetic resonance (NMR) shieldings and chemical shifts in diamagnetic actinide compounds. Two relativistic DFT methods are used, ZORA ("zeroth-order regular approximation") and the quasirelativistic (QR) method. In the given second paper, NMR shieldings and chemical shifts are calculated and discussed for a wide range of compounds. The molecules studied comprise uranyl complexes, [UO(2)L(n)](+/-)(q); UF(6); inorganic UF(6) derivatives, UF(6-n)Cl(n), n = 0-6; and organometallic UF(6) derivatives, UF(6-n)(OCH(3))(n), n = 0-5. Uranyl complexes include [UO(2)F(4)](2-), [UO(2)Cl(4)](2-), [UO(2)(OH)(4)](2-), [UO(2)(CO(3))(3)](4-), and [UO(2)(H(2)O)(5)](2+). For the ligand NMR, moderate (e.g., (19)F NMR chemical shifts in UF(6-n)Cl(n)) to excellent agreement [e.g., (19)F chemical shift tensor in UF(6) or (1)H NMR in UF(6-n)(OCH(3))(n)] has been found between theory and experiment. The methods have been used to calculate the experimentally unknown (235)U NMR chemical shifts. A large chemical shift range of at least 21,000 ppm has been predicted for the (235)U nucleus. ZORA spin-orbit appears to be the most accurate method for predicting actinide metal chemical shifts. Trends in the (235)U NMR chemical shifts of UF(6-n)L(n) molecules are analyzed and explained in terms of the calculated electronic structure. It is argued that the energy separation and interaction between occupied and virtual orbitals with f-character are the determining factors.
Measuring space radiation shielding effectiveness
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 ...
Structural shielding at irradiation tests with x-rays up to 400 kV
International Nuclear Information System (INIS)
Rabitsch, H.; Schachinger, E.
1979-12-01
Tables worked out in accordance to the dimensioning method following DIN 54113 are given for the determination protective barriers thicknesses against effective and stray radiation in practical operation conditions. The tables comprise irradiation operation with directional and non-directional radiation sources radiators beams collimated and are restricted to steel as dispersive material for the barrier thicknesses against stray radiation. The calculations of protective barriers against stray radiation show that operation with properly chosen primary beam limiting orifice yields in considerable savings in protective shield thickness when compared to operation without or with mismatched orifice to film size. It further shows a relative indifference against changes of the stray area (film size) or the focus - film sheet distance. The protective shield thickness for the most common exposure conditions calculated according to the Austrian Regulations for Radiation Protection are on the safe side when compared to protective barriere thicknesses calculated according to DIN 54113. (V.M.)
International Nuclear Information System (INIS)
Kim, Jae Cheon; Lee, Hwan Soo; Ha, Pham Nhu Viet; Kim, Soon Young; Shin, Chang Ho; Kim, Jong Kyung
2007-01-01
EASYQAD had been previously developed by using MATLAB GUI (Graphical User Interface) in order to perform conveniently gamma and neutron shielding calculations at Hanyang University. It had been completed as version α of radiation shielding analysis code. In this study, EASYQAD was upgraded to version β with many additional functions and more user-friendly graphical interfaces. For general users to run it on Windows XP environment without any MATLAB installation, this version was developed into a standalone code system
Dosimetric evaluation of photon dose calculation under jaw and MLC shielding
International Nuclear Information System (INIS)
Fogliata, A.; Clivio, A.; Vanetti, E.; Nicolini, G.; Belosi, M. F.; Cozzi, L.
2013-01-01
Purpose: The accuracy of photon dose calculation algorithms in out-of-field regions is often neglected, despite its importance for organs at risk and peripheral dose evaluation. The present work has assessed this for the anisotropic analytical algorithm (AAA) and the Acuros-XB algorithms implemented in the Eclipse treatment planning system. Specifically, the regions shielded by the jaw, or the MLC, or both MLC and jaw for flattened and unflattened beams have been studied.Methods: The accuracy in out-of-field dose under different conditions was studied for two different algorithms. Measured depth doses out of the field, for different field sizes and various distances from the beam edge were compared with the corresponding AAA and Acuros-XB calculations in water. Four volumetric modulated arc therapy plans (in the RapidArc form) were optimized in a water equivalent phantom, PTW Octavius, to obtain a region always shielded by the MLC (or MLC and jaw) during the delivery. Doses to different points located in the shielded region and in a target-like structure were measured with an ion chamber, and results were compared with the AAA and Acuros-XB calculations. Photon beams of 6 and 10 MV, flattened and unflattened were used for the tests.Results: Good agreement between calculated and measured depth doses was found using both algorithms for all points measured at depth greater than 3 cm. The mean dose differences (±1SD) were −8%± 16%, −3%± 15%, −16%± 18%, and −9%± 16% for measurements vs AAA calculations and −10%± 14%, −5%± 12%, −19%± 17%, and −13%± 14% for Acuros-XB, for 6X, 6 flattening-filter free (FFF), 10X, and 10FFF beams, respectively. The same figures for dose differences relative to the open beam central axis dose were: −0.1%± 0.3%, 0.0%± 0.4%, −0.3%± 0.3%, and −0.1%± 0.3% for AAA and −0.2%± 0.4%, −0.1%± 0.4%, −0.5%± 0.5%, and −0.3%± 0.4% for Acuros-XB. Buildup dose was overestimated with AAA, while Acuros-XB gave
International Nuclear Information System (INIS)
Alkhatib, Sari F.; Park, Chang Je; Jeong, Hae Yong; Lee, Yongdeok
2016-01-01
Highlights: • A simple formalism is suggested for the recursive approach and then it is used to produce buildup factors for certain multilayer shields. • The newly layer-splitting technique is implemented on the studied cases for testing the suggested formalism performance. • The buildup factors are generated using cubic polynomial fitting functions that are produced based on previous well-acknowledge data. - Abstract: This study illustrates the implementation of the newly suggested layer-splitting testing technique. This technique is introduced in order to be implemented in examining suggested formalisms for the recursive scheme (or iterative scheme). The recursive scheme is a concept used in treating and producing the gamma ray buildup factors in the case of multilayer shields. The layer-splitting technique simply enforces the scheme to treat a single layer of one material as two separated layers with similar characteristics. Thus it subjects the scheme to an abnormal definition of the multilayer shield that will test its performance in treating the successive layers. Thus, it will act as a method of verification for the approximations and assumptions taken in consideration. A simple formalism was suggested for the recursive scheme then the splitting technique was implemented on it. The results of implementing both the suggested formalism and the splitting technique are then illustrated and discussed. Throughout this study, cubic polynomial fitting functions were used to generate the data of buildup factors for the basic single-media that constitute the multilayer shields understudy. This study is limited to the cases of multiple shields consisting of repeated consecutive thin layers of lead–water and iron–water shields for 1 MeV gamma rays. The produced results of the buildup factor values through the implementation of the suggested formalism showed good consistency with the Monte Carlo simulation results of Lin and Jiang work. In the implementation of
The trade experiment: Shielding calculations for the building hosting the subcritical system
International Nuclear Information System (INIS)
Burn, K. W.; Carta, M.; Casalini, L.; Kadi, Y.; Monti, S.; Nava, E.; Palomba, M.; Petrovich, C.; Picardi, L.; Rubbia, C.; Troiani, F.
2005-01-01
The TRADE project (TRIGA Accelerator Driven Experiment), to be performed at the existing TRIGA reactor at ENEA Casaccia, has been proposed as a validation of the accelerator-driven system (ADS) concept. TRADE will be the first experiment in which the three main components of an ADS-the accelerator, spallation target and sub-critical blanket-are coupled at a power level sufficient to encounter reactivity feedback effects. As such, TRADE represents the necessary intermediate step in the development of hybrid transmutation systems, its expected outcomes being considered crucial-in terms of proof of stability of operation, dynamic behaviour and licensing issues-for the subsequent realisation of an ADS Transmutation Demonstrator. An essential role in the feasibility study of the experiment is played by radioprotection calculations. Such a system exhibits new characteristics with respect to a traditional reactor, owing to the presence of the proton accelerator. As beam losses always occur under normal operating conditions of an accelerator, shielding studies need to be performed not only around the reactor but also along the beam line from the accelerator to the spallation target. This paper illustrates a preliminary evaluation, using Monte Carlo methods, of the additional shielding to be located around the reactor structures, the beam transport line and the existing reactor building to allow access into the reactor hall and to restrict the doses outside to their legal limits. (authors)
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
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.
Investigations of some building materials for γ-rays shielding effectiveness
Mann, Kulwinder Singh; Kaur, Baljit; Sidhu, Gurdeep Singh; Kumar, Ajay
2013-06-01
For construction of residential and non-residential buildings bricks are used as building blocks. Bricks are made from mixtures of sand, clay, cement, fly ash, gypsum, red mud and lime. Shielding effectiveness of five soil samples and two fly ash samples have been investigated using some energy absorption parameters (Mass attenuation coefficients, mass energy absorption coefficients, KERMA (kinetic energy released per unit mass), HVL, equivalent atomic number and electron densities) firstly at 14 different energies from 81-1332 keV then extended to wide energy range 0.015-15 MeV. The soil sample with maximum shielding effectiveness has been used for making eight fly ash bricks [(Lime)0.15 (Gypsum)0.05 (Fly Ash)x (Soil)0.8-x, where values of x are from 0.4-0.7]. High Purity Germanium (HPGe) detector has been used for gamma-ray spectroscopy. The elemental compositions of samples were analysed using an energy dispersive X-ray fluorescence (EDXRF) spectrometer. The agreements of theoretical and experimental values of mass attenuation coefficient have been found to be quite satisfactory. It has been verified that common brick possess the maximum shielding effectiveness for wide energy range 0.015-15 MeV. The results have been shown graphically with some useful conclusions for making radiation safe buildings.
Effect of molybdenum on gamma ray shielding and structural properties of PbO-B2O3 glasses
Dogra, Mridula; Singh, K. J.; Kaur, Kulwinder
2018-04-01
The present study is aimed at developing new shielding materials for gamma ray shielding applications. Transparent glasses of the composition xMoO3-0.7PbO-(0.3-x)B2O3 where x= 0.03 to 0. 06 (mole fraction) have been prepared by using melt-quenchingtechnique. Gamma ray shielding properties have been evaluated in terms of mass attenuation coefficient and half value layer parameter at photon energies 662 and 1173 keV. These shielding parameters are also compared with standard shielding material`concretes'. It has been found that prepared glass system shows better shielding properties than barite and ordinary concretes proving the possibility of its usage as an alternate to conventional concrete for gamma ray shielding applications. The density, molar volume, X-Ray Diffraction, Fourier Transform InfraRed and Raman studies have been performed to study the structural properties of the glass system. It has been analyzed from FTIR and Raman studies that bridging oxygens increase with the decrease of MoO3 content in the glass composition.
Calculation of cosmic ray induced single event upsets: Program CRUP (Cosmic Ray Upset Program)
Shapiro, P.
1983-09-01
This report documents PROGRAM CRUP, COSMIC RAY UPSET PROGRAM. The computer program calculates cosmic ray induced single-event error rates in microelectronic circuits exposed to several representative cosmic-ray environments.
Inhomogeneity of neutron and gamma-ray attenuation in biological shields
Energy Technology Data Exchange (ETDEWEB)
El-bakkoush, F A; El-Ghobary, A M; Megahid, R M [Reactor and Neutron physics Department, Nuclear Research Center, A.E.A., Cairo (Egypt)
1997-12-31
Measurements have been carried-out to investigate the attenuation properties of some materials which are used as biological shields around nuclear radiation sources. Investigation was performed by measuring the transmitted fast neutron and gamma-spectra through cylindrical samples of magnetite- limonite, steel and cellulose shields. The neutron and gamma spectra were measured by a neutron-gamma spectrometer with stilbene scintillator. Discrimination between neutron and gamma pulses was achieved by a discrimination method. The obtained results are displayed in the form of neutron and gamma spectra and attenuation relations which are used to derive the total macroscopic cross-sections for neutrons and total linear attenuation coefficients for gamma-rays. The values of neutron and gamma relaxation lengths are also derived for the investigated materials. 10 figs., 1 tabs.
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)
Neutron Radiation Shielding For The NIF Streaked X-Ray Detector (SXD) Diagnostic
International Nuclear Information System (INIS)
Song, P; Holder, J; Young, B; Kalantar, D; Eder, D; Kimbrough, J
2006-01-01
The National Ignition Facility (NIF) at Lawrence Livermore National Laboratory (LLNL) is preparing for the National Ignition Campaign (NIC) scheduled in 2010. The NIC is comprised of several ''tuning'' physics subcampaigns leading up to a demonstration of Inertial Confinement Fusion (ICF) ignition. In some of these experiments, time-resolved x-ray imaging of the imploding capsule may be required to measure capsule trajectory (shock timing) or x-ray ''bang-time''. A capsule fueled with pure tritium (T) instead of a deutriun-tritium (DT) mixture is thought to offer useful physics surrogacy, with reduced yields of up to 5e14 neutrons. These measurements will require the use of the NIF streak x-ray detector (SXD). The resulting prompt neutron fluence at the planned SXD location (∼1.7 m from the target) would be ∼1.4e9/cm 2 . Previous measurements suggest the onset of significant background at a neutron fluence of ∼ 1e8/cm 2 . The radiation damage and operational upsets which starts at ∼1e8 rad-Si/sec must be factored into an integrated experimental campaign plan. Monte Carlo analyses were performed to predict the neutron and gamma/x-ray fluences and radiation doses for the proposed diagnostic configuration. A possible shielding configuration is proposed to mitigate radiation effects. The primary component of this shielding is an 80 cm thickness of Polyethylene (PE) between target chamber center (TCC) and the SXD diagnostic. Additionally, 6-8 cm of PE around the detector provide from the large number of neutrons that scatter off the inside of the target chamber. This proposed shielding configuration reduces the high-energy neutron fluence at the SXD by approximately a factor ∼50
Neutron Radiation Shielding For The NIF Streaked X-Ray Detector (SXD) Diagnostic
Energy Technology Data Exchange (ETDEWEB)
Song, P; Holder, J; Young, B; Kalantar, D; Eder, D; Kimbrough, J
2006-11-02
The National Ignition Facility (NIF) at Lawrence Livermore National Laboratory (LLNL) is preparing for the National Ignition Campaign (NIC) scheduled in 2010. The NIC is comprised of several ''tuning'' physics subcampaigns leading up to a demonstration of Inertial Confinement Fusion (ICF) ignition. In some of these experiments, time-resolved x-ray imaging of the imploding capsule may be required to measure capsule trajectory (shock timing) or x-ray ''bang-time''. A capsule fueled with pure tritium (T) instead of a deutriun-tritium (DT) mixture is thought to offer useful physics surrogacy, with reduced yields of up to 5e14 neutrons. These measurements will require the use of the NIF streak x-ray detector (SXD). The resulting prompt neutron fluence at the planned SXD location ({approx}1.7 m from the target) would be {approx}1.4e9/cm{sup 2}. Previous measurements suggest the onset of significant background at a neutron fluence of {approx} 1e8/cm{sup 2}. The radiation damage and operational upsets which starts at {approx}1e8 rad-Si/sec must be factored into an integrated experimental campaign plan. Monte Carlo analyses were performed to predict the neutron and gamma/x-ray fluences and radiation doses for the proposed diagnostic configuration. A possible shielding configuration is proposed to mitigate radiation effects. The primary component of this shielding is an 80 cm thickness of Polyethylene (PE) between target chamber center (TCC) and the SXD diagnostic. Additionally, 6-8 cm of PE around the detector provide from the large number of neutrons that scatter off the inside of the target chamber. This proposed shielding configuration reduces the high-energy neutron fluence at the SXD by approximately a factor {approx}50.
Calculated volumes of individual shield volcanoes at the young end of the Hawaiian Ridge
Robinson, Joel E.; Eakins, Barry W.
2006-03-01
High-resolution multibeam bathymetry and a digital elevation model of the Hawaiian Islands are used to calculate the volumes of individual shield volcanoes and island complexes (Niihau, Kauai, Oahu, the Maui Nui complex, and Hawaii), taking into account subsidence of the Pacific plate under the load of the Hawaiian Ridge. Our calculated volume for the Island of Hawaii and its submarine extent (213 × 10 3 km 3) is nearly twice the previous estimate (113 × 10 3 km 3), due primarily to crustal subsidence that had not been accounted for in the earlier work. The volcanoes that make up the Island of Hawaii (Mahukona, Kohala, Mauna Kea, Hualalai, Mauna Loa, Kilauea and Loihi) are generally considered to have been formed within the past million years, and our revised volume for the island indicates that magma supply rates are greater than previously estimated, 0.21 km 3/yr as opposed to ˜ 0.1 km 3/yr. This result also shows that compared with rates calculated for the Hawaiian Islands (0-6 Ma, 0.095 km 3/yr), the Hawaiian Ridge (0-45 Ma, 0.017 km 3/yr), and the Emperor Seamounts (45-80 Ma, 0.010 km 3/yr), magma supply rates have increased dramatically to build the Island of Hawaii.
Shielding calculations for neutron calibration bunker using Monte Carlo code MCNP-4C
International Nuclear Information System (INIS)
Suman, H.; Kharita, M. H.; Yousef, S.
2008-02-01
In this work, the dose arising from an Am-Be source of 10 8 neutron/sec strength located inside the newly constructed neutron calibration bunker in the National Radiation Metrology Laboratories, was calculated using MCNP-4C code. It was found that the shielding of the neutron calibration bunker is sufficient. As the calculated dose is not expected to exceed in inhabited areas 0.183 μSv/hr, which is 10 times smaller than the regulatory dose constraints. Hence, it can be concluded that the calibration bunker can house - from the external exposure point of view - an Am-Be neutron source of 10 9 neutron/sec strength. It turned out that the neutron dose from the source is few times greater than the photon dose. The sky shine was found to contribute significantly to the total dose. This contribution was estimated to be 60% of the neutron dose and 10% of the photon dose. The systematic uncertainties due to various factors have been assessed and was found to be between 4 and 10% due to concrete density variations; 15% due to the dose estimation method; 4 -10% due to weather variations (temperature and moisture). The calculated dose was highly sensitive to the changes in source spectra. The uncertainty due to the use of two different neutron spectra is about 70%.(author)
FFTF reactor-characterization program: gamma-ray measurements and shield characterization
International Nuclear Information System (INIS)
Bunch, W.L.; Moore, F.S. Jr.
1983-02-01
A series of experiments is to be made during the acceptance test program of the Fast Flux Test Facility (FFTF) to measure the gamma ray characteristics of the Fast Test Reactor (FTR) and to establish the performance characteristics of the reactor shield. These measurements are a part of the FFTF Reactor Characterization Program (RCP). Detailed plans have been developed for these experiments. During the initial phase of the Characteristics Program, which will be carried out in the In-Reactor Thimble (IRT), both active and passive measurement methods will be employed to obtain as much information concerning the gamma ray environment as is practical. More limited active gamma ray measurements also will be made in the Vibration Open Test Assembly (VOTA)
SCALE6 Hybrid Deterministic-Stochastic Shielding Methodology for PWR Containment Calculations
International Nuclear Information System (INIS)
Matijevic, Mario; Pevec, Dubravko; Trontl, Kresimir
2014-01-01
CADIS and also analog MC simulations, the FW-CADIS drastically improved MC dose rate calculations in quality as well in quantity. Large shielding problems such as portions and complete PWR facility require not only extensive computational resources but also understanding of the underlying physics, which is inevitable in interpreting results of hybrid deterministic-stochastic methodology. (authors)
A study of shielding properties of x-ray and gamma in barium compounds
Energy Technology Data Exchange (ETDEWEB)
Seenappa, L.; Manjunatha, H. C. [Dept. of of Physics, Government College for women, Karnataka (India); Chandrika, B. M. [PC Extension, St. Annes School, Karnataka (India); Chikka, Hanumantharayappa [Vivekananda Degree College, Karnataka (India)
2017-03-15
Ionizing radiation is known to be harmful to human health. The shielding of ionizing radiation depends on the attenuation which can be achieved by three main rules, i.e. time, distance and absorbing material. The mass attenuation coefficient, linear attenuation coefficient, Half Value Layer (HVL) and Tenth Value Layer (TVL) of X-rays (32 keV, 74 keV) and gamma rays (662 keV) are measured in Barium compounds. The measured values agree well with the theory. The effective atomic numbers (Z{sub eff}) and electron density (Ne) of Barium compounds have been computed in the wide energy region 1 keV to 100 GeV using an accurate database of photon-interaction cross sections and the WinXCom program. The mass attenuation coefficient and linear attenuation coefficient for BaCO{sub 3} is higher than the BaCl{sub 2}, Ba(No{sub 3}){sub 2} and BaSO{sub 4}. HVL, TVL and mean free path are lower for BaCO{sub 3} than the BaCl{sub 2}, Ba(No{sub 3}){sub 2} and BaSO{sub 4}. Among the studied barium compounds, BaCO{sub 3} is best material for x-ray and gamma shielding.
Poster - 28: Shielding of X-ray Rooms in Ontario in the Absence of Best Practice
Energy Technology Data Exchange (ETDEWEB)
Frimeth, Jeff; Richer, Jeff; Nesbitt, James [Xspect Inc., Windsor Regional Hospital, XRCT Ltd. (Canada)
2016-08-15
This poster will be strictly based on the Healing Arts Radiation Protection (HARP) Act, Regulation 543 under this Act (X-ray Safety Code), and personal communication the presenting author has had. In Ontario, the process of approval of an X-ray machine installation by the Director of the X-ray Inspection Service (XRIS) follows a certain protocol. Initially, the applicant submits a series of forms, including recommended shielding amounts, in order to satisfy the law. This documentation is then transferred to a third-party vendor (i.e. a professional engineer – P.Eng.) outsourced by the Ministry of Health and Long-term Care (MOHLTC). The P.Eng. then evaluates the submitted documentation for appropriate fulfillment of the HARP Act and Reg. 543 requirements. If the P.Eng.’s evaluation of the documentation is to their satisfaction, the XRIS is then notified. Finally, the Director will then issue a letter of approval to install the equipment at the facility. The methodology required to be used by the P.Eng. in order to determine the required amounts of protective barriers, and recommended to be used by the applicant, is contained within Safety Code 20A. However, Safety Code 35 has replaced the obsolete Safety Code 20A document and employs best practices in shielding design. This talk will focus further on specific intentions and limitations of Safety Code 20A. Furthermore, this talk will discuss the definition of the “practice of professional engineering” in Ontario. COMP members who are involved in shielding design are strongly encouraged to attend.
Poster - 28: Shielding of X-ray Rooms in Ontario in the Absence of Best Practice
International Nuclear Information System (INIS)
Frimeth, Jeff; Richer, Jeff; Nesbitt, James
2016-01-01
This poster will be strictly based on the Healing Arts Radiation Protection (HARP) Act, Regulation 543 under this Act (X-ray Safety Code), and personal communication the presenting author has had. In Ontario, the process of approval of an X-ray machine installation by the Director of the X-ray Inspection Service (XRIS) follows a certain protocol. Initially, the applicant submits a series of forms, including recommended shielding amounts, in order to satisfy the law. This documentation is then transferred to a third-party vendor (i.e. a professional engineer – P.Eng.) outsourced by the Ministry of Health and Long-term Care (MOHLTC). The P.Eng. then evaluates the submitted documentation for appropriate fulfillment of the HARP Act and Reg. 543 requirements. If the P.Eng.’s evaluation of the documentation is to their satisfaction, the XRIS is then notified. Finally, the Director will then issue a letter of approval to install the equipment at the facility. The methodology required to be used by the P.Eng. in order to determine the required amounts of protective barriers, and recommended to be used by the applicant, is contained within Safety Code 20A. However, Safety Code 35 has replaced the obsolete Safety Code 20A document and employs best practices in shielding design. This talk will focus further on specific intentions and limitations of Safety Code 20A. Furthermore, this talk will discuss the definition of the “practice of professional engineering” in Ontario. COMP members who are involved in shielding design are strongly encouraged to attend.
International Nuclear Information System (INIS)
Sasamoto, Nobuo; Kotegawa, Hiroshi
1984-11-01
In order to improve computational efficiency of PALLAS code, an accuracy is estimated on the neutron penetration calculation through a concrete shield, using bunched component nuclides of concrete. The calculated fast neutron flux is observed to depend weakly on how the nuclides are bunched. Contrary to this, the calculated thermal neutron fluxes are strongly dependent on the manner of bunching, mainly due to the fact that iron cross section has exceptionally large negative sensitivity to thermal neutron flux. (author)
On The Utilization of (1-X)Cu-X Pb) Alloys for Gamma-Rays Shielding
International Nuclear Information System (INIS)
Abd El-Latif, A.A.; Saeid, Kh.S.; Abd El-Latif, A.A.
2011-01-01
The present work deals with the study of the attenuation properties of gamma rays for [(1-X) Cu -X Pb] alloys where, x=10%, 20%, 30%, and 40% Pb waste by weight. Investigation has been performed by measuring the transmitted gamma ray spectra behind cylindrical samples of [(1-X) Cu - X Pb] alloys of different thicknesses. A collimated beam of gamma ray measured by using γ - ray spectrometer NaI(Tl) Scintillation detector with multichannel analyzer (MCA) cassy. Total mass attenuation coefficients (μ/ρ) of γ-ray have been evaluated and calculated using measured results and XCOM code respectively . Comparison between measured and calculated results shows a reasonable divergence at 0.511 MeV ,and 0.662 MeV γ-ray energies, in addition there is a convergence at 1.17 MeV, 1.274 MeV, and 1.3 MeV γ-ray energies
ZnO-PbO-B2O3 glasses as gamma-ray shielding materials
International Nuclear Information System (INIS)
Singh, Harvinder; Singh, Kulwant; Gerward, Leif; Singh, Kanwarjit; Sahota, Hari Singh; Nathuram, Rohila
2003-01-01
Values of the gamma-ray mass-attenuation coefficient, the photon mean free path (MFP), the effective atomic number and the effective electron density have been determined experimentally for xZnO · 2xPbO · (1-3x)B 2 O 3 (x=0.1-0.26) glasses at photon energies 511, 662, 1173 and 1332 keV and compared with theoretical data. The specific volume of the glasses has been derived from density measurements and studied as a function of composition. It is pointed out that these glasses have potential applications in radiation shielding
Church, Cody; Mawko, George; Archambault, John Paul; Lewandowski, Robert; Liu, David; Kehoe, Sharon; Boyd, Daniel; Abraham, Robert; Syme, Alasdair
2018-02-01
Radiopaque microspheres may provide intraprocedural and postprocedural feedback during transarterial radioembolization (TARE). Furthermore, the potential to use higher resolution x-ray imaging techniques as opposed to nuclear medicine imaging suggests that significant improvements in the accuracy and precision of radiation dosimetry calculations could be realized for this type of therapy. This study investigates the absorbed dose kernel for novel radiopaque microspheres including contributions of both short and long-lived contaminant radionuclides while concurrently quantifying the self-shielding of the glass network. Monte Carlo simulations using EGSnrc were performed to determine the dose kernels for all monoenergetic electron emissions and all beta spectra for radionuclides reported in a neutron activation study of the microspheres. Simulations were benchmarked against an accepted 90 Y dose point kernel. Self-shielding was quantified for the microspheres by simulating an isotropically emitting, uniformly distributed source, in glass and in water. The ratio of the absorbed doses was scored as a function of distance from a microsphere. The absorbed dose kernel for the microspheres was calculated for (a) two bead formulations following (b) two different durations of neutron activation, at (c) various time points following activation. Self-shielding varies with time postremoval from the reactor. At early time points, it is less pronounced due to the higher energies of the emissions. It is on the order of 0.4-2.8% at a radial distance of 5.43 mm with increased size from 10 to 50 μm in diameter during the time that the microspheres would be administered to a patient. At long time points, self-shielding is more pronounced and can reach values in excess of 20% near the end of the range of the emissions. Absorbed dose kernels for 90 Y, 90m Y, 85m Sr, 85 Sr, 87m Sr, 89 Sr, 70 Ga, 72 Ga, and 31 Si are presented and used to determine an overall kernel for the
Calculation of double energy angle differential neutron albedos for radiation shielding applications
International Nuclear Information System (INIS)
Litaize, O.; Diop, C.M.; Nimal, J.C.
2000-01-01
Void radiation shielding problems can be dealt with albedo concept which is an alternative to the complex bringing into operation of the 'exact' transport method calculations (SN, Monte Carlo). Up to here, differential albedos are used for single reflections from walls in the NARCISSE-3 propagation albedo code developed at CEA and used for project calculations. For taking into account the neutron multiple reflections on lacunar medium walls, double energy-angle differential albedos are needed. TRIPOLI-4 neutral particle transport Monte Carlo code in three dimensional geometries, has been chosen to implement a double differential albedo calculus routine and therefore to generate albedo data for different kinds of medium. The surfacic estimator, which could be used, is not enough efficient because all neutrons do not contribute to the result. A new estimator is carried out. At each collision site, during the neutron history simulation, it allows to compute the probability of the neutron to go through the medium and to come through the reflection surface in the direction and at the energy considered. This estimator is about hundred times more efficient than the surfacic estimator. (author)
International Nuclear Information System (INIS)
Ghassoun, Jillali; Jehoauni, Abdellatif
2000-01-01
In practice, the estimation of the flux obtained by Fredholm integral equation needs a truncation of the Neuman series. The order N of the truncation must be large in order to get a good estimation. But a large N induces a very large computation time. So the conditional Monte Carlo method is used to reduce time without affecting the estimation quality. In a previous works, in order to have rapid convergence of calculations it was considered only weakly diffusing media so that has permitted to truncate the Neuman series after an order of 20 terms. But in the most practical shields, such as water, graphite and beryllium the scattering probability is high and if we truncate the series at 20 terms we get bad estimation of flux, so it becomes useful to use high orders in order to have good estimation. We suggest two simple techniques based on the conditional Monte Carlo. We have proposed a simple density of sampling the steps for the random walk. Also a modified stretching factor density depending on a biasing parameter which affects the sample vector by stretching or shrinking the original random walk in order to have a chain that ends at a given point of interest. Also we obtained a simple empirical formula which gives the neutron flux for a medium characterized by only their scattering probabilities. The results are compared to the exact analytic solution, we have got a good agreement of results with a good acceleration of convergence calculations. (author)
Application of the MCNPX-McStas interface for shielding calculations and guide design at ESS
DEFF Research Database (Denmark)
Klinkby, Esben Bryndt; Bergbäck Knudsen, Erik; Willendrup, Peter Kjær
2014-01-01
Recently, an interface between the Monte Carlo code MCNPX and the neutron ray-tracing code MCNPX was developed [1, 2]. Based on the expected neutronic performance and guide geometries relevant for the ESS, the combined MCNPX-McStas code is used to calculate dose rates along neutron beam guides......, and by using newly developed event logging capability, the neutron state parameters corresponding to un-reflected neutrons are recorded at each scattering. This information is handed back to MCNPX where it serves as neutron source input for a second MCNPX simulation. This simulation enables calculation of dose...
Self-Shielding Treatment to Perform Cell Calculation for Seed Furl In Th/U Pwr Using Dragon Code
Directory of Open Access Journals (Sweden)
Ahmed Amin El Said Abd El Hameed
2015-08-01
Full Text Available Time and precision of the results are the most important factors in any code used for nuclear calculations. Despite of the high accuracy of Monte Carlo codes, MCNP and Serpent, in many cases their relatively long computational time leads to difficulties in using any of them as the main calculation code. Usually, Monte Carlo codes are used only to benchmark the results. The deterministic codes, which are usually used in nuclear reactor’s calculations, have limited precision, due to the approximations in the methods used to solve the multi-group transport equation. Self- Shielding treatment, an algorithm that produces an average cross-section defined over the complete energy domain of the neutrons in a nuclear reactor, is responsible for the biggest error in any deterministic codes. There are mainly two resonance self-shielding models commonly applied: models based on equivalence and dilution and models based on subgroup approach. The fundamental problem with any self-shielding method is that it treats any isotope as there are no other isotopes with resonance present in the reactor. The most practical way to solve this problem is to use multi-energy groups (50-200 that are chosen in a way that allows us to use all major resonances without self-shielding. In this paper, we perform cell calculations, for a fresh seed fuel pin which is used in thorium/uranium reactors, by solving 172 energy group transport equation using the deterministic DRAGON code, for the two types of self-shielding models (equivalence and dilution models and subgroup models Using WIMS-D5 and DRAGON data libraries. The results are then tested by comparing it with the stochastic MCNP5 code. We also tested the sensitivity of the results to a specific change in self-shielding method implemented, for example the effect of applying Livolant-Jeanpierre Normalization scheme and Rimman Integration improvement on the equivalence and dilution method, and the effect of using Ribbon
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.)
International Nuclear Information System (INIS)
Zitouni, Y.
1987-04-01
In the field of shielding, the requirement of radiation transport calculations in severe conditions, characterized by irreducible three-dimensional geometries has increased the use of the Monte Carlo method. The latter has proved to be the only rigorous and appropriate calculational method in such conditions. However, further efforts at optimization are still necessary to render the technique practically efficient, despite recent improvements in the Monte Carlo codes, the progress made in the field of computers and the availability of accurate nuclear data. Moreover, the personal experience acquired in the field and the control of sophisticated calculation procedures are of the utmost importance. The aim of the work which has been carried out is the gathering of all the necessary elements and features that would lead to an efficient utilization of the Monte Carlo method used in connection with shielding problems. The study of the general aspects of the method and the exploitation techniques of the MORSE code, which has proved to be one of the most comprehensive of the Monte Carlo codes, lead to a successful analysis of an actual case. In fact, the severe conditions and difficulties met have been overcome using such a stochastic simulation code. Finally, a critical comparison between calculated and high-accuracy experimental results has allowed the final confirmation of the methodology used by us
Revised Calculated Volumes Of Individual Shield Volcanoes At The Young End Of The Hawaiian Ridge
Robinson, J. E.; Eakins, B. W.
2003-12-01
Recent, high-resolution multibeam bathymetry and a digital elevation model of the Hawaiian Islands allow us to recalculate Bargar and Jackson's [1974] volumes of coalesced volcanic edifices (Hawaii, Maui-Nui, Oahu, Kauai, and Niihau) and individual shield volcanoes at the young end of the Hawaiian Ridge, taking into account subsidence of the Pacific plate under the load of the volcanoes as modeled by Watts and ten Brink [1989]. Our volume for the Island of Hawaii (2.48 x105 km3) is twice the previous estimate (1.13 x105 km3), due primarily to crustal subsidence, which had not been accounted for in the earlier work. The volcanoes that make up the Hawaii edifice (Mahukona, Kohala, Mauna Kea, Hualalai, Mauna Loa, Kilauea, and Loihi) are generally considered to have formed within the past million years and our revised volume for Hawaii indicates that either magma-supply rates are greater than previously estimated (0.25 km3/yr as opposed to 0.1 km3/yr) or that Hawaii's volcanoes have erupted over a longer period of time (>1 million years). Our results also indicate that magma supply rates have increased dramatically to build the Hawaiian edifices: the average rate of the past 5 million years (0.096 km3/yr) is substantially greater than the overall average of the Hawaiian Ridge (0.018km3/yr) or Emperor Seamounts (0.012 km3/yr) as calculated by Bargar and Jackson, and that rates within the past million years are greater still (0.25 km3/yr). References: Bargar, K. E., and Jackson, E. D., 1974, Calculated volumes of individual shield volcanoes along the Hawaiian-Emperor Chain, Jour. Research U.S. Geol. Survey, Vol. 2, No. 5, p. 545-550. Watts, A. B., and ten Brink, U. S., 1989, Crustal structure, flexure, and subsidence history of the Hawaiian Islands, Jour. Geophys. Res., Vol. 94, No. B8, p. 10,473-10,500.
Evaluation of gamma ray durability and its application of shielded RF tags
International Nuclear Information System (INIS)
Teraura, Nobuyuki; Ito, Kunio; Kobayashi, Daisuke; Sakurai, Kouichi
2015-01-01
In this study, the RF (Radio Frequency) tag with radiation shield is developed and its gamma ray durability is evaluated. RFID (RF Identification) is a radio-wave-based identification technology that can be used for various items. RF tags find use in many applications, including item tracing, access control, etc. RF tags can be classified as active RF tags, which have inbuilt voltaic cells, and passive RF tags without these cells. Passive RF tags, known for their low price and durability, are used in various fields. For instance, they are used for equipment maintenance in factories and thermal power plants. Several frequencies are used for RF tags. Further, RF tagging on the UHF (Ultra High Frequency) frequencies allows a communication range of approximately 10 m, and thus, remote reading is possible. When used in radiation environments such as in nuclear power plants, remote reading can contribute to the reduction of radiation exposure. However, because semiconductors are the primary elements used in the manufacture of RF tags, they can be damaged by radiation, and operational errors can occur. Therefore, this technology has not been used in environments affected by relatively high radiation levels. Therefore, in nuclear power plants, the use of RF tags is limited in areas of low radiation levels. In our study, we develop and manufacture a new RF tag with a radiation shield cover that provides error correction functionality. It is expected that radiation shielded RF tags will improve the radiation-proof feature, and its application range will be expanded. Using the radiation-proof RF tag, we have conducted radiation durability tests. These tests are of two types: one using low energy gamma ray, and the other using high-energy gamma ray. Experimental results are then analyzed. The number of applications for radiation shielded RF tags is considerably increasing, because it can be used in various radiation environments other than nuclear power plants as well, such as
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.
Directory of Open Access Journals (Sweden)
Adams Mike
2017-01-01
Full Text Available For nuclear waste packages the expected dose rates and nuclide inventory are beforehand calculated. Depending on the package of the nuclear waste deterministic programs like MicroShield® provide a range of results for each type of packaging. Stochastic programs like “Monte-Carlo N-Particle Transport Code System” (MCNP® on the other hand provide reliable results for complex geometries. However this type of program requires a fully trained operator and calculations are time consuming. The problem here is to choose an appropriate program for a specific geometry. Therefore we compared the results of deterministic programs like MicroShield® and stochastic programs like MCNP®. These comparisons enable us to make a statement about the applicability of the various programs for chosen types of containers. As a conclusion we found that for thin-walled geometries deterministic programs like MicroShield® are well suited to calculate the dose rate. For cylindrical containers with inner shielding however, deterministic programs hit their limits. Furthermore we investigate the effect of an inhomogeneous material and activity distribution on the results. The calculations are still ongoing. Results will be presented in the final abstract.
Adams, Mike; Smalian, Silva
2017-09-01
For nuclear waste packages the expected dose rates and nuclide inventory are beforehand calculated. Depending on the package of the nuclear waste deterministic programs like MicroShield® provide a range of results for each type of packaging. Stochastic programs like "Monte-Carlo N-Particle Transport Code System" (MCNP®) on the other hand provide reliable results for complex geometries. However this type of program requires a fully trained operator and calculations are time consuming. The problem here is to choose an appropriate program for a specific geometry. Therefore we compared the results of deterministic programs like MicroShield® and stochastic programs like MCNP®. These comparisons enable us to make a statement about the applicability of the various programs for chosen types of containers. As a conclusion we found that for thin-walled geometries deterministic programs like MicroShield® are well suited to calculate the dose rate. For cylindrical containers with inner shielding however, deterministic programs hit their limits. Furthermore we investigate the effect of an inhomogeneous material and activity distribution on the results. The calculations are still ongoing. Results will be presented in the final abstract.
Shield calculations, optimization vs. paradigm; Calculos de blindajes, optimizacion vs. paradigma
Energy Technology Data Exchange (ETDEWEB)
Cornejo D, N.; Hernandez S, A.; Martinez G, A. [Centro de Proteccion e Higiene de las Radiaciones, Calle 20 No. 4113 e/41 y 47 Playa C.P. 11300 LaHabana (Cuba)]. e-mail: nestor@cphr.edu.cu
2006-07-01
Many shieldings have been designed under the criteria of 'Maximum dose rates of project'. It has created the paradigm of those 'low dose rates', for the one which not few specialists would consider unacceptable levels of dose rate superior to the units of {mu}Sv.h{sup -1}, independently of the exposure times. At the present time numerous shieldings are being designed considering dose restrictions in real times of exposure. After these new shieldings, the dose rates could be notably superior to those after traditional shieldings, without it implies inadequate designs or constructive errors. In the work significant differences in levels of dose rates and thickness of shieldings estimated by both methods for some typical facilities. It was concluded that the use of real times of exposure is more adequate for the optimization of the Radiological Protection, although this method demands bigger care in its application. (Author)
Elevated gamma-rays shielding property in lead-free bismuth tungstate by nanofabricating structures
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.
Lead oxide-decorated graphene oxide/epoxy composite towards X-Ray radiation shielding
Hashemi, Seyyed Alireza; Mousavi, Seyyed Mojtaba; Faghihi, Reza; Arjmand, Mohammad; Sina, Sedigheh; Amani, Ali Mohammad
2018-05-01
In this study, employing modified Hummers method coupled with a multi-stage manufacturing procedure, graphene oxide (GO) decorated with Pb3O4 (GO-Pb3O4) at different weight ratios was synthesized. Thereupon, via the vacuum shock technique, composites holding GO-Pb3O4 at different filler loadings (5 and 10 wt%) and thicknesses (4 and 6 mm) were fabricated. Successful decoration of GO with Pb3O4 was confirmed via FTIR analysis. Moreover, particle size distribution of the produced fillers was examined using particle size analyzer. X-ray attenuation examination revealed that reinforcement of epoxy-based composites with GO-Pb3O4 led to a significant improvement in the overall attenuation rate of X-ray beam. For instance, composites containing 10 wt% GO-Pb3O4 with 6 mm thickness showed 4.06, 4.83 and 3.91 mm equivalent aluminum thickness at 40, 60 and 80 kVp energies, denoting 124.3, 124.6 and 103.6% improvement in the X-ray attenuation rate compared to a sample holding neat epoxy resin, respectively. Simulation results revealed that the effect of GO-Pb3O4 loading on the X-ray shielding performance undermined with increase in the voltage of the applied X-ray beam.
International Nuclear Information System (INIS)
Nascimento, Erika M.; Lopes Filho, Ferdinand J.; Souza, Milena Thays B. de; Aragao Filho, Geraldo L.
2013-01-01
The thickness of shielding for controlled and surrounding areas must be considered in terms of the PET-CT equipment installation. However, for a project to install a PET-CT requires the participation of technologists, engineers and architects so that together we can meet the requirements of radiological protection at low cost, enabling its installation and maintenance in nuclear medical centers and hospitals. The objective of this paper is to describe the calculations needed to shield a nuclear medicine center that will install a PET-CT equipment and describe how the participation of other professionals can contribute to a lower cost
Calculation of point isotropic buildup factors of gamma rays for water and lead
Directory of Open Access Journals (Sweden)
A. S. H.
2001-12-01
Full Text Available Exposure buildup factors for water and lead have been calculated by the Monte-Carlo method for an isotropic point source in an infinite homogeneous medium, using the latest cross secions available on the Internet. The types of interactions considered are ,photoelectric effect, incoherent (or bound-electron Compton. Scattering, coherent (or Rayleigh scattering and pair production. Fluorescence radiations have also been taken into acount for lead. For each material, calculations were made at 10 gamma ray energies in the 40 keV to 10 MeV range and up to penetration depths of 10 mean free paths at each energy point. The results presented in this paper can be considered as modified gamma ray exposure buildup factors and be used in radiation shielding designs.
International Nuclear Information System (INIS)
Berezhkov, A.B.; Gordeeva, E.K.; Mazanov, V.L.; Solov'ev, V.Yu.; Ryabov, A.V.; Khokhlov, V.F.; Shejno, I.N.
1987-01-01
Programs for obtaining phantom images when calculating the radiation shield structure for nuclear-engineering plants, using computer graphics, are developed. Programs are designed to accompany calculational investigations using the SUPER2/RRI3-PICSCH program and ZAMOK-TOMOGRAF program comutering complexes. Design geometry techniques, allowing to present three-dimensional object in the form of two-dimensional perspective projection to the screen plane, are realized in the programs
Adams Mike; Smalian Silva
2017-01-01
For nuclear waste packages the expected dose rates and nuclide inventory are beforehand calculated. Depending on the package of the nuclear waste deterministic programs like MicroShield® provide a range of results for each type of packaging. Stochastic programs like “Monte-Carlo N-Particle Transport Code System” (MCNP®) on the other hand provide reliable results for complex geometries. However this type of program requires a fully trained operator and calculations are time consuming. The prob...
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.
Gamma-ray shielding properties of CaO–SrO–B2O3 glasses
DEFF Research Database (Denmark)
Singh, K; Singh, H; Sharma, G
2005-01-01
, the molar volume of the glasses has been derived from density measurements, and the excess volume has been determined as a function of composition. The reported data should be useful for potential applications of these glasses in radiation shielding. (C) 2004 Elsevier Ltd. All rights reserved.......Mass attenuation coefficients, effective atomic numbers and effective electron densities have been determined experimentally for the glass system xCaO . (0.3-x)SrO . 0.7B(2)O(3) at photon energies 511, 662, 1173, and 1332KeV. The results are compared with theoretical calculations. In addition...
Energy Technology Data Exchange (ETDEWEB)
Kondo, Keitaro, E-mail: kondo.keitaro@jaea.go.jp; Arbeiter, Frederik; Fischer, Ulrich; Lu, Lei; Qiu, Yuefeng; Tian, Kuo
2015-10-15
Highlights: • A detailed geometry model with pipe penetrations and gaps was prepared for the IFMIF test cell. • The neutron streaming effect due to gaps and pipes with shielding plugs was investigated. • The present analysis revealed that the streaming effect can be mitigated by some counter measures. • Occupational workers can access to the room above the test cell during operation. - Abstract: The IFMIF Test Cell (TC) design was developed and optimized in the EVEDA phase, and finally the reference TC design was proposed. The present study is devoted to further investigations of open issues on the reference TC design. In order to examine the neutron streaming effect caused by pipe penetrations and gaps around removable shielding plugs, a new geometry model for neutronic analyses has been prepared directly from engineering CAD data by utilizing the McCad conversion software. All removable shielding plugs are separately described in the model and a detailed description of pipes was incorporated into the model. The calculation result suggests that the streaming effect is mitigated if the pipe penetration is designed appropriately, while the gaps around the shielding plugs above the TC have large impact on the radiation dose in the access cell. The concept of the reference TC design has been basically validated from the neutronics point of view, although the streaming effect should be compensated by the shielding capability of the test cell cover plate so that occupational workers can access to the access cell during operation.
Directory of Open Access Journals (Sweden)
Shane Stimpson
2017-09-01
Full Text Available An essential component of the neutron transport solver is the resonance self-shielding calculation used to determine equivalence cross sections. The neutron transport code, MPACT, is currently using the subgroup self-shielding method, in which the method of characteristics (MOC is used to solve purely absorbing fixed-source problems. Recent efforts incorporating multigroup kernels to the MOC solvers in MPACT have reduced runtime by roughly 2×. Applying the same concepts for self-shielding and developing a novel lumped parameter approach to MOC, substantial improvements have also been made to the self-shielding computational efficiency without sacrificing any accuracy. These new multigroup and lumped parameter capabilities have been demonstrated on two test cases: (1 a single lattice with quarter symmetry known as VERA (Virtual Environment for Reactor Applications Progression Problem 2a and (2 a two-dimensional quarter-core slice known as Problem 5a-2D. From these cases, self-shielding computational time was reduced by roughly 3–4×, with a corresponding 15–20% increase in overall memory burden. An azimuthal angle sensitivity study also shows that only half as many angles are needed, yielding an additional speedup of 2×. In total, the improvements yield roughly a 7–8× speedup. Given these performance benefits, these approaches have been adopted as the default in MPACT.
International Nuclear Information System (INIS)
Stimpson, Shane G.; Liu, Yuxuan; Collins, Benjamin S.; Clarno, Kevin T.
2017-01-01
An essential component of the neutron transport solver is the resonance self-shielding calculation used to determine equivalence cross sections. The neutron transport code, MPACT, is currently using the subgroup self-shielding method, in which the method of characteristics (MOC) is used to solve purely absorbing fixed-source problems. Recent efforts incorporating multigroup kernels to the MOC solvers in MPACT have reduced runtime by roughly 2×. Applying the same concepts for self-shielding and developing a novel lumped parameter approach to MOC, substantial improvements have also been made to the self-shielding computational efficiency without sacrificing any accuracy. These new multigroup and lumped parameter capabilities have been demonstrated on two test cases: (1) a single lattice with quarter symmetry known as VERA (Virtual Environment for Reactor Applications) Progression Problem 2a and (2) a two-dimensional quarter-core slice known as Problem 5a-2D. From these cases, self-shielding computational time was reduced by roughly 3–4×, with a corresponding 15–20% increase in overall memory burden. An azimuthal angle sensitivity study also shows that only half as many angles are needed, yielding an additional speedup of 2×. In total, the improvements yield roughly a 7–8× speedup. Furthermore given these performance benefits, these approaches have been adopted as the default in MPACT.
International Nuclear Information System (INIS)
Kondo, Keitaro; Arbeiter, Frederik; Fischer, Ulrich; Lu, Lei; Qiu, Yuefeng; Tian, Kuo
2015-01-01
Highlights: • A detailed geometry model with pipe penetrations and gaps was prepared for the IFMIF test cell. • The neutron streaming effect due to gaps and pipes with shielding plugs was investigated. • The present analysis revealed that the streaming effect can be mitigated by some counter measures. • Occupational workers can access to the room above the test cell during operation. - Abstract: The IFMIF Test Cell (TC) design was developed and optimized in the EVEDA phase, and finally the reference TC design was proposed. The present study is devoted to further investigations of open issues on the reference TC design. In order to examine the neutron streaming effect caused by pipe penetrations and gaps around removable shielding plugs, a new geometry model for neutronic analyses has been prepared directly from engineering CAD data by utilizing the McCad conversion software. All removable shielding plugs are separately described in the model and a detailed description of pipes was incorporated into the model. The calculation result suggests that the streaming effect is mitigated if the pipe penetration is designed appropriately, while the gaps around the shielding plugs above the TC have large impact on the radiation dose in the access cell. The concept of the reference TC design has been basically validated from the neutronics point of view, although the streaming effect should be compensated by the shielding capability of the test cell cover plate so that occupational workers can access to the access cell during operation.
Using MCBEND for neutron or gamma-ray deterministic calculations
Directory of Open Access Journals (Sweden)
Geoff Dobson
2017-01-01
Full Text Available MCBEND 11 is the latest version of the general radiation transport Monte Carlo code from AMEC Foster Wheeler’s ANSWERS® Software Service. MCBEND is well established in the UK shielding community for radiation shielding and dosimetry assessments. MCBEND supports a number of acceleration techniques, for example the use of an importance map in conjunction with Splitting/Russian Roulette. MCBEND has a well established automated tool to generate this importance map, commonly referred to as the MAGIC module using a diffusion adjoint solution. This method is fully integrated with the MCBEND geometry and material specification, and can easily be run as part of a normal MCBEND calculation. An often overlooked feature of MCBEND is the ability to use this method for forward scoping calculations, which can be run as a very quick deterministic method. Additionally, the development of the Visual Workshop environment for results display provides new capabilities for the use of the forward calculation as a productivity tool. In this paper, we illustrate the use of the combination of the old and new in order to provide an enhanced analysis capability. We also explore the use of more advanced deterministic methods for scoping calculations used in conjunction with MCBEND, with a view to providing a suite of methods to accompany the main Monte Carlo solver.
Using MCBEND for neutron or gamma-ray deterministic calculations
Geoff, Dobson; Adam, Bird; Brendan, Tollit; Paul, Smith
2017-09-01
MCBEND 11 is the latest version of the general radiation transport Monte Carlo code from AMEC Foster Wheeler's ANSWERS® Software Service. MCBEND is well established in the UK shielding community for radiation shielding and dosimetry assessments. MCBEND supports a number of acceleration techniques, for example the use of an importance map in conjunction with Splitting/Russian Roulette. MCBEND has a well established automated tool to generate this importance map, commonly referred to as the MAGIC module using a diffusion adjoint solution. This method is fully integrated with the MCBEND geometry and material specification, and can easily be run as part of a normal MCBEND calculation. An often overlooked feature of MCBEND is the ability to use this method for forward scoping calculations, which can be run as a very quick deterministic method. Additionally, the development of the Visual Workshop environment for results display provides new capabilities for the use of the forward calculation as a productivity tool. In this paper, we illustrate the use of the combination of the old and new in order to provide an enhanced analysis capability. We also explore the use of more advanced deterministic methods for scoping calculations used in conjunction with MCBEND, with a view to providing a suite of methods to accompany the main Monte Carlo solver.
Contribution to gamma ray transport calculation in heterogeneous media
International Nuclear Information System (INIS)
Bourdet, L.
1985-04-01
This thesis presents the development of gamma transport calculation codes in three dimension heterogeneous geometries. These codes allow us to define the protection against gamma-rays or verify their efficiency. The laws that govern the interactions of gamma-rays with matters are briefly revised. A library with the all necessary constants for these codes is created. TRIPOLI-2, a code that treats in exact way the neutron transport in matters using Monte-Carlo method, has been adapted to deal with the transport of gamma-rays in matters as well. TRINISHI, a code which considers only one collision, has been realized to treat heterogeneous geometries containing voids. Elaborating a formula that calculates the albedo for gamma-ray reflection (the code ALBANE) allows us to solve the problem of gamma-ray reflection on plane surfaces. NARCISSE-2 deals with gamma-rays that suffer only one reflection on the inner walls of any closed volume (rooms, halls...) [fr
Aggregate effects on γ-ray shielding characteristics and compressive strength on concrete
International Nuclear Information System (INIS)
Oh, Jeong Hwan; Choi, Soo Seok; Mun, Young Bun; Lee, Jae Hyung; Choi, Hyun Kook
2016-01-01
We observed the γ-ray shielding characteristics and compressive strength of five types of concrete using general aggregates and high-weight aggregates. The aggregates were classified into fine aggregate and coarse aggregate according to the average size. The experimental results obtained an attenuation coefficient of 0.371 cm-1 from a concrete with the oxidizing slag sand (OSS) and oxidizing slag gravel (OSG) for a γ-ray of "1"3"7Cs, which is improved by 2% compared with a concrete with typical aggregates of sand and gravel. In the unit weight measurement, a concrete prepared by iron ore sand (IOS) and OSG had the highest value of 3,175 kg·m"-"3. Although the unit weight of the concrete with OSS and OSG was 3,052 kg·m"-"3, which was lower than the maximum unit weight condition by 123 kg·m"-"3, its attenuation coefficient was improved by 0.012 cm-1. The results of chemical analysis of aggregates revealed that the magnesium content in oxidizing slag was lower than that in iron ore, while the calcium content was higher. The concrete with oxidizing slag aggregates demonstrated enhanced γ-ray shielding performance due to a relatively high calcium content compared with the concrete with OSS and OSG in spite of a low unit weight. All sample concretes mixed with high-weight aggregates had higher compressive strength than the concrete with typical sand and gravel. When OSS and IOS were used, the highest compressive strength was 50.2 MPa, which was an improvement by 45% over general concrete, which was achieved after four weeks of curing
Aggregate effects on γ-ray shielding characteristics and compressive strength on concrete
Energy Technology Data Exchange (ETDEWEB)
Oh, Jeong Hwan; Choi, Soo Seok [Jeju National University, Jeju (Korea, Republic of); Mun, Young Bun; Lee, Jae Hyung; Choi, Hyun Kook [Sungshin Cement Co., Ltd, Sejong (Korea, Republic of)
2016-12-15
We observed the γ-ray shielding characteristics and compressive strength of five types of concrete using general aggregates and high-weight aggregates. The aggregates were classified into fine aggregate and coarse aggregate according to the average size. The experimental results obtained an attenuation coefficient of 0.371 cm-1 from a concrete with the oxidizing slag sand (OSS) and oxidizing slag gravel (OSG) for a γ-ray of {sup 137}Cs, which is improved by 2% compared with a concrete with typical aggregates of sand and gravel. In the unit weight measurement, a concrete prepared by iron ore sand (IOS) and OSG had the highest value of 3,175 kg·m{sup -3}. Although the unit weight of the concrete with OSS and OSG was 3,052 kg·m{sup -3}, which was lower than the maximum unit weight condition by 123 kg·m{sup -3}, its attenuation coefficient was improved by 0.012 cm-1. The results of chemical analysis of aggregates revealed that the magnesium content in oxidizing slag was lower than that in iron ore, while the calcium content was higher. The concrete with oxidizing slag aggregates demonstrated enhanced γ-ray shielding performance due to a relatively high calcium content compared with the concrete with OSS and OSG in spite of a low unit weight. All sample concretes mixed with high-weight aggregates had higher compressive strength than the concrete with typical sand and gravel. When OSS and IOS were used, the highest compressive strength was 50.2 MPa, which was an improvement by 45% over general concrete, which was achieved after four weeks of curing.
International Nuclear Information System (INIS)
Peng, Y.K.M.; Marcus, F.B.; Dory, R.A.; Moore, J.R.
1975-01-01
A poloidal magnetic field shielding system is proposed for a tokamak EPR. This coil system minimizes the pulsed poloidal field that intersects the TF (toroidal field) coils and hence reduces the risk of superconductor quenching and structural failure of the coils. Based on an idealized shielding model, we have determined the configurations for the OH (ohmic heating), the S-VF (shield-vertical field), and the T-VF (trimming-vertical field) coils in a typical tokamak EPR. It is found that the pulsed poloidal field strength is greatly reduced in the TF coil region. The overall requirement in stored plasma and vertical field energy is also substantially reduced when compared with conventional EPR designs. Use of this field shielding system is expected to enhance reliability of the superconducting TF coils in a tokamak EPR
Application of the MCNPX-McStas interface for shielding calculations and guide design at ESS
DEFF Research Database (Denmark)
Klinkby, Esben Bryndt; Bergbäck Knudsen, Erik; Willendrup, Peter Kjær
2013-01-01
. The generation and moderation of neutrons is simulated using a full scale MCNPX model of the ESS target monolith. Upon entering the beam extraction region, the individual neutron states are handed to McStas via the MCNPX-McStas interface. McStas transports the neutrons through the beam guide and by using newly......Recently, an interface between the Monte Carlo code MCNPX and the neutron ray-tracing code MCNPX was developed[1]. Based on the expected neutronic performance and guide geometries relevant for the ESS, the combined MCNPX-McStas code is used to calculate dose rates along neutron beam guides...... developed event logging capability, the neutron state parameters corresponding to un-reflected neutrons are recorded at each scattering. This information is handed back to MCNPX where it serves as neutron source input for a second MCNPX simulation. This simulation enables calculation of dose rates...
Neutron shielding point kernel integral calculation code for personal computer: PKN-pc
International Nuclear Information System (INIS)
Kotegawa, Hiroshi; Sakamoto, Yukio; Nakane, Yoshihiro; Tomita, Ken-ichi; Kurosawa, Naohiro.
1994-07-01
A personal computer version of PKN code, PKN-pc, has been developed to calculate neutron and secondary gamma-ray 1cm depth dose equivalents in water, ordinary concrete and iron for neutron source. Characteristics of PKN code are, to able to calculate dose equivalents in multi-layer three-dimensional system, which are described with two-dimensional surface, for monoenergetic neutron source from 0.01 to 14.9 MeV, 252 Cf fission and 241 Am-Be neutron source quick and easily. In addition to these features, the PKN-pc is possible to process interactive input and to get graphical system configuration and graphical results easily. (author)
Gamma ray shielding and structural properties of PbO-P2O5-Na2WO4 glass system
Dogra, Mridula; Singh, K. J.; Kaur, Kulwinder; Anand, Vikas; Kaur, Parminder
2017-05-01
The present work has been undertaken to study the gamma ray shielding properties of PbO-P2O5-Na2WO4 glass system. The values of mass attenuation coefficient and half value layer parameter at photon energies 511, 662 and 1173 KeV have been determined using XCOM computer software developed by National Institute of Standards and Technology. The density, molar volume, XRD, UV-VIS and Raman studies have been performed to study the structural properties of the prepared glass system to check the possibility of the use of prepared samples as an alternate to conventional concrete for gamma ray shielding applications.
Calculation of characteristics of X-ray devices
Directory of Open Access Journals (Sweden)
Orobinskyi A. N.
2015-12-01
Full Text Available Actuality of this work is related to human radiation safety during tuning and regulation of X-ray devices in the process of their development and production. The more precise the calculations for the device are, the less time is required for its tuning and regulation, and thus people are less exposed to radiation. When developing an X-ray device, it is necessary to choose an X-ray tube and filters taking into account the application domain of the device. In order to do this, one should know anode voltage, X-ray tube anode current, material and thickness of filters, i.e. to calculate these characteristics at the set quality of X-ray radiation. The known published studies do not give any solution to this problem. The scientific novelty of this work is that it establishes the interdependence between main characteristics of the X-ray device: the function of the device defines the quality of X-ray radiation (mean photon energy and air kerma power; mean photon energy depends on the X-ray anode tube voltage and spectral resolution; air kerma power depends on anode tube voltage, current of X-ray tube anode, spectral resolution, thicknesses of the filters and their materials; spectral resolution depends on thicknesses of filters and their materials; thickness of filters depends on the material of the filter (the linear coefficient of weakening of X-ray radiation. Knowledge of interdependence of basic characteristics of the X-ray devices allowes developing simple algorithm for their calculation at the values of homogeneity coefficient from 0,8 to 1, which makes it possible to choose an X-ray tube and filters with the purpose of obtaining X-ray radiation of the set quality.
International Nuclear Information System (INIS)
Coste-Delclaux, M.; Aggery, A.; Huot, N.
2005-01-01
APOLLO2 is a modular multigroup transport code developed by Cea in Saclay. Until last year, the self-shielding module could only treat one resonant isotope mixed with moderator isotopes. Consequently, the resonant mixture self-shielding treatment was an iterative one. Each resonant isotope of the mixture was treated separately, the other resonant isotopes of the mixture being then considered as moderator isotopes, that is to say non-resonant isotopes. This treatment could be iterated. Last year, we have developed a new method that consists in treating the resonant mixture as a unique entity. A main feature of APOLLO2 self-shielding module is that some implemented models are very general and therefore very powerful and versatile. We can give, as examples, the use of probability tables in order to describe the microscopic cross-section fluctuations or the TR slowing-down model that can deal with any resonance shape. The self-shielding treatment of a resonant mixture was developed essentially thanks to these two models. The calculations of a simplified Jules Horowitz reactor using a Monte-Carlo code (TRIPOLI4) as a reference and APOLLO2 in its standard and improved versions, show that, as far as the effective multiplication factor is concerned, the mixture treatment does not bring an improvement, because the new treatment suppresses compensation between the reaction rate discrepancies. The discrepancy of 300 pcm that appears with the reference calculation is in accordance with the technical specifications of the Jules Horowitz reactor
Design of a PET/CT facility considering the shielding calculation in accordance with AAPM TG-108
International Nuclear Information System (INIS)
Guevara R, V. Y.; Romero C, N.; Berrocal T, M.
2014-08-01
A Positron Emission Tomography / Computed Tomography facility may require protection barriers on floor, ceiling and walls, because the patient becomes a radioactive source that emits photons of 0.511 MeV, after having received a radiopharmaceutical, usually F-18 fluorodeoxyglucose (F-18 FDG). This work has as objective to propose the design of a PET/CT facility, taking into account technical and radiation protection considerations applied internationally, and also develop the necessary shielding for such installation by applying as published by the American Association of Physicists in Medicine Task Group Report 108. A shielding spreadsheet in Excel program was developed with reference to the recommendations of the AAPM TG - 08, to determine the shielding required for the walls, floor and ceiling. For fixing the radiation levels in the shielding calculation has been considered the actual restrictions for the occupationally exposed personnel (100 μSv/week) as well as the people in general (20 μSv/ week). The radiopharmaceutical used as a reference for the shielding calculation was the F-18 FDG. With the assistance of an architectural plan were determined distances from potential sources of radiation in facility (uptake and image acquisition living rooms) to points of interest around them. Finally the thickness of the protective barriers in lead and concrete necessary to achieve the established radiation levels were calculated and these results were stored in a table. This paper shows that technical aspects considered in the design of the installation and environments distribution can improve work processes within the PET/CT facility, consequently resulting in a reduction of the dose levels for people in general. (author)
International Nuclear Information System (INIS)
Greene, B.F.; Neistat, M.D.
1983-01-01
An unobtrusive observation system was developed to determine the extent to which dental professionals in two communities provided lead shielding to patients during X-ray exams. A lengthy baseline revealed low and irregular provision of shielding among half of these professionals. Subsequently, a program was undertaken by a consumer's group in which these professionals were requested to provide shielding and were given confidential feedback regarding its use during the baseline period. The provision of shielding dramatically increased at all offices and was maintained throughout a follow-up period extending to more than 9 months after the program's implementation. Little or no generalized effect was observed in the occurrence of three collateral behaviors that were also assessed throughout the study
Attenuation of 60Co gamma rays by barium acrylic resin composite shields
International Nuclear Information System (INIS)
Abdulla, Riaz; Fidha, Mariyam; Sripathi Rao, B.H.; Kudkuli, Jagadish; Rekha, P.D.; Sharma, S.D.
2015-01-01
Oral squamous cell carcinoma is the sixth most common cancer reported globally, with an annual incidence of over 300,000 cases, of which 62% arise in developing countries. Radiation therapy is a treatment modality that uses ionizing radiation as a therapeutic agent. It is widely employed in the treatment of head and neck cancer, as a primary therapy coupled with surgical procedure and chemotherapy or as a palliative treatment for advanced tumors. However, radiotherapy can cause a series of complications such as xerostomia, mucositis, osteoradionecrosis, and radiation caries. Composite circular disc containing different ratios of acrylic and barium sulfate (BaSO 4 ) were made in-house. The purpose of this study was to evaluate the percentage attenuation from these composite shields in 60 Co gamma rays. A maximum of 8% radiation attenuation was achieved using 1:4 ratio of acrylic-BaSO 4 composite shields. The study proposes BaSO 4 as one of the compounds in combination with acrylic resin or any other thermoplastic substances for making biocompatible radiation attenuating devices. (author)
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)
Adapting an x-ray/debris shield to the cascade ICF power plant: Neutronics issues
International Nuclear Information System (INIS)
Tobin, M.T.
1990-01-01
A neutronics analysis has been carried out to determine the effects on the Cascade ICF reactor concept of adding a solid-lithium x-ray and debris shield to each ICF capsule. Results indicate that tritium breeding in LiAlO 2 is possible with a modest isotopic enhancement in 6 Li (to 15%). The shallow-burial index is greater than 1 (indicating that deep burial may be required) if the blanket is kept in the reactor for more than 2.5 yr. Nine percent of the total thermal power is unrecoverable. Parts of the chamber wall may require replacement once during the reactor life due to radiation damage. Part of the SiC chamber end cap must be replaced annually. The reactor may not require any nuclear-grade construction. 20 refs., 4 figs., 3 tabs
International Nuclear Information System (INIS)
Johnson, J.O.; Miller, L.F.; Kam, F.B.K.
1981-05-01
A neutronics model for the Oak Ridge National Laboratory Bulk Shielding Reactor (ORNL-SAR) was developed and verified by experimental measurements. A cross-section library was generated from the 218 group Master Library using the AMPX Block Code system. A series of one-, two-, and three-dimensional neutronics calculations were performed utilizing both transport and diffusion theory. Spectral comparison was made with 58 Ni(n,p) reaction. The results of the comparison between the calculational model and other experimental measurements showed agreement within 10% and therefore the model was determined to be adequate for calculating the neutron fluence for future irradiation experiments in the ORNL-BSR
Energy Technology Data Exchange (ETDEWEB)
Johnson, J.O.; Miller, L.F.; Kam, F.B.K.
1981-05-01
A neutronics model for the Oak Ridge National Laboratory Bulk Shielding Reactor (ORNL-SAR) was developed and verified by experimental measurements. A cross-section library was generated from the 218 group Master Library using the AMPX Block Code system. A series of one-, two-, and three-dimensional neutronics calculations were performed utilizing both transport and diffusion theory. Spectral comparison was made with /sup 58/Ni(n,p) reaction. The results of the comparison between the calculational model and other experimental measurements showed agreement within 10% and therefore the model was determined to be adequate for calculating the neutron fluence for future irradiation experiments in the ORNL-BSR.
International Nuclear Information System (INIS)
Blum, P.; Cagnon, R.; Nimal, J.C.
1982-01-01
This report gives the results of a campaign of gamma dose rates measurement in the vicinity of a transport package loaded with 12 PWR spent fuel assemblies, so that the characteristics of the package and the fuel. It describes the measuring methods, and gives the accuracy of the data which will be usefull, as benchmarks, to the control of the calculation methods used to verify the gamma shielding of the packages. It shows how to calculate gamma dose rates from the data given on the package and the fuel, and gives the results of a calculation with the Mecure IV code and compares them to the measurements
Characterization of barite and crystal glass as attenuators in X-ray and gamma radiation shieldings
International Nuclear Information System (INIS)
Almeida Junior, Airton Tavares de
2005-03-01
Aiming to determine the barium sulphate (BaSO 4 ) ore and crystal glass attenuation features, both utilized as shieldings against ionizing X and gamma radiations in radiographic installations, a study of attenuation using barite plaster and barite concrete was carried out, which are used, respectively, on wall coverings and in block buildings. The crystal glass is utilized in screens and in windows. To do so, ten plates of barite plaster and three of barite concrete with 900 cm 2 and with an average thickness ranging from 1 to 5 cm, and three plates of crystal glass with 323 cm 2 and with thicknesses of 1, 2 and 4 cm were analyzed. The samples were irradiated with X-rays with potentials of 60, 80, 110 and 150 kilovolts, and also with 60 Co gamma rays. Curves of attenuation were obtained for barite plaster and barite concrete (mGy/mA.min) and (mGy/h), both at 1 meter, as a function of thickness and curve of transmission through barite plaster and barite concrete as a function of the thickness. The equivalent thicknesses of half and tenth value layers for barite plaster, barite concrete and crystal glass for all X-Ray energies were also determined. (author)
International Nuclear Information System (INIS)
Kaewkhao, J.; Pokaipisit, A.; Limsuwan, P.
2010-01-01
In this work, the mass attenuation coefficients and shielding parameters of borate glass matrices containing with Bi 2 O 3 and BaO have been investigated at 662 keV, and compare with PbO in same glass structure. The theoretical values were calculated by WinXCom software and compare with experiential data. The results found that the mass attenuation coefficients were increased with increasing of Bi 2 O 3 , BaO and PbO concentration, due to increase photoelectric absorption of all glass samples. However, Compton scattering gives dominant contribution to the total mass attenuation coefficients for studied glass samples. Moreover the half value layers (HVL) of glass samples were also better than ordinary concretes and commercial window glass. These results reflecting that the Bi-based glass can use replace Pb in radiation shielding glass. In the case of Ba, may be can use at appropriate energy such as X-rays or lower.
International Nuclear Information System (INIS)
Ganesan, S.; Gopalakrishnan, V.; Ramanadhan, M.M.; Cullan, D.E.
1986-01-01
We investigate the effect of interpolation error in the pre-processing codes LINEAR, RECENT and SIGMA1 on calculations of self-shielding factors and their temperature derivatives. We consider the 2.0347 to 3.3546 keV energy region for 238 U capture, which is the NEACRP benchmark exercise on unresolved parameters. The calculated values of temperature derivatives of self-shielding factors are significantly affected by interpolation error. The sources of problems in both evaluated data and codes are identified and eliminated in the 1985 version of these codes. This paper helps to (1) inform code users to use only 1985 versions of LINEAR, RECENT, and SIGMA1 and (2) inform designers of other code systems where they may have problems and what to do to eliminate their problems. (author)
International Nuclear Information System (INIS)
Ganesan, S.; Gopalakrishnan, V.; Ramanadhan, M.M.; Cullen, D.E.
1985-01-01
The authors investigate the effect of interpolation error in the pre-processing codes LINEAR, RECENT and SIGMA1 on calculations of self-shielding factors and their temperature derivatives. They consider the 2.0347 to 3.3546 keV energy region for /sup 238/U capture, which is the NEACRP benchmark exercise on unresolved parameters. The calculated values of temperature derivatives of self-shielding factors are significantly affected by interpolation error. The sources of problems in both evaluated data and codes are identified and eliminated in the 1985 version of these codes. This paper helps to (1) inform code users to use only 1985 versions of LINEAR, RECENT, and SIGMA1 and (2) inform designers of other code systems where they may have problems and what to do to eliminate their problems
Shielding and activation calculations around the reactor core for the MYRRHA ADS design
Ferrari, Anna; Mueller, Stefan; Konheiser, J.; Castelliti, D.; Sarotto, M.; Stankovskiy, A.
2017-09-01
In the frame of the FP7 European project MAXSIMA, an extensive simulation study has been done to assess the main shielding problems in view of the construction of the MYRRHA accelerator-driven system at SCK·CEN in Mol (Belgium). An innovative method based on the combined use of the two state-of-the-art Monte Carlo codes MCNPX and FLUKA has been used, with the goal to characterize complex, realistic neutron fields around the core barrel, to be used as source terms in detailed analyses of the radiation fields due to the system in operation, and of the coupled residual radiation. The main results of the shielding analysis are presented, as well as the construction of an activation database of all the key structural materials. The results evidenced a powerful way to analyse the shielding and activation problems, with direct and clear implications on the design solutions.
Energy Technology Data Exchange (ETDEWEB)
Endo, Akira; Kawasaki, Katsuya; Kikuchi, Masamitsu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Harada, Yasunori
1997-07-01
Measurement has been made to study distributions of {gamma}-ray emitters induced in the concrete shield of the JAERI electron linac facility. Core boring was carried out at seven positions to take samples from the concrete shield, and {gamma}-ray counting rates and {gamma}-ray spectra of these samples were measured with a NaI(Tl) detector and a Ge semiconductor detector, respectively. The following radionuclides were detected in the concrete samples: {sup 60}Co, {sup 134}Cs, {sup 152}Eu and {sup 154}Eu generated through thermal neutron capture reaction, and {sup 22}Na and {sup 54}Mn generated through nuclear reactions by bremsstrahlung and fast neutrons. The relation between the distributions of {gamma}-ray emitters, as a function of the depth of concrete, and the positions of core boring is discussed. (author)
Calculation of self–shielding factor for neutron activation experiments using GEANT4 and MCNP
Energy Technology Data Exchange (ETDEWEB)
Romero–Barrientos, Jaime, E-mail: jaromero@ing.uchile.cl [Comisión Chilena de Energía Nuclear, Nueva Bilbao 12501, Las Condes, Santiago (Chile); Universidad de Chile, DFI, Facultad de Ciencias Físicas Y Matemáticas, Avenida Blanco Encalada 2008, Santiago (Chile); Molina, F. [Comisión Chilena de Energía Nuclear, Nueva Bilbao 12501, Las Condes, Santiago (Chile); Aguilera, Pablo [Comisión Chilena de Energía Nuclear, Nueva Bilbao 12501, Las Condes, Santiago (Chile); Universidad de Chile, Depto. de Física, Facultad de Ciencias, Las Palmeras 3425, Ñuñoa, Santiago (Chile); Arellano, H. F. [Universidad de Chile, DFI, Facultad de Ciencias Físicas Y Matemáticas, Avenida Blanco Encalada 2008, Santiago (Chile)
2016-07-07
The neutron self–shielding factor G as a function of the neutron energy was obtained for 14 pure metallic samples in 1000 isolethargic energy bins from 1·10{sup −5}eV to 2·10{sup 7}eV using Monte Carlo simulations in GEANT4 and MCNP6. The comparison of these two Monte Carlo codes shows small differences in the final self–shielding factor mostly due to the different cross section databases that each program uses.
Adjustable lead glass shielding device for use with a over-the-table x-ray tube
International Nuclear Information System (INIS)
Eubig, C.; Groves, B.M.; Davey, G.
1978-01-01
Sources of scattered radiation exposure to personnel from a ceiling-mounted x-ray tube were examined at the side of cardiac catheterization patients. A fully adjustable mounting for a lead glass shield was designed to afford maximum radiation protection to the attending physician's head and neck area, while minimizing interference with the procedure
Adjustable lead glass shielding device for use with an over-the-table x-ray tube.
Eubig, C; Groves, B M; Davey, G
1978-12-01
Sources of scattered radiation exposure to personnel from a ceiling-mounted x-ray tube were examined at the side of cardiac catheterization patients. A fully adjustable mounting for a lead glass shield was designed to afford maximum radiation protection to the attending physician's head and neck area, while minimizing interference with the procedure.
International Nuclear Information System (INIS)
Parminder Kaur; Singh, K.J.; Kulwinder Kaur; Anand, Vikas; Dogra, Mridula
2017-01-01
Glasses doped with heavy metal oxides have been proposed to shield the hazardous gamma rays originating from nuclear reactors as alternative to the conventional concretes. In this work, transparent glasses with composition 65Bi_2O_3-xBaO-(35-x) B_2O_3 (with x =0, 4, 8 wt %) have been prepared by using melt quenching technique in the laboratory. XRD and FTIR studies have been undertaken to explore the structural properties. The amorphous nature of the prepared samples is confirmed by XRD studies. Structural changes in the system have been explored by FTIR studies. The FTIR results reveal the conversion of (BO_3) triangular units to (BO_4) tetrahedral units with the addition of barium oxide along with the creation of non-bridging oxygen in the prepared glass system. Gamma-ray shielding properties have been explored with the help of WinXCom software developed by National Institute Standards and Technology at photon energy 662 keV. Gamma ray shielding properties in terms of mass attenuation coefficient, half value layer, tenth value layer and mean free path have been found to be superior as compared to the ordinary as well as barite concrete. Therefore, it is speculated that our prepared glass samples can serve as better gamma ray shielding materials. (author)
Yoshizawa, Terutaka; Zou, Wenli; Cremer, Dieter
2017-04-01
A new method for calculating nuclear magnetic resonance shielding constants of relativistic atoms based on the two-component (2c), spin-orbit coupling including Dirac-exact NESC (Normalized Elimination of the Small Component) approach is developed where each term of the diamagnetic and paramagnetic contribution to the isotropic shielding constant σi s o is expressed in terms of analytical energy derivatives with regard to the magnetic field B and the nuclear magnetic moment 𝝁 . The picture change caused by renormalization of the wave function is correctly described. 2c-NESC/HF (Hartree-Fock) results for the σiso values of 13 atoms with a closed shell ground state reveal a deviation from 4c-DHF (Dirac-HF) values by 0.01%-0.76%. Since the 2-electron part is effectively calculated using a modified screened nuclear shielding approach, the calculation is efficient and based on a series of matrix manipulations scaling with (2M)3 (M: number of basis functions).
International Nuclear Information System (INIS)
Cai Ling; Kosov, Daniel S.; Fushman, David
2011-01-01
We performed density functional calculations of backbone 15 N shielding tensors in the regions of beta-sheet and turns of protein G. The calculations were carried out for all twenty-four beta-sheet residues and eight beta-turn residues in the protein GB3 and the results were compared with the available experimental data from solid-state and solution NMR measurements. Together with the alpha-helix data, our calculations cover 39 out of the 55 residues (or 71%) in GB3. The applicability of several computational models developed previously (Cai et al. in J Biomol NMR 45:245–253, 2009) to compute 15 N shielding tensors of alpha-helical residues is assessed. We show that the proposed quantum chemical computational model is capable of predicting isotropic 15 N chemical shifts for an entire protein that are in good correlation with experimental data. However, the individual components of the predicted 15 N shielding tensor agree with experiment less well: the computed values show much larger spread than the experimental data, and there is a profound difference in the behavior of the tensor components for alpha-helix/turns and beta-sheet residues. We discuss possible reasons for this.
International Nuclear Information System (INIS)
Park, J; Lee, J; Kim, H; Kim, I; Ye, S
2015-01-01
Purpose: To evaluate the effect of a tungsten eye-shield on the dose distribution of a patient. Methods: A 3D scanner was used to extract the dimension and shape of a tungsten eye-shield in the STL format. Scanned data was transferred into a 3D printer. A dummy eye shield was then produced using bio-resin (3D systems, VisiJet M3 Proplast). For a patient with mucinous carcinoma, the planning CT was obtained with the dummy eye-shield placed on the patient’s right eye. Field shaping of 6 MeV was performed using a patient-specific cerrobend block on the 15 x 15 cm 2 applicator. The gantry angle was 330° to cover the planning target volume near by the lens. EGS4/BEAMnrc was commissioned from our measurement data from a Varian 21EX. For the CT-based dose calculation using EGS4/DOSXYZnrc, the CT images were converted to a phantom file through the ctcreate program. The phantom file had the same resolution as the planning CT images. By assigning the CT numbers of the dummy eye-shield region to 17000, the real dose distributions below the tungsten eye-shield were calculated in EGS4/DOSXYZnrc. In the TPS, the CT number of the dummy eye-shield region was assigned to the maximum allowable CT number (3000). Results: As compared to the maximum dose, the MC dose on the right lens or below the eye shield area was less than 2%, while the corresponding RTP calculated dose was an unrealistic value of approximately 50%. Conclusion: Utilizing a 3D scanner and a 3D printer, a dummy eye-shield for electron treatment can be easily produced. The artifact-free CT images were successfully incorporated into the CT-based Monte Carlo simulations. The developed method was useful in predicting the realistic dose distributions around the lens blocked with the tungsten shield
Bulk density calculations from prompt gamma ray yield
International Nuclear Information System (INIS)
Naqvi, A.A.; Nagadi, M.M.; Al-Amoudi, O.S.B.; Maslehuddin, M.
2006-01-01
Full text: The gamma ray yield from a Prompt Gamma ray Neutron Activation Analysis (PGNAA) setup is a linear function of element concentration and neutron flux in a the sample with constant bulk density. If the sample bulk density varies as well, then the element concentration and the neutron flux has a nonlinear correlation with the gamma ray yield [1]. The measurement of gamma ray yield non-linearity from samples and a standard can be used to estimate the bulk density of the samples. In this study the prompt gamma ray yield from Blast Furnace Slag, Fly Ash, Silica Fumes and Superpozz cements samples have been measured as a function of their calcium and silicon concentration using KFUPM accelerator-based PGNAA setup [2]. Due to different bulk densities of the blended cement samples, the measured gamma ray yields have nonlinear correlation with calcium and silicon concentration of the samples. The non-linearity in the yield was observed to increase with gamma rays energy and element concentration. The bulk densities of the cement samples were calculated from ratio of gamma ray yield from blended cement and that from a Portland cement standard. The calculated bulk densities have good agreement with the published data. The result of this study will be presented
International Nuclear Information System (INIS)
Huggett, J.; Mukonoweshuro, W.; Loader, R.
2013-01-01
Three commercially available in-plane patient organ shields (barium eye, bismuth eye and bismuth breast) for computed tomography (CT) examinations were evaluated to determine their effectiveness for dose reduction. Absorbed doses were measured using metal oxide semiconductor field effect transistor dosemeters fastened to a Kyoto CT Torso phantom. Resultant images were visually compared with those minus shielding by an experienced radiologist. Approximate dose reductions of 21, 38 and 50 % were achieved by the barium eye, bismuth eye and bismuth breast shields, respectively, at a cost of increased image noise and streak artefacts. Shielded images produced varied levels of image artefact, particularly those resulting from the eye shields. Measured dose reductions were not consistent with the potential dose savings stated by the manufacturers of the shields. When evaluating the breast shield, similar dose reduction was achieved without shield-induced artefact by simply reducing the X-ray tube current. (authors)
Kramer, S. L.; Ghosh, V. J.; Breitfeller, M.; Wahl, W.
2016-11-01
Third generation high brightness light sources are designed to have low emittance and high current beams, which contribute to higher beam loss rates that will be compensated by Top-Off injection. Shielding for these higher loss rates will be critical to protect the projected higher occupancy factors for the users. Top-Off injection requires a full energy injector, which will demand greater consideration of the potential abnormal beam miss-steering and localized losses that could occur. The high energy electron injection beam produces significantly higher neutron component dose to the experimental floor than a lower energy beam injection and ramped operations. Minimizing this dose will require adequate knowledge of where the miss-steered beam can occur and sufficient EM shielding close to the loss point, in order to attenuate the energy of the particles in the EM shower below the neutron production threshold (weaknesses in the design before a high radiation incident occurs. The effort required to adequately define the accelerator geometry for these codes has been greatly reduced with the implementation of the graphical interface of FLAIR to FLUKA. This made the effective shielding process for NSLS-II quite accurate and reliable. The principles used to provide supplemental shielding to the NSLS-II accelerators and the lessons learned from this process are presented.
Shielding features of quarry stone
International Nuclear Information System (INIS)
Hernandez V, C.; Contreras S, H.; Hernandez A, L.; Baltazar R, A.; Escareno J, E.; Mares E, C. A.; Vega C, H. R.
2010-10-01
Quarry stone lineal attenuation coefficient for gamma-rays has been obtained. In Zacatecas, quarry stone is widely utilized as a decorative item in buildings, however its shielding features against gamma-rays unknown. The aim of this work is to determine the shielding properties of quarry stone against γ-rays using Monte Carlo calculations where a detailed model of a good geometry experimental setup was carried out. In the calculations 10 pieces 10 X 10 cm 2 of different thickness were utilized to evaluate the photons transmission as the quarry stone thickness is increased. It was noticed that transmitted photons decay away as the shield thickness is increased, these results were fitted to an exponential function were the linear attenuation coefficient was estimated. Also, using XCOM code the linear attenuation coefficient from several keV up to 100 MeV was estimated. From the comparison between Monte Carlo results and XCOM calculations a good agreement was found. For 0.662 MeV γ-rays the attenuation coefficient of quarry stone, whose density is 2.413 g-cm -3 , is 0.1798 cm -1 , this mean a X 1/2 = 3.9 cm, X 1/4 = 7.7 cm, X 1/10 = 12.8 cm, and X 1/100 = 25.6 cm. Having the information of quarry stone performance as shielding give the chance to use this material to shield X and γ-ray facilities. (Author)
High performance dosimetry calculations using adapted ray-tracing
International Nuclear Information System (INIS)
Perrotte, Lancelot; Saupin, Guillaume
2010-01-01
When preparing interventions on nuclear sites, it is interesting to study different scenarios, to identify the most appropriate one for the operator(s). Using virtual reality tools is a good way to simulate the potential scenarios. Thus, taking advantage of very efficient computation times can help the user studying different complex scenarios, by immediately evaluating the impact of any changes. In the field of radiation protection, people often use computation codes based on the straight line attenuation method with build-up factors. As for other approaches, geometrical computations (finding all the interactions between radiation rays and the scene objects) remain the bottleneck of the simulation. We present in this paper several optimizations used to speed up these geometrical computations, using innovative GPU ray-tracing algorithms. For instance, we manage to compute every intersection between 600 000 rays and a huge 3D industrial scene in a fraction of second. Moreover, our algorithm works the same way for both static and dynamic scenes, allowing easier study of complex intervention scenarios (where everything moves: the operator(s), the shielding objects, the radiation sources).
Neutron and gamma ray streaming calculations for the ETF neutral beam injectors
International Nuclear Information System (INIS)
Lillie, R.A.; Santoro, R.T.; Alsmiller, R.G. Jr.; Barnes, J.M.
1981-02-01
Two-dimensional radiation transport methods have been used to estimate the effects of neutron and gamma ray streaming on the performance of the Engineering Test Facility (ETF) neutral beam injectors. The calculations take into account the spatial, angular, and spectral distributions of the radiation entering the injector duct. The instantaneous nuclear heating rate averaged over the length of the cryopumping panel in the injector is 7.5 x 10 -3 MW/m 3 which implies a total heat load of 2.2 x 10 -4 MW. The instantaneous dose rate to the ion gun insulators was estimated to be 3200 rad/s. The radial dependence of the instantaneous dose equivalent rate in the neutral beam injector duct shield was also calculated
ZnO-PbO-B2O3 glasses as gamma-ray shielding materials
DEFF Research Database (Denmark)
Singh, H.; Singh, K.; Gerward, Leif
2003-01-01
Values of the gamma-ray mass-attenuation coefficient, the photon mean free path (MFP), the effective atomic number and the effective electron density have been determined experimentally for xZnO.2xPbO.(1-3x)B2O3 (x = 0.1-0.26) glasses at photon energies 511, 662, 1173 and 1332 keV and compared wi...... with theoretical data. The specific volume of the glasses has been derived from density measurements and studied as a function of composition. It is pointed out that these glasses have potential applications in radiation shielding.......Values of the gamma-ray mass-attenuation coefficient, the photon mean free path (MFP), the effective atomic number and the effective electron density have been determined experimentally for xZnO.2xPbO.(1-3x)B2O3 (x = 0.1-0.26) glasses at photon energies 511, 662, 1173 and 1332 keV and compared...
Crystal glass used for X ray and gamma radiation shielding - Part two
International Nuclear Information System (INIS)
Antonio Filho, Joao
2007-01-01
Crystal glass has been widely used as shielding material in gamma radiation sources as well as x-ray generating equipment to replace the plumbiferous glass, in order to minimize exposure to individuals. However, properties of the radiation attenuation of crystal glass commercially available in Brazil, for the different types of energy are not known. For this reason, this work was carried out aiming to determine the radiation attenuation, transmission curves and Half Value Layer. In this work, ten plates of crystal glass, with dimensions of 20 cm x 20 cm and range of thicknesses from 0.5 to 2.0 cm, were used. The plates were X-ray irradiated with potential constants of 60, 80, 110, 150 kV and gamma radiation of 60 Co. Analysis in the properties of the 60 Co radiation attenuation of barite plaster and barite concrete commercially available in Brazil were also carried out. The curves of attenuation and of transmission were obtained for crystal glass, barite plaster and barite concrete (mGy/mA.min) at 1 meter as a function of thickness. The thickness equivalent of a half value layer and deci value layer of crystal glass for all types of radiation and energies studied was also determined. (author)
Fundamentals for cost calculations of X-ray equipment
International Nuclear Information System (INIS)
Bossard, F.
1985-01-01
Economic implications of running an X-ray departement in Switzerland will be illustrated by comparing operating costs of private radiological institutes with and without CT to the operating costs of large radiological departments in hospitals with and without CT and to the operating costs of simple X-ray equipment in general practicioners' offices. - These costs calculations form the basis for cost-benefit analyses. (orig.) [de
Dogra, Mridula; Singh, K. J.; Kaur, Kulwinder
2018-05-01
Quaternary system of the composition (0.15+x) Bi2O3-(0.55-x) B2O3-0.15CdO-0.15Na2O (where x=0, 0.1, 0.3 and 0.5 mole fraction) has been synthesized using melt-quenching technique. Gamma ray shielding properties are measured in terms of mass attenuation coefficient and half value layer at photon energies 662, 1173 and 1332 keV. These parameters are compared with standard nuclear radiation shielding `barite and ferrite' concretes. The results reflect better radiation shielding properties as compared to barite and ferrite concretes. Effective atomic number is calculated at photon energies 662 and 1173 keV. Density, molar volume and XRD studies are analyzed to know physical and structural properties of the glass system. Optical band gap, refractive index and molar refraction are calculated from UV-Visible measurements. Decrease in optical band gap and increase in molar refraction have been observed indicating the increase of non-bridging oxygens in the structure.
International Nuclear Information System (INIS)
Palma, Daniel A.; Goncalves, Alessandro C.; Martinez, Aquilino S.; Silva, Fernando C.
2008-01-01
The activation technique allows much more precise measurements of neutron intensity, relative or absolute. The technique requires the knowledge of the Doppler broadening function ψ(x,ξ) to determine the resonance self-shielding factors in the epithermal range G epi (τ,ξ). Two new analytical approximations for the Doppler broadening function ψ(x,ξ) are proposed. The approximations proposed are compared with other methods found in literature for the calculation of the ψ(x,ξ) function, that is, the 4-pole Pade method and the Frobenius method, when applied to the calculation of G epi (τ,ξ). The results obtained provided satisfactory accuracy. (authors)
International Nuclear Information System (INIS)
Kazarnovskij, M.V.; Matushko, G.K.; Matushko, V.L.; Par'ev, Eh.Ya.; Serezhnikov, S.V.
1981-01-01
The problem on propagation of the internuclear cascade initiated by nucleons of 0.1-1 GeV energy in accelerator schielding is solved approximately in the analytical form. Analytical expressions for the function of spatial, angular and energy distribution of the flux density of nucleons with the energy above 20 MeV and some functionals from it are obtained. The results of the calculations obtained by the developed methods are compared with calculations obtained by the method of direct simulation. It is shown that at the atomic mass of shielding material [ru
International Nuclear Information System (INIS)
Gel'fand, E.K.; Man'ko, B.V.; Serov, A.Ya.; Sychev, B.S.
1979-01-01
A complex of programs for modelling various radiation situations at high energy proton accelerators is considered. The programs are divided into there main groups according to their purposes. The first group includes programs for preparing constants describing the processes of different particle interaction with a substanc The second group of programs calculates the complete function of particle distribution arising in shields under irradiation by high energy nucleons. Concrete radiation situations arising at high energy proton accelerators are calculated by means of the programs of the third group. A list of programs as well as their short characteristic are given
Calculation of gamma-rays and fast neutrons fluxes with the program Mercure-4
International Nuclear Information System (INIS)
Baur, A.; Dupont, C.; Totth, B.
1978-01-01
The program MERCURE-4 evaluates gamma ray or fast neutron attenuation, through laminated or bulky three-dimensionnal shields. The method used is that of line of sight point attenuation kernel, the scattered rays being taken into account by means of build-up factors for γ and removal cross sections for fast neutrons. The integration of the point kernel over the range of sources distributed in space and energy, is performed by the Monte-Carlo method, with an automatic adjustment of the importance functions. Since it is operationnal the program MERCURE-4 has been intensively used for many various problems, for example: - the calculation of gamma heating in reactor cores, control rods and shielding screens, as well as in experimental devices and irradiation loops; - the evaluation of fast neutron fluxes and corresponding damage in structural materials of reactors (vessel steels...); - the estimation of gamma dose rates on nuclear instrumentation in the reactors, around the reactor circuits and around spent fuel shipping casks
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)
Shield nuclear design for the 5-kWe TE system
International Nuclear Information System (INIS)
Keshishian, V.
1972-01-01
The nuclear analysis of the 5-kW(e) reactor shield is presented. Calculation methods and optimization techniques used are presented. Borated stainless steel was selected for the gamma ray shield with tungsten alloy as an alternate. The total shield weight was calculated to be 355 lb. (U.S.)
International Nuclear Information System (INIS)
Charbonneau, B.W.; Killeen, P.G.; Carson, J.M.; Cameron, G.W.; Richardson, K.A.
1976-01-01
Results of airborne gamma-ray spectrometer surveys conducted by the Geological Survey of Canada are presented as maps contoured in units of radioelement and concentration ratios. These contoured values represent the average surface concentrations of the radioelements over areas of the order of several square kilometres. The relationship between this ''average surface concentration'' and the radioelement concentration in bedrock underlying the area depends on: (1) the percentage of outcrop; (2) the relation between overburden and bedrock radioelement concentration; (3) percentage of marshland or surface water in the area; (4) soil moisture; and (5) density of vegetation. More than 2500 portable gamma-ray spectrometer analyses of outcrop and overburden have been made in the Bancroft, Elliot Lake and Fort Smith areas of the Canadian Precambrian Shield. In the areas examined, the radioelement concentrations in glacial drift reflect the concentrations in the underlying bedrock. Rocks with near-crustal average contents of thorium, uranium and potassium are overlain by glacial drift having approximately the same concentrations. As the concentration in bedrock increases, the concentration in the local overburden also increases, but not to the same extent. In addition, in-situ gamma-ray spectrometry measurements were made at almost 1000 stations within the area of airborne surveys near Mont Laurier and Elliot Lake. These ground measurements were compared with the airborne measurements by averaging the values for all those ground stations located in the areas between each contour level on airborne maps. Radioelement concentrations in bedrock are considerably higher than corresponding airborne measurements, and this difference between bedrock and airborne values increases at higher radioelement concentrations. Radioelement concentrations in glacial drift are only slightly higher than airborne contour values for the same area. Airborne contour maps of the radioelement ratios
Application of a calculational model for thermal neutrons through biological shields
Energy Technology Data Exchange (ETDEWEB)
Hathout, A M [Nuclear engineering safety department, national center for nuclear safety and radiation, Nasr City Cairo, (Egypt)
1995-10-01
In this work a computational program, based on the Boltzmann transport integrodifferential equation, is applied. The scattering kernel is represented by the synthetic scattering model. The behaviour of thermal neutron in hydrogenous materials, which can be used as biological shields, are studied. These materials are water, polyethylene, Oak-Ridge concrete, ordinary concrete and manganese concrete. The data obtained are presented in tables. The results are analysed and compared with similar experimental values. Safety evaluation and environmental impact are discussed. 2 tabs.
Calculation of {beta}-ray spectra. Odd-odd nuclei
Energy Technology Data Exchange (ETDEWEB)
Tachibana, Takahiro [Waseda Univ., Tokyo (Japan). Advanced Research Center for Science and Engineering
1996-05-01
In order to study {beta}-ray of atomic nucleus, it is natural to consider {beta}-ray data fundamental and important. In a recent experiment, Rudstam measured {beta}-ray spectra from short term nuclear fission product species in 1990. It is an important check point in theoretical study on {beta}-ray to investigate if these experimental data can be reproduced by any theoretical calculation. As there are several spectrum studies of {beta}-ray through decay heat for its various properties due to the general theory of the {beta}-decay, little descriptions can be found. In even such studies, spectra under high excitation state of daughter species difficult to measure and apt to short experimental results were treated with combination spectra composed of experimental and calculated values such as substitution of a part of the general theory with calculated value. In this paper, the {beta} spectra supposed by only the general theory was reported without using such data combination in order to confirm effectiveness of the theory. In particular, this report was described mainly on the results using recent modification of odd-odd nucleus species. (G.K.)
Calculation of β-ray spectra. Odd-odd nuclei
International Nuclear Information System (INIS)
Tachibana, Takahiro
1996-01-01
In order to study β-ray of atomic nucleus, it is natural to consider β-ray data fundamental and important. In a recent experiment, Rudstam measured β-ray spectra from short term nuclear fission product species in 1990. It is an important check point in theoretical study on β-ray to investigate if these experimental data can be reproduced by any theoretical calculation. As there are several spectrum studies of β-ray through decay heat for its various properties due to the general theory of the β-decay, little descriptions can be found. In even such studies, spectra under high excitation state of daughter species difficult to measure and apt to short experimental results were treated with combination spectra composed of experimental and calculated values such as substitution of a part of the general theory with calculated value. In this paper, the β spectra supposed by only the general theory was reported without using such data combination in order to confirm effectiveness of the theory. In particular, this report was described mainly on the results using recent modification of odd-odd nucleus species. (G.K.)
Energy Technology Data Exchange (ETDEWEB)
Bourgeois, J; Lafore, P; Millot, J P; Rastoin, J; Vathaire, F de [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires
1959-07-01
1) The concept of an effective removal cross section has been developed in order more easily to compute reactor shielding thicknesses. We have built an experimental facility for the purpose of measuring effective removal cross sections, the value of which had not been published at that time. The first part of this paper describes the device or facility used, the computation method applied, and the results obtained. 2) Starting from this concept, we endeavoured to define a removal cross section as a function of energy. This enabled us to use the method for computations bearing on the attenuation of fast neutrons of any spectrum. An experimental verification was carried out for the case of fission neutrons filtered by a substantial thickness of graphite. 3) Finally, we outline a computation method enabling us to determine the sources of captured gamma rays by the age theory and we give an example of the application in a composite shield. (author) [French] 1) La notion de section efficace effective de deplacement a ete introduite pour calculer commodement les epaisseurs de protection des reacteurs. Nous avons construit un dispositif experimental destine a mesurer les sections efficaces effectives de deplacement dont la valeur n'avait pas ete publiee a cette epoque. La premiere partie de cette communication decrit le dispositif utilise, la methode de calcul employee et les resultats obtenus. 2) A partir de cette notion, nous avons essaye de definir une section efficace de deplacement fonction de l'energie. Ceci permet d'utiliser la methode du deplacement pour des calculs d'attenuation de neutrons rapides dont le spectre est quelconque. Une verification experimentale a ete faite dans le cas de neutrons de fission filtres par une epaisseur notable de graphite. 3) Enfin une methode de calcul permettant de determiner les sources de gamma de capture par la theorie de l'age est exposee et un exemple d'application donne dans une protection composite. (auteur)
Energy Technology Data Exchange (ETDEWEB)
Bourgeois, J; Lafore, P; Millot, J P; Rastoin, J; Vathaire, F de [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires
1958-07-01
1. The concept of an effective removal cross section has been developed in order more easily to compute reactor shielding thicknesses. We have built an experimental facility for the purpose of measuring effective removal cross sections, the value of which had not been published at that time. The first part of this paper describes the device or facility used, the computation method applied, and the results obtained. 2. Starting from this concept, we endeavored to define a removal cross section as a function of energy. This enabled us to use the method for computations bearing on the attenuation of fast neutrons of any spectrum. An experimental verification was carried out for the case of fission neutrons filtered by a substantial thickness of graphite. 3. Finally, we outline a computation method enabling us to determine the sources of captured gamma rays by the age theory and we give an example of the application in a composite shield. (author)Fren. [French] 1. La notion de section efficace effective de deplacement a ete introduite pour calculer commodement les epaisseurs de protection des reacteurs. Nous avons construit un dispositif experimental destine a mesurer les sections efficaces effectives de deplacement dont la valeur n'avait pas ete publiee a cette epoque. La premiere partie de cette communication decrit le dispositif utilise, la methode de calcul employee et les resultats obtenus. 2. A partir de cette notion, nous avons essaye de definir une section efficace de deplacement fonction de l'energie. Ceci permet d'utiliser la methode du deplacement pour des calculs d'attenuation de neutrons rapides dont le spectre est quelconque. Une verification experimentale a ete faite dans le cas de neutrons de fission filtres par une epaisseur notable de graphite. 3. Enfin une mde de calcul permettant de determiner les sources de gamma de capture par la theorie de l'age est exposee et un exemple d'application donne dans une protection composite. (auteur)
Calculation of Dose Gamma Ray Build up Factor in Some ...
African Journals Online (AJOL)
The gamma ray buildup factor was calculated by analyzing the narrow- beam and broad-beam geometry equations using Taylor's formula for isotropic sources and homogeneous materials. The buildup factor was programmed using MATLAB software to operate with any radiation energy (E), atomic number (Z) and the ...
Gamma ray shielding characteristic of BiZnBo-SLS and PbZnBo-SLS glass
Syuhada Ahmad, Nor; Shahrim Mustafa, Iskandar; Mansor, Ishak; Malik, Muhammad Fadhirul Izwan bin Abdul; Ain Nabilah Razali, Nur; Nordin, Sufiniza
2018-05-01
The radiation shielding and optical properties of x [RmOn] (0.5‑x) [ZnO] 0.2 [B2O3] 0.3 [SLS], where RmOn are Bi2O3 and PbO with x = 0.05, 0.10, 0.20, 0.30, 0.40, and 0.45 have been prepared by using the melt-quenching method at 1200 °C and was investigated on their physical, structural and gamma ray shielding properties. Field-emission scanning electron microscope (FESEM) data revealed that the particle morphologies is aggregated and irregular in shapes and size. Energy dispersive x-ray spectroscopy (EDS) elemental mapping data confirmed that all mentioned element all present on the prepared glass. Soda Lime Silica (SLS) that is mainly composed of SiO2 has been utilized in this study as the source of SiO2 for fabrication of glass system. From the result, the density and molar volume of both glass samples increased as Bi2O3 and PbO content increased. The gamma ray shielding properties, such as linear attenuation and mass attenuation coefficient, were increased while half value layer (HVL) and mean free path (MFP) were decreased as the increased in Bi2O3 and PbO concentrations. It is recognized that the mass attenuation coefficient value of Bi2O3 and PbO glass are slightly different. From this study, it can be concluded that from the non-toxicity and shielding point of view, the bismuth glass is a good shield to gamma radiation as compared to lead glass.
Ab initio theory and calculations of X-ray spectra
International Nuclear Information System (INIS)
Rehr, J.J.; Kas, J.J.; Prange, M.P.; Sorini, A.P.; Takimoto, Y.; Vila, F.
2009-01-01
There has been dramatic progress in recent years both in the calculation and interpretation of various x-ray spectroscopies. However, current theoretical calculations often use a number of simplified models to account for many-body effects, in lieu of first principles calculations. In an effort to overcome these limitations we describe in this article a number of recent advances in theory and in theoretical codes which offer the prospect of parameter free calculations that include the dominant many-body effects. These advances are based on ab initio calculations of the dielectric and vibrational response of a system. Calculations of the dielectric function over a broad spectrum yield system dependent self-energies and mean-free paths, as well as intrinsic losses due to multielectron excitations. Calculations of the dynamical matrix yield vibrational damping in terms of multiple-scattering Debye-Waller factors. Our ab initio methods for determining these many-body effects have led to new, improved, and broadly applicable x-ray and electron spectroscopy codes. (authors)
Gamma rays shielding and sensing application of some rare earth doped lead-alumino-phosphate glasses
Kaur, Preet; Singh, Devinder; Singh, Tejbir
2018-03-01
Seven rare earth (Sm3+, Eu3+ and Nd3+) doped lead alumino phosphate glasses were prepared. The protective and sensing measures from gamma rays were analysed in terms of parameters viz. density (ρ), refractive index, energy band gap (Eg), mean free path (mfp), effective atomic number (Zeff) and buildup factors (energy absorption EABF as well as exposure buildup factor EBF). The energy dependent parameters (mfp, Zeff, EABF and EBF) were investigated in the energy region from 15 keV to 15 MeV. EABF and EBF values were observed to be maximum in the intermediate energy region. Besides, the EABF and EBF values for the prepared samples are shown to have strong dependence on chemical composition of the glass at lower energy, whereas, it is almost independent of chemical composition in higher energy region. The prepared glass samples are found to have potential applications in radiation shielding as well as radiation sensing, which further find numerous applications in the field of medicine and industry.
Crystal glass and barite used for x ray and gamma radiation shielding
International Nuclear Information System (INIS)
Antonio Filho, Joao
2008-01-01
Full text: Crystal glass, barite plaster and barite concrete has been widely used as shielding material in gamma radiation sources as well as x-ray generating equipment to replace the plumbiferous glass and in the wall covering, in order to minimize exposure to individuals. However, properties of the radiation attenuation of crystal glass commercially available in Brazil, for the different types of energy are not known. For this reason, this work was carried out aiming to determine the radiation attenuation, transmission curves and Half Value Layer. In this work, ten plates of crystal glass, with dimensions of 20 cm x 20 cm and range of thicknesses from 0.5 to 2.0 cm, and ten plates of barite plaster and five plates of barite concrete, with dimensions of 20 x 20 cm 2 and range of thicknesses from 1,0 to 5,0 cm, were used. The plates were X-ray irradiated with potential constants of 60, 80, 110, 150 kV and gamma radiation of 60 Co. Analysis in the properties of the 60 Co radiation attenuation of barite plaster and barite concrete commercially available in Brazil were also carried out. The curves of attenuation and of transmission were obtained for crystal glass, barite plaster and barite concrete (mGy/m A.min) at 1 meter as a function of thickness. The thickness equivalent of a half value layer and deci value layer of crystal glass for all types of radiation and energies studied was also determined. Although their use permits the dimensioning of the armor covering for external x-radiation whit precision and safety without elevating the cost of protection. (author)
Calculation of neutron and gamma-ray flux-to-dose-rate conversion factors
International Nuclear Information System (INIS)
Kwon, S.G.; Lee, S.Y.; Yook, C.C.
1981-01-01
This paper presents flux-to-dose-rate conversion factors for neutrons and gamma rays based on the American National Standard Institute (ANSI) N666. These data are used to calculate the dose rate distribution of neutron and gamma ray in radiation fields. Neutron flux-to-dose-rate conversion factors for energies from 2.5 x 10 -8 to 20 MeV are presented; the corresponding energy range for gamma rays is 0.01 to 15 MeV. Flux-to-dose-rate conversion factors were calculated, under the assumption that radiation energy distribution has nonlinearity in the phantom, have different meaning from those values obtained by monoenergetic radiation. Especially, these values were determined with the cross section library. The flux-to-dose-rate conversion factors obtained in this work were in a good agreement to the values presented by ANSI. Those data will be useful for the radiation shielding analysis and the radiation dosimetry in the case of continuous energy distributions. (author)
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)
International Nuclear Information System (INIS)
Abd El-Latifa, A.A.; Ikraiam, F.A.; Abd El-Latifa, A.A.; Abd Elazziz, A.; Abd Elazziz, A.
2010-01-01
The mass attenuation coefficient , the build up factor , the half value thickness X 1/2 , and tenth value thickness X 1/10 of fiber concrete , 0% , 1% , 2%, 3%, and 4% by weight fiber content were measured at different gamma ray energies in MeV, 0.511,1.274 from Na-22 ,1.17 ,1.33 from Co-60 and 0.662 from Cs-137 . Appreciable variations were noted in the former nuclear parameters, due to the changes in the fiber content and gamma ray energies .A comparison of shielding properties of concrete with fiber content and reference sample(concrete without fiber ) have proven that the addition of steel fibers by weight to concrete have a potential application as a radiation shielding
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.
Dose Distribution Calculation Using MCNPX Code in the Gamma-ray Irradiation Cell
International Nuclear Information System (INIS)
Kim, Yong Ho
1991-02-01
60 Co-gamma irradiators have long been used for foods sterilization, plant mutation and development of radio-protective agents, radio-sensitizers and other purposes. The Applied Radiological Science Research Institute of Cheju National University has a multipurpose gamma irradiation facility loaded with a MDS Nordin standard 60 Co source (C188), of which the initial activity was 400 TBq (10,800 Ci) on February 19, 2004. This panoramic gamma irradiator is designed to irradiate in all directions various samples such as plants, cultured cells and mice to administer given radiation doses. In order to give accurate doses to irradiation samples, appropriate methods of evaluating, both by calculation and measurement, the radiation doses delivered to the samples should be set up. Computational models have been developed to evaluate the radiation dose distributions inside the irradiation chamber and the radiation doses delivered to typical biolological samples which are frequently irradiated in the facility. The computational models are based on using the MCNPX code. The horizontal and vertical dose distributions has been calculated inside the irradiation chamber and compared the calculated results with measured data obtained with radiation dosimeters to verify the computational models. The radiation dosimeters employed are a Famer's type ion chamber and MOSFET dosimeters. Radiation doses were calculated by computational models, which were delivered to cultured cell samples contained in test tubes and to a mouse fixed in a irradiation cage, and compared the calculated results with the measured data. The computation models are also tested to see if they can accurately simulate the case where a thick lead shield is placed between the source and detector. Three tally options of the MCNPX code, F4, F5 and F6, are alternately used to see which option produces optimum results. The computation models are also used to calculate gamma ray energy spectra of a BGO scintillator at
Hanni, Matti; Lantto, Perttu; Runeberg, Nino; Jokisaari, Jukka; Vaara, Juha
2004-09-22
Quantum chemical calculations of the nuclear shielding tensor, the nuclear quadrupole coupling tensor, and the spin-rotation tensor are reported for the Xe dimer using ab initio quantum chemical methods. The binary chemical shift delta, the anisotropy of the shielding tensor Delta sigma, the nuclear quadrupole coupling tensor component along the internuclear axis chi( parallel ), and the spin-rotation constant C( perpendicular ) are presented as a function of internuclear distance. The basis set superposition error is approximately corrected for by using the counterpoise correction (CP) method. Electron correlation effects are systematically studied via the Hartree-Fock, complete active space self-consistent field, second-order Møller-Plesset many-body perturbation, and coupled-cluster singles and doubles (CCSD) theories, the last one without and with noniterative triples, at the nonrelativistic all-electron level. We also report a high-quality theoretical interatomic potential for the Xe dimer, gained using the relativistic effective potential/core polarization potential scheme. These calculations used valence basis set of cc-pVQZ quality supplemented with a set of midbond functions. The second virial coefficient of Xe nuclear shielding, which is probably the experimentally best-characterized intermolecular interaction effect in nuclear magnetic resonance spectroscopy, is computed as a function of temperature, and compared to experiment and earlier theoretical results. The best results for the second virial coefficient, obtained using the CCSD(CP) binary chemical shift curve and either our best theoretical potential or the empirical potentials from the literature, are in good agreement with experiment. Zero-point vibrational corrections of delta, Delta sigma, chi (parallel), and C (perpendicular) in the nu=0, J=0 rovibrational ground state of the xenon dimer are also reported.
Calculation of Si(Li) x-ray detector efficiencies
International Nuclear Information System (INIS)
Zaluzec, N.; Holton, R.
1984-01-01
The calculation of detector efficiency functions is an important step in the quantitative analysis of x-ray spectra when approached by a standardless technique. In this regard, it becomes essential that the analyst not only model the physical aspects of the absorption and transmission of the various windows present, but also use the most accurate data available for the mass absorption coefficients required in these calculations. The topic of modeling the size and shape of the windows present is beyond the scope of this paper and the authors instead concentrate on the mass absorption coefficients used in the calculations and their implications to efficiency calculations. For the purposes of this paper, the authors consider that the relative detector efficiency function of a conventional Si(Li) detector can be modeled by a simple expression
Dose calculation of X-ray in medium
International Nuclear Information System (INIS)
Liu Yanmei; Xue Dingyu; Xu Xinhe; Chen Zhen; Dong Zaili
2006-01-01
The photon transportation in radiotherapy is studied based on Monte Carlo method. The dose calculation based on the MC simulation package DPM has been carried out, and the results have been visualized using MEX technology of Matlab. The dose results of X-ray in homogeneity and inhomogeneity medium have been compared with experimental data and those of other MC simulation package, and these results all agree. The calculation method we proposed has the advantage of high speed and good accuracy, therefore, is applicable in practice. (authors)
Calculation of gamma ray exposure rates from uranium ore bodies
International Nuclear Information System (INIS)
Thomson, J.E.; Wilson, O.J.
1980-02-01
The planning of operations associated with uranium mines often requires that estimates be made of the exposure rates from various ore bodies. A straight-forward method of calculating the exposure rate from an arbitrarily shaped body is presented. Parameters for the calculation are evaluated under the assumption of secular equilibrium of uranium with its daughters and that the uranium is uniformly distributed throughout an average soil mixture. The spectral distribution of the emitted gamma rays and the effect of air attenuation are discussed. Worked examples are given of typical situations encountered in uranium mines
Directory of Open Access Journals (Sweden)
Todd M. Alam
2002-08-01
Full Text Available Abstract: Ring formation in phosphate systems is expected to influence both the magnitude and orientation of the phosphorus (31P nuclear magnetic resonance (NMR chemical shielding anisotropy (CSA tensor. Ab initio calculations of the 31P CSA tensor in both cyclic and acyclic phosphate clusters were performed as a function of the number of phosphate tetrahedral in the system. The calculation of the 31P CSA tensors employed the GAUSSIAN 98 implementation of the gauge-including atomic orbital (GIAO method at the Hartree-Fock (HF level. It is shown that both the 31P CSA tensor anisotropy, and the isotropic chemical shielding can be used for the identification of cyclic phosphates. The differences between the 31P CSA tensor in acyclic and cyclic phosphate systems become less pronounced with increasing number of phosphate groups within the ring. The orientation of the principal components for the 31P CSA tensor shows some variation due to cyclization, most notably with the smaller, highly strained ring systems.
International Nuclear Information System (INIS)
Lawrence, R.D.; Dorning, J.J.
1980-01-01
A coarse-mesh discrete nodal integral transport theory method has been developed for the efficient numerical solution of multidimensional transport problems of interest in reactor physics and shielding applications. The method, which is the discrete transport theory analogue and logical extension of the nodal Green's function method previously developed for multidimensional neutron diffusion problems, utilizes the same transverse integration procedure to reduce the multidimensional equations to coupled one-dimensional equations. This is followed by the conversion of the differential equations to local, one-dimensional, in-node integral equations by integrating back along neutron flight paths. One-dimensional and two-dimensional transport theory test problems have been systematically studied to verify the superior computational efficiency of the new method
International Nuclear Information System (INIS)
Canetta, E.; Maino, G.; Menapace, E.
2001-01-01
The matter is reviewed, also following previous discussions at ICRS-9, concerning evaluation and related theoretical activities on nuclear data for radiation shielding within the framework of international co-operation initiatives, according to recognised needs and priorities. Both cross-section data.- for reactions induced by neutrons and photons - and nuclear structure data have been considered. In this context, main contributions and typical results are presented from theoretical and evaluation activities at the ENEA Applied Physics Division, especially concerning neutron induced reaction data up to 20 MeV and photonuclear reaction data such as photon absorption and (gamma,n) cross-sections. Relevant aspects of algebraic nuclear models and of evaporation and pre-equilibrium models are discussed. (authors)
Radiation calculations and shielding considerations for the design of the Next Linear Collider
International Nuclear Information System (INIS)
Nelson, W.R.; Rokni, S.H.; Vylet, V.
1996-11-01
The authors describe some of the work that they have done as a contribution to the Next Linear Collider (NLC) Zeroth-Order Design Report (ZDR), with specific emphasis placed on radiation-protection issues. However, because of the very nature of this machine--namely, extremely-small beam spots of high intensity--a new approach in accelerator radiation-protection philosophy appears to be warranted. Accordingly, the presentation will first take a look at recent design studies directed at protecting the machine itself, since this has resulted in a much better understanding of the very short exposure times involved whenever beam is lost and radiation sources are created. At the end of the paper, the authors suggest a Beam Containment System (BCS) that would provide an independent, redundant guarantee that exposure times are, indeed, kept very short. This, in turn, has guided them in the determination of the transverse shield thickness for the machine
International Nuclear Information System (INIS)
Isman MT; Elisabeth Supriatni; Tochrul Binowo
2002-01-01
Calculation of build up factor ordinary concrete, baryte concrete and blast-furnace sludge concrete have been carried out. The calculations have been carried out by dose rate measurement of Cs 137 source before and after passing through shielding. The investigated variables were concrete type, thickness of concrete and relative possession of concrete. Concrete type variables are ordinary concrete, baryte concrete and blast sludge furnace concrete. The thickness variables were 6, 12, 18, 24, 30 and 36 cm. The relative position variables were dose to the source and close to detector. The result showed that concrete type and position did not have significant effect to build-up factor value, while the concrete thickness (r) and the attenuation coefficient (μ) were influenced to the build-up factor. The higher μr value the higher build-up factor value. (author)
Energy Technology Data Exchange (ETDEWEB)
Palma, Daniel A. [CEFET QUIMICA de Nilopolis/RJ, Rio de Janeiro (Brazil); Goncalves, Alessandro C.; Martinez, Aquilino S.; Silva, Fernando C. [COPPE/UFRJ - Programa de Engenharia Nuclear, Rio de Janeiro (Brazil)
2008-07-01
The activation technique allows much more precise measurements of neutron intensity, relative or absolute. The technique requires the knowledge of the Doppler broadening function psi(x,xi) to determine the resonance self-shielding factors in the epithermal range G{sub epi} (tau,xi). Two new analytical approximations for the Doppler broadening function psi(x,xi) are proposed. The approximations proposed are compared with other methods found in literature for the calculation of the psi(x,xi) function, that is, the 4-pole Pade method and the Frobenius method, when applied to the calculation of G{sub epi} (tau,xi). The results obtained provided satisfactory accuracy. (authors)
Energy Technology Data Exchange (ETDEWEB)
Park, J; Lee, J [Program in Biomedical Radiation Sciences, Department of Transdisciplinary Studies, Graduate School of Convergence Science and Technology, Seoul National University, Seoul (Korea, Republic of); Institute of Radiation Medicine, Seoul National University Medical Research Center, Seoul (Korea, Republic of); Kim, H [Institute of Radiation Medicine, Seoul National University Medical Research Center, Seoul (Korea, Republic of); Interdisciplinary Program in Radiation Applied Life Science, Seoul National University College of Medicine, Seoul (Korea, Republic of); Kim, I [Department of Radiation Oncology, Seoul National University Hospital, Seoul (Korea, Republic of); Interdisciplinary Program in Radiation Applied Life Science, Seoul National University College of Medicine, Seoul (Korea, Republic of); Ye, S [Program in Biomedical Radiation Sciences, Department of Transdisciplinary Studies, Graduate School of Convergence Science and Technology, Seoul National University, Seoul (Korea, Republic of); Institute of Radiation Medicine, Seoul National University Medical Research Center, Seoul (Korea, Republic of); Department of Radiation Oncology, Seoul National University Hospital, Seoul (Korea, Republic of); Interdisciplinary Program in Radiation Applied Life Science, Seoul National University College of Medicine, Seoul (Korea, Republic of); Advanced Institutes of Convergence Technology, Seoul National University, Suwon (Korea, Republic of)
2015-06-15
Purpose: To evaluate the effect of a tungsten eye-shield on the dose distribution of a patient. Methods: A 3D scanner was used to extract the dimension and shape of a tungsten eye-shield in the STL format. Scanned data was transferred into a 3D printer. A dummy eye shield was then produced using bio-resin (3D systems, VisiJet M3 Proplast). For a patient with mucinous carcinoma, the planning CT was obtained with the dummy eye-shield placed on the patient’s right eye. Field shaping of 6 MeV was performed using a patient-specific cerrobend block on the 15 x 15 cm{sup 2} applicator. The gantry angle was 330° to cover the planning target volume near by the lens. EGS4/BEAMnrc was commissioned from our measurement data from a Varian 21EX. For the CT-based dose calculation using EGS4/DOSXYZnrc, the CT images were converted to a phantom file through the ctcreate program. The phantom file had the same resolution as the planning CT images. By assigning the CT numbers of the dummy eye-shield region to 17000, the real dose distributions below the tungsten eye-shield were calculated in EGS4/DOSXYZnrc. In the TPS, the CT number of the dummy eye-shield region was assigned to the maximum allowable CT number (3000). Results: As compared to the maximum dose, the MC dose on the right lens or below the eye shield area was less than 2%, while the corresponding RTP calculated dose was an unrealistic value of approximately 50%. Conclusion: Utilizing a 3D scanner and a 3D printer, a dummy eye-shield for electron treatment can be easily produced. The artifact-free CT images were successfully incorporated into the CT-based Monte Carlo simulations. The developed method was useful in predicting the realistic dose distributions around the lens blocked with the tungsten shield.
A novel removable shield attached to C-arm units against scattered X-rays from a patient's side
Energy Technology Data Exchange (ETDEWEB)
Mori, Hiroshige [Hokkaido Social Insurance Hospital, Department of Radiology, Sapporo, Hokkaido (Japan); Kanazawa University, Department of Quantum Medical Technology, Division of Health Sciences, Graduate School of Medical Science, Kanazawa, Ishikawa (Japan); Koshida, Kichiro; Matsubara, Kosuke [Kanazawa University, School of Health Sciences, College of Medical, Pharmaceutical and Health Sciences, Kanazawa, Ishikawa (Japan); Ishigamori, Osamu [Hokkaido Social Insurance Hospital, Department of Radiology, Sapporo, Hokkaido (Japan)
2014-08-15
We invented a drape-like shield against scattered X-rays that can safely come into contact with medical equipment or people during fluoroscopically guided procedures. The shield can be easily removed from a C-arm unit using one hand. We evaluated the use of the novel removable shield during the endoscopic retrograde cholangiopancreatography (ERCP) procedure. We measured the dose rate of scattered X-rays around endoscopists with and without this removable shield and surveyed the occupational doses to the ERCP staff. We also examined the endurance of the shield. The removable shield reduced the dose rate of scattered X-rays to one-tenth and reduced the monthly dose to an endoscopist by at least two-fifths. For 2.5 years, there was no damage to the shield and no loosening of the seam. The bonding of the hook-and-loop fasteners did not weaken, although the powerful double-sided tapes made especially for plastic did. The removable shield can reduce radiation exposure to the ERCP staff and may contribute to reducing the exposure to the eye lenses of operators. It would also be possible to expand its use to other fluoroscopically guided procedures besides ERCP because it is a light, simple, and useful device. (orig.)
Design and Fabrication of Radiation Shielded Micro X-Ray Diffraction System
International Nuclear Information System (INIS)
Park, Yang Soon; Han, Sun Ho; Ha, Kyeong Yeong; Jee, Kwang Yong
2006-12-01
It has been observed that microstructure changes occur at the radial edge of pellet(rim) of the fuel at a high burn-up and extended fuel cycle. The thickness of a rim is some hundreds of micrometers. Despite its narrow range, a rim would affect the behaviour of nuclear fuel. To determine lattice parameter with micro-XRD at intervals as small as 30 - 50 μm in radial direction of irradiated fuel samples, a radiation shielded micro-XRD system was designed and fabricated. This report describes the concept, shielding analysis, the structural design and the fabrication of a radiation shielded glove box for micro-XRD system. This radiation shielded micro-XRD system will be used for analysis of lattice parameter change and the phase distribution at intervals as small as 30 - 50 μm in radial direction of the rim of an irradiated fuel sample and a fuel cladding
Design and Fabrication of Radiation Shielded Micro X-Ray Diffraction System
Energy Technology Data Exchange (ETDEWEB)
Park, Yang Soon; Han, Sun Ho; Ha, Kyeong Yeong; Jee, Kwang Yong
2006-12-15
It has been observed that microstructure changes occur at the radial edge of pellet(rim) of the fuel at a high burn-up and extended fuel cycle. The thickness of a rim is some hundreds of micrometers. Despite its narrow range, a rim would affect the behaviour of nuclear fuel. To determine lattice parameter with micro-XRD at intervals as small as 30 - 50 {mu}m in radial direction of irradiated fuel samples, a radiation shielded micro-XRD system was designed and fabricated. This report describes the concept, shielding analysis, the structural design and the fabrication of a radiation shielded glove box for micro-XRD system. This radiation shielded micro-XRD system will be used for analysis of lattice parameter change and the phase distribution at intervals as small as 30 - 50 {mu}m in radial direction of the rim of an irradiated fuel sample and a fuel cladding.
Directory of Open Access Journals (Sweden)
Pin Gong
2018-04-01
Full Text Available With a highly functional methyl vinyl silicone rubber (VMQ matrix and filler materials of B4C, PbO, and benzophenone (BP and through powder surface modification, silicone rubber mixing, and vulcanized molding, a flexible radiation shielding and resistant composite was prepared in the study. The dispersion property of the powder in the matrix filler was improved by powder surface modification. BP was added into the matrix to enhance the radiation resistance performance of the composites. After irradiation, the tensile strength, elongation, and tear strength of the composites decreased, while the Shore hardness of the composites and the crosslinking density of the VMQ matrix increased. Moreover, the composites with BP showed better mechanical properties and smaller crosslinking density than those without BP after irradiation. The initial degradation temperatures of the composites containing BP before and after irradiation were 323.6°C and 335.3°C, respectively. The transmission of neutrons for a 2-mm thick sample was only 0.12 for an Am–Be neutron source. The transmission of γ-rays with energies of 0.662, 1.173, and 1.332 MeV for 2-cm thick samples were 0.7, 0.782, and 0.795, respectively. Keywords: Flexible Composite, Neutron Shielding, Radiation Resistance, γ-ray Shielding
Computer code ANISN multiplying media and shielding calculation II. Code description (input/output)
International Nuclear Information System (INIS)
Maiorino, J.R.
1990-01-01
The user manual of the ANISN computer code describing input and output subroutines is presented. ANISN code was developed to solve one-dimensional transport equation for neutron or gamma rays in slab, sphere or cylinder geometry with general anisotropic scattering. The solution technique is the discrete ordinate method. (M.C.K.)
International Nuclear Information System (INIS)
Li Hengxiang; Zhang Huafang; Chen Peizhong; Ma Aihua; Wang Yutao
2007-01-01
Objective: To develop a mobile shield for X-ray radiography in order to improve the protection of operators against the exposure in diagnosis of endemic diseases at site. Methods: Firstly, X-ray tube was routinely fixed on the stand while the operator was 1 and 2 m respectively away from the axis of it. When the shanks and forearms of the patient were exposed fractionally, the dose-equivalent rates of operator were detected at their forehead, chest and abdomen by pressurized ionization chamber. Secondly, the mobile shield was added before X-ray tube. The same parts were exposed and the dose-equivalent rates at the corresponding positions were detected as above. Fifty-five people were examined in areas with endemic diseases. The X-ray films were sham-exposed as control. The photographing quality was assessed. Results: In the routine way, the dose-equivalent rates for the forearms were 1043, 1191, and 1297 μSv/h at 1 m, and 512, 673, and 798 μSv/h at 2 m, while for the shanks were 1356, 1489, 1675 μSv/h at 1 m, and 743, 978 and 1127 μSv/h at 2 m, respectively. After the shield was added, the dose-equivalent rates for the forearms reduced to 1.7, 2.9 and 7.9 μSv/h at 1 m, and 0.6, 1.3 and 3.1 μSv/h at 2 m, while for the shanks were 2.2, 3.4, 9.1 μSv/h at 1 m, and 0.9, 1.2, and 3.9 μSv/h at 2 m respectively, which were 1/150-1/800 of the former. Moreover, the percentage of high-quality photos was increased by 30%. Conclusions: As a small device, the mobile shield for X-ray radiography is convenient for photograph of multi-parts, which helps to perform the X-ray test safely and effectively in areas of endemic diseases. (authors)
International Nuclear Information System (INIS)
Barranco R, F.
2015-01-01
In this thesis criticality and shielding calculations to evaluate the design of a container of dry storage of spent nuclear fuel generated in research reactors were made. The design of such container was originally proposed by Argentina and Brazil, and the Instituto Nacional de Investigaciones Nucleares (ININ) of Mexico. Additionally, it is proposed to modify the design of this container to store spent fuel 120 that are currently in the pool of TRIGA Mark III reactor, the Nuclear Center of Mexico and calculations and analyzes are made to verify that the settlement of these fuel elements is subcritical limits and dose rates to workers and the general public are not exceeded. These calculations are part of the design criteria for security protection systems in dry storage system (Dss for its acronym in English) proposed by the Nuclear Regulatory Commission (NRC) of the United States. To carry out these calculations simulation codes of Monte Carlo particle transport as MCNPX and MCNP5 were used. The initial design (design 1) 78 intended to store spent fuel with a maximum of 115. The ININ has 120 fuel elements and spent 3 control rods (currently stored in the reactor pool). This leads to the construction of two containers of the original design, but for economic reasons was decided to modify (design 2) to store in a single container. Criticality calculations are performed to 78, 115 and fresh fuel elements 124 within the container, to the two arrangements described in Chapter 4, modeling the three-dimensional geometry assuming normal operating conditions and accident. These calculations are focused to demonstrate that the container will remain subcritical, that is, that the effective multiplication factor is less than 1, in particular not greater than 0.95 (as per specified by the NRC). Spent fuel 78 and 124 within the container, both gamma radiation to neutron shielding calculations for only two cases were simulated. First actinides and fission products generated
International Nuclear Information System (INIS)
Hogenbirk, A.; Verschuur, K.A.
1993-09-01
The calculations were performed using both the Monte Carlo code MCNP and the 2D discrete ordinates neutron/γ transport code DORT. Consistent nuclear data were used: For the Monte Carlo calculations data were taken from the EFF-1.3 library, for the discrete ordinates calculations data were taken from the MAT175 library. Both libraries are based on the JEF/EFF-1 evaluation. Care was taken to model the 2.0 cm wide gaps between two blanket segments, as the neutron flux behind the vacuum vessel is largely determined by neutrons streaming through these gaps. The resulting neutron- and γ-flux spectra are in excellent agreement up to the end of the cryostat. It is noted, that at this position the attenuation of the neutron flux is about 11 orders of magnitude. Due to precautions in the Monte Carlo calculations the uncertainty in neutron- and γ-flux spectra calculated by MCNP is only small: The uncertainty in the integrated neutron spectrum amounts to approximately 15% at the end of the cryostat. Also the dose-rates as calculated by MCNP and DORT agree well. Differences occur when heating data are compared. This is clearly due to the different way in which nuclear heating is treated in MCNP (direct calculation of heating) and DORT (kerma factors used; including radioactive decay contributions). (orig.)
From user of photoreaction data. Application to shielding calculation of SR beamline
International Nuclear Information System (INIS)
Asano, Yoshihiro
1996-01-01
Photoneutron production-yield has been calculated by using the data of track-length distribution caused by gas bremsstrahlung and photoreaction cross-section on a insertion device beamline of SPring-8. The track-length distribution and gas bremsstrahlung spectra were calculated by the Monte Carlo code EGS4. About 1.7 μSv/h has been obtained for the total photoneutron dose-equivalent rate at 1 m from the center of a thick target of Pb by assuming that the target is a point source and neutron emission is isotropic. The benchmark calculations of photoneutron production presented at the SATIF2 meeting have been also discussed. (author)
International Nuclear Information System (INIS)
Baguena, A.; Shaw, M.; Williart, A.; Baguena, A.; Garcia, G.
2006-01-01
We describe the calculations and preliminary measures made for the installation of a X-ray generator tube. This device is going to be used for the secondary electron production from photonic primary radiation of up to 125 keV. With this experimental system, we will study the energetic and space distribution of produced secondary electrons by obtaining its spectrum of energies and its angular distribution. This method of measurement is going to be applied in different targets of radiological, environmental and biological interest. Calculations in the present article include: theoretical yield of X-rays production of the designed equipment, necessary shielding for the radiological safety of the installation staff, and an estimated dose due to their use. Characteristics of the installation and the equipment are described with this purpose. (author)
Energy Technology Data Exchange (ETDEWEB)
Baguena, A.; Shaw, M.; Williart, A. [Universidad Nacional de Educacion a Distancia, Dpto. Fisica de los Materiales, Madrid (Spain); Baguena, A. [Consejo de Seguridad Nuclear, Madrid (Spain); Garcia, G. [Instituto de Matematicas y Fisica Fundamental, Consejo Superior de Investigaciones Cientificas, Madrid (Spain)
2006-07-01
We describe the calculations and preliminary measures made for the installation of a X-ray generator tube. This device is going to be used for the secondary electron production from photonic primary radiation of up to 125 keV. With this experimental system, we will study the energetic and space distribution of produced secondary electrons by obtaining its spectrum of energies and its angular distribution. This method of measurement is going to be applied in different targets of radiological, environmental and biological interest. Calculations in the present article include: theoretical yield of X-rays production of the designed equipment, necessary shielding for the radiological safety of the installation staff, and an estimated dose due to their use. Characteristics of the installation and the equipment are described with this purpose. (author)
International Nuclear Information System (INIS)
Blanco, A; Gomez S
2012-01-01
After completing the rearrangement of the Spent Fuel Elements (SFE) into a compact arrangement in the two storage water pools, Atucha Nuclear Reactor 1 (ANR 1) will leave free position for the wet storage of the SFE discharged until December 2014. Even, in two possible scenarios, such as extending operation from 2015 or the cessation of operation after that date, it will be necessary to empty the interim storage water pools transferring the SFE to a temporary dry storage system. Because the law 25.018 'Management of Radioactive Wastes' implies for the first scenario - operation beyond 2015 - that Nucleoelectrica Argentina S.A. will still be in charge of the dry storage system and for the second - the cessation of operation after 2015 - the National Commission of Atomic Energy (CNEA) will be in charge by the National Management Program of Radioactive Wastes, the interim dry storage system of SNF is an issue of common interest which justifies go forward together. For that purpose and in accordance with the criticality and shielding calculations relevant to the project, in this paper we present the dose rate calculations for shielding conceptual design of a system for dry interim storage of the SFE of ANR 1. The specifications includes that the designed system must be suitable without modification for the SFE of the ANR 2. The results for the calculation of the photon dose rate, in touch and at one meter far, for the Transport Module and the Container of the SFE, are presented, which are required and controlled by the National Regulatory Authority (NRA) and the International Atomic Energy Agency (IAEA). It was used the SAS4 module of SCALE5.1 system and MCNP5. As a design tool for the photon shielding in order to meet current standards for allowable dose rates, a radial and axial parametric analysis were developed based on the thickness of lead of the Transport Module. The results were compared and verified between the two computing systems. Before this
International Nuclear Information System (INIS)
Vuong Huu Tan; Tran Tuan Anh; Nguyen Kien Cuong; Nguyen Canh Hai; Nguyen Xuan Hai; Pham Ngoc Son; Ho Huu Thang
2011-01-01
The tangential horizontal channel of No. 3 of the Dalat Research Reactor has been opened and used during the 1990s. The utilizations of the thermal neutron beam at this channel were the Neutron Radiography and the Prompt Gamma Neutron Activation Analysis method (PGNAA). At present, the neutron beam used for nuclear structure data researches based on the Summing of Amplitude Coincident Pulses system (SACP). Beside, several related research equipments have been set up and operated for the research purposes. A renovation of the neutron channel, therefore, will play an important role in safe and effective utilizations of the neutron beam in fields of nuclear physic training and researches. A new configuration for radiation shielding has been simulated by MCNP code. The calculated results of dose rates for neutron and gamma at working positions are in range of dose rate limit. (author)
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 comparison of the calculation methods of the maze shielding dose
International Nuclear Information System (INIS)
Li Wenqian; Li Junli; Li Pengyu; Tao Yinghua
2009-01-01
This paper gives a theoretical calculating method for the dose rate of the maze of the low-energy accelerators or high-energy accelerators, based on the NCRP report Nos.49, 51 and 151. The multi-legged maze of the Miyun CT workshop of the NUCTECH Company Limited and the arc maze of the radiation laboratory of the Academy of Military Medical Sciences were calculated using this method. The calculating results were compared with the MCNP simulating results and the measured results. For the commonly estimation of the maze dose rate, as long as the parameters chosen properly, this method can give a conservative result, and save time from simulation. It's hoped that this work could offer a reference for the maze design and the dose estimation method in the aftertime. (authors)
International Nuclear Information System (INIS)
Hong, Ser Gi; Kim, Kang-Seog
2011-01-01
This paper describes the iteration methods using resonance integral tables to estimate the effective resonance cross sections in heterogeneous transport lattice calculations. Basically, these methods have been devised to reduce an effort to convert resonance integral table into subgroup data to be used in the physical subgroup method. Since these methods do not use subgroup data but only use resonance integral tables directly, these methods do not include an error in converting resonance integral into subgroup data. The effective resonance cross sections are estimated iteratively for each resonance nuclide through the heterogeneous fixed source calculations for the whole problem domain to obtain the background cross sections. These methods have been implemented in the transport lattice code KARMA which uses the method of characteristics (MOC) to solve the transport equation. The computational results show that these iteration methods are quite promising in the practical transport lattice calculations.
Infantino, Angelo; Marengo, Mario; Baschetti, Serafina; Cicoria, Gianfranco; Longo Vaschetto, Vittorio; Lucconi, Giulia; Massucci, Piera; Vichi, Sara; Zagni, Federico; Mostacci, Domiziano
2015-11-01
Biomedical cyclotrons for production of Positron Emission Tomography (PET) radionuclides and radiotherapy with hadrons or ions are widely diffused and established in hospitals as well as in industrial facilities and research sites. Guidelines for site planning and installation, as well as for radiation protection assessment, are given in a number of international documents; however, these well-established guides typically offer analytic methods of calculation of both shielding and materials activation, in approximate or idealized geometry set up. The availability of Monte Carlo codes with accurate and up-to-date libraries for transport and interactions of neutrons and charged particles at energies below 250 MeV, together with the continuously increasing power of nowadays computers, makes systematic use of simulations with realistic geometries possible, yielding equipment and site specific evaluation of the source terms, shielding requirements and all quantities relevant to radiation protection. In this work, the well-known Monte Carlo code FLUKA was used to simulate two representative models of cyclotron for PET radionuclides production, including their targetry; and one type of proton therapy cyclotron including the energy selection system. Simulations yield estimates of various quantities of radiological interest, including the effective dose distribution around the equipment, the effective number of neutron produced per incident proton and the activation of target materials, the structure of the cyclotron, the energy degrader, the vault walls and the soil. The model was validated against experimental measurements and comparison with well-established reference data. Neutron ambient dose equivalent H*(10) was measured around a GE PETtrace cyclotron: an average ratio between experimental measurement and simulations of 0.99±0.07 was found. Saturation yield of 18F, produced by the well-known 18O(p,n)18F reaction, was calculated and compared with the IAEA recommended
International Nuclear Information System (INIS)
Silva, M
2006-01-01
The Atucha I Nuclear Power Plant (CNA-I) has enough room to store its spent fuel (SF) in damp in its two pool houses until the middle of 2015.Before that date there is the need to have an interim dry storage system for spent fuel that would make possible to empty at least one of the pools, whether to keep the plant operating if its useful life is extended, or to be able to empty the reactor core in case of decommissioning.Nucleolectrica Argentina S.A. (NA-SA) and the Comision Nacional de Energia Atomica (CNEA), due to their joint responsibility in the management of the SF, have proposed interim dry storage systems.These systems have to be evaluated in order to choose one of them by the end of 2006.In this work the Monte Carlo code MCNP was used to make the criticality and shielding calculations corresponding to the model proposed by CNEA.This model suggests the store of sealed containers with 36 or 37 SF in concrete modules.Each one of the containers is filled in the pool houses and transported to the module in a transference cask with lead walls.The results of the criticality calculations indicates that the solutions of SF proposed have widely fulfilled the requirements of subcriticality, even in supposed extreme accidental situations.Regarding the transference cask, the SF dose rate estimations allow us to make a feedback for the design aiming to the geometry and shielding improvements.Regarding the store modules, thicknesses ranges of concrete walls are suggested in order to fulfill the dose requirements stated by the Autoridad Regulatoria Nuclear Argentina [es
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)
Multiband method for resonance self-shielding calculation of fuel assembly in arbitrary geometries
International Nuclear Information System (INIS)
Huang Weibin; Wu Hongchun; Cao Liangzhi; Yang Weiyan
2009-01-01
A formula to calculate the multiband parameters is derived based on the multiband method. Adopting the method combining two-band and three-band, and based on the WIMSD4-69 library, a code named RESCAL is developed. The validation shows that the results of RESCAL code are well in accordance with MCNP's, and the numerical errors meet the practical requirement. Due to the limitation of WIMSD4 69-group library and the method adopted to calculate multiband parameters, the precision of RESCAL code is highly affected by the ratio of water to uranium. (authors)
Hagen, E.C.; Hudson, C.L.
1995-07-25
A new deflection structure which deflects a beam of charged particles, such as an electron beam, includes a serpentine set for transmitting a deflection field, and a shielding frame for housing the serpentine set. The serpentine set includes a vertical serpentine deflection element and a horizontal serpentine deflection element. These deflection elements are identical, and are interdigitatedly and orthogonally disposed relative to each other, for forming a central transmission passage, through which the electron beam passes, and is deflected by the deflection field, so as to minimize drift space signal distortion. The shielding frame includes a plurality of ground blocks, and forms an internal serpentine trough within these ground blocks, for housing the serpentine set. The deflection structure further includes a plurality of feedthrough connectors which are inserted through the shielding frame, and which are electrically connected to the serpentine set. 10 figs.
Energy Technology Data Exchange (ETDEWEB)
Lee, Yi-Kang, E-mail: yi-kang.lee@cea.fr
2016-11-01
Highlights: • Verification and validation of TRIPOLI-4 radiation transport calculations for ITER shielding benchmark. • Evaluation of CEA-V5.1.1 and FENDL-3.0 nuclear data libraries on D–T fusion neutron continuous energy transport calculations. • Advances in nuclear analyses for nuclear heating and radiation damage in iron. • This work also demonstrates that the “safety factors” concept is necessary in the nuclear analyses of ITER. - Abstract: With the growing interest in using the continuous-energy TRIPOLI-4{sup ®} Monte Carlo radiation transport code for ITER applications, a key issue that arises is whether or not the released TRIPOLI-4 code and its associated nuclear data libraries are verified and validated for the D–T fusion neutronics calculations. Previous published benchmark results of TRIPOLI-4 code on the ITER related activities have concentrated on the first wall loading, the reactor dosimetry, the nuclear heating, and the tritium breeding ratio. To enhance the TRIPOLI-4 verification and validation on neutron-gamma coupled calculations for fusion device application, the computational ITER shielding benchmark of M. E. Sawan was performed in this work by using the 2013 released TRIPOLI-4.9S code and the associated CEA-V5.1.1 data library. First wall, blanket, vacuum vessel and toroidal field magnet of the inboard and outboard components were fully modelled in this 1-D toroidal cylindrical benchmark. The 14.1 MeV source neutrons were sampled from a uniform isotropic distribution in the plasma zone. Nuclear responses including neutron and gamma fluxes, nuclear heating, and material damage indicator were benchmarked against previous published results. The capabilities of the TRIPOLI-4 code on the evaluation of above physics parameters were presented. The nuclear data library from the new FENDL-3.0 evaluation was also benchmarked against the CEA-V5.1.1 results for the neutron transport calculations. The results show that both data libraries
International Nuclear Information System (INIS)
Lee, Yi-Kang
2016-01-01
Highlights: • Verification and validation of TRIPOLI-4 radiation transport calculations for ITER shielding benchmark. • Evaluation of CEA-V5.1.1 and FENDL-3.0 nuclear data libraries on D–T fusion neutron continuous energy transport calculations. • Advances in nuclear analyses for nuclear heating and radiation damage in iron. • This work also demonstrates that the “safety factors” concept is necessary in the nuclear analyses of ITER. - Abstract: With the growing interest in using the continuous-energy TRIPOLI-4 ® Monte Carlo radiation transport code for ITER applications, a key issue that arises is whether or not the released TRIPOLI-4 code and its associated nuclear data libraries are verified and validated for the D–T fusion neutronics calculations. Previous published benchmark results of TRIPOLI-4 code on the ITER related activities have concentrated on the first wall loading, the reactor dosimetry, the nuclear heating, and the tritium breeding ratio. To enhance the TRIPOLI-4 verification and validation on neutron-gamma coupled calculations for fusion device application, the computational ITER shielding benchmark of M. E. Sawan was performed in this work by using the 2013 released TRIPOLI-4.9S code and the associated CEA-V5.1.1 data library. First wall, blanket, vacuum vessel and toroidal field magnet of the inboard and outboard components were fully modelled in this 1-D toroidal cylindrical benchmark. The 14.1 MeV source neutrons were sampled from a uniform isotropic distribution in the plasma zone. Nuclear responses including neutron and gamma fluxes, nuclear heating, and material damage indicator were benchmarked against previous published results. The capabilities of the TRIPOLI-4 code on the evaluation of above physics parameters were presented. The nuclear data library from the new FENDL-3.0 evaluation was also benchmarked against the CEA-V5.1.1 results for the neutron transport calculations. The results show that both data libraries can be
International Nuclear Information System (INIS)
Moon, S.H.; Ha, C.W.; Kwon, S.K.; Lee, J.K.; Choi, H.S.
1982-01-01
The theoretical bases of radiation streaming analysis in power reactors, such as ducts or reactor cavity, have been investigated. Discrete ordinates-Monte Carlo or Monte Carlo-Monte Carlo coupling techniques are suggested for the streaming analysis of ducts or reactor cavity. Single albedo scattering approximation code (SINALB) has been developed for simple and quick estimation of gamma-ray ceiling scattering, where the ceiling is assumed to be semi-infinite medium. This code has been employed to calculate the gamma-ray ceiling scattering effects in the laboratory containing a Co-60 source. The SINALB is applicable to gamma-ray scattering, only where the ceiling is thicker than Σsup(-1) and the height is at least twice higher than the shield wall. This code can be used for the purpose of preliminary radiation shield design. The MORSE code has been improved to analyze the gamma-ray scattering problem with on approximation method in respect to the random walk and estimation processes. This improved MORSE code has been employed to the gamma-ray ceiling scattering problem. The results of the improved MORSE calculation are in good agreement with the SINALB and standard MORSE. (Author)
Monte-Carlo calculation of irradiation dose content beyond shielding of high-energy accelerators
International Nuclear Information System (INIS)
Mokhov, N.V.; Frolov, V.V.
1975-01-01
The MARS programme, designed for calculating the three-dimensional internuclear cascade in defence of the accelerators by the Monte Carlo method, is described. The methods used to reduce the dispersion and the system of semi-empirical formulas made it possible to exceed the parameters of the existing programmes. By means of a synthesis of the results, registered by MARS and HAMLET programmes, the dosage fields for homogeneous and heterogeneous defence were evaluated. The results of the calculated absorbed and equivalent dose behind the barrier, irradiated by a proton beam, having the energy of Esub(o)=1/1000 GeV are exposed. The dependence of the high- and low-energy neutron, proton, pion, kaon, muonium and γ-quantum dosage on the initial energy and thickness, on the material and the composition of the defence is investigated
International Nuclear Information System (INIS)
Yamaji, Akio; Saito, Tetsuo.
1988-01-01
To investigate a proximity effect of ducts on shield performance against γ radiation, an experiment was performed at JRR-4 by entering the γ-ray beam into a concrete shield wall of 100 cm-thickness with 3 or 5 straight cylindrical ducts of radius of 4.45 cm placed in a straight line or crosswise at interval of 8.9 cm. The dose rates were measured using digital dosimeters on a horizontal line 20 cm apart from the rear of the wall with 0, 1, 3 and 5 ducts, and with the incident angles of 0deg, 7deg, 14deg and 20deg, respectively. The dose rate distributions depended on the number of ducts and the incident angle, and the dose rate ratios of with-three-ducts to no-duct distributed within 3.6∼12, 1.3∼5.0 and 1.1∼4.3, for the incident angles of 7deg, 14deg and 20deg, while those of with-single-duct to no-duct within 1.2∼7.1, 1.1∼2.7 and 1.0∼1.9, respectively. The experiment was analyzed using a multigroup single scattering code G33YSN able to deal with the geometry of the ducts exactly. For each incident angle, the calculation agreed with the experiment within a factor of 2. (author)
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.)
International Nuclear Information System (INIS)
Archer, Benjamin R.; Gray, Joel E.
2005-01-01
The recently published Report No. 147 of The National Council on Radiation Protection and Measurements entitled 'Structural shielding design for medical x-ray imaging facilities' provides an update of shielding recommendations for x rays used for medical imaging. The goal of this report is to ensure that the shielding in these facilities limits radiation exposures to employees and members of the public to acceptable levels. Board certified medical and health physicists, as defined in this report, are the 'qualified experts' who are competent to design radiation shielding for these facilities. As such, physicists must be aware of the new technical information and the changes from previous reports that Report No. 147 supersedes. In this article we summarize the new data, models and recommendations for the design of radiation barriers in medical imaging facilities that are presented in Report No. 147
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
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.
REMTWO - WRS system module number 299 for calculating the removal flux in a slab shield
International Nuclear Information System (INIS)
Grimstone, M.J.
1978-06-01
The WRS Modular Programming System has been developed as a means by which programmes may be more efficiently constructed, maintained and modified. In this system a module is a self-contained unit typically composed of one or more Fortran routines, and a programme is constructed from a number of such modules. This report describes one WRS module, the function of which is to calculate the uncollided flux due to a distributed source in a one-dimensional plane geometry system. A method is used in which the flux is approximated by the sum of diffusion equation solutions. The information given in this manual is of use both to the programmer wishing to incorporate the module in a programme, and to the user of such a programme. (author)
International Nuclear Information System (INIS)
Caldwell, J.
1984-01-01
Martinelli and Morini have used an analytical method for calculating values and distribution of the magnetic field in superconducting magnets. Using Fourier series the magnetic field is determined by carrying out a series expansion of the current density distribution of the system of coils. This Fourier method can be modified to include axial iron to a far greater accuracy (for finite permeability) by incorporating the image series approach of Caldwell and Zisserman. Also an exact solution can be obtained for the case of infinite permeability. A comparison of the results derived from the expansion of Martinelli and Morini with the exact solution of Caldwell and Zisserman shows excellent agreement for the iron-free case but the accuracy deteriorates as the permeability μ/sub z/ increases. The exact solution should be used for infinite permeability and also gives satisfactory results for permeability μ/sub z/ >100. A symmetric geometry is used throughout the communication for simplicity of presentation
International Nuclear Information System (INIS)
Broadhead, B.L.; Brady, M.C.; Parks, C.V.
1990-11-01
In 1985, the Nuclear Energy Agency-Committee on Reactor Physics (NEACRP) established a working group on shielding assessment of transportation packages. Following the initial distribution of a set of six problems, discussions were held at the Organization for Economic Cooperation and Development (OECD) Headquarters in Paris, France, in June/July 1986, May 1988, and February/March 1990. The US contribution to the working group is documented in this report. The results from this effort permit the evaluation of a number of approximations and effects that must be considered in a typical shielding analysis of a transportation cask. Among the effects reported here are the performance of multiple cross-section sets, the comparison of several source generation codes, and multidimensional versus one-dimensional (1-D) analyses. 18 refs., 16 figs., 33 tabs
International Nuclear Information System (INIS)
Santos, Maira R.; Silveira, Thiago B.; Garcia, Paulo L.; Trindade, Cassia; Martins, Lais P.; Batista, Delano V.S.
2013-01-01
Given the new methodology introduced in the shielding calculation due to recent modulated techniques in radiotherapy treatment, it became necessary to evaluate the impact of changes in the accelerator routine using such techniques. Based on a group of 30 patients from the National Cancer Institute (INCA) the workload multiplier factors for intensity modulated radiotherapy (IMRT factor) and for RapidArc™ (RA factor) were established. Four different routines in a 6 MV generic accelerator were proposed to estimate the impact of these modified workloads in the building cost of the secondary barriers. The results indicate that if 50% of patients are treating with IMRT, the secondary barrier becomes 14,1% more expensive than the barrier calculated for conformal treatments exclusive. While RA, in the same proportion, leads to a barrier only 3,7% more expensive. Showing that RA can, while reducing treatment time, increase the proportion of patients treated with modulation technique, without increasing the cost of the barrier, when compared with IMRT. (author)
International Nuclear Information System (INIS)
Reeves, J.H.; Hensley, W.K.; Brodzinski, R.L.
1984-10-01
Our current paper describes the development of a low cost shielding system using liquid scintillator for the cosmic ray detector-neutron moderator which accounts for a tenfold reduction in the cosmic continuum. Our primary objective was to develop a low cost anticoincidence shield for laboratory use which would substantially reduce the background from cosmic ray interactions. The minimum thickness of scintillator which would provide the necessary moderation of neutrons as well as furnish detectable quantities of light generated from cosmic ray interactions was determined experimentally. Tanks holding the liquid scintillator were constructed from stainless steel and were partitioned in such a manner that 10, 20, 30, or 40 cm thicknesses could be selected for background measurements. Lucite was used for construction of a tank which would allow the comparison of light output relative to stainless steel for a 10 cm thickness of liquid scintillator. Plastic scintillator was used for the bottom layer in all cases, however, liquid scintillator could be used with proper internal support. A 20 cm x 20 cm x 40 cm plastic scintillator was machined to completely surround the detector and fit inside 15 cm thick walls of lead which in turn, fit inside the stainless steel scintilllator tanks. Background measurements were taken with this inner scintillator both active and inactive. Measurements were also made using copper as well as iron as replacements for the inner scintillator
P(VDF-TrFE)/ZrO{sub 2} polymer-composites for X-ray shielding
Energy Technology Data Exchange (ETDEWEB)
Fontainha, Crissia Carem Paiva [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear; Baptista Neto, Annibal Theotonio; Santos, Adelina Pinheiro; Faria, Luiz Oliveira de, E-mail: farialo@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)
2016-03-15
Poly(vinylidene fluoride - tryfluorethylene) [P(VDF-TrFE)] copolymers were mixed with zirconia nanoparticles. The investigation was conducted with the intention to produce nanocompounds with potential to be used as protective patient shielding in radiological procedures. Polymer based nanocomposites with 1, 2, 3, 5 and 10 wt% of ZrO{sub 2} nanoparticles were prepared using sol-gel route with zirconium butoxide as the precursor for zirconium oxide nanoclusters. UV-Vis and FTIR spectrometry and differential scanning calorimetry (DSC) were used to characterize the composite samples. We observed a more homogeneous distribution of ZrO{sub 2} nanoparticles encapsulated by methyl methacrylate (MMA) into the polymeric matrix, when compared to composites made without the use of surface modifiers from methacrylate group. Apparently, this property is related to the absence of the strong MMA absorption band at 1745 cm{sup -1}, attributed to C=O bond, in the P(VDF-TrFE)/ZrO{sub 2} -MMA nanocomposites. The radiation damage due to high dose exposure was performed for gamma doses ranging from 100 kGy to 1,000 kGy. The radiation shielding characterization conducted using x-rays with effective energy of 40 keV has demonstrated that composites with 10% of ZrO{sub 2}, and only 1.0 mm thick, can attenuate 60% of the x-rays beam. (author)
Radiation Shielding Analyses of A 10 MeV, 15kW LINAC for Electron Beam and X-ray at KACST
Energy Technology Data Exchange (ETDEWEB)
Kang, W. G.; Pyo, S. H.; Han, B. S.; Kang, C. M. [EB Tech Co., Daejeon (Korea, Republic of); Alkhuraiji, T. S. [King AbdulAziz City for Science and Technology, Riyadh (Saudi Arabia)
2016-10-15
The King AbdulAziz City for Science and Technology (KACST) in the Kingdom of Saudi Arabia has a plan to build a 10 MeV, 15kW linear accelerator (LINAC) for electron beam and X-ray, which is to be supplied by EB Tech in Republic of Korea. The design and construction of the accelerator building will be carried out jointly between EB Tech and KACST. Recommendations for the design and installation of radiation shielding for x-ray and gamma-ray can be found in NCRP No. 49(1976) and for accelerators with energies over 10 MeV in NCRP No. 151 (2005). Monte Carlo calculations were conducted using the MCNP6 code to determine photon fluxes and doses at the point detectors locations around the accelerator building. The problem was run as an electron, photon and neutron transport problem to account for all reactions including the (γ,n) reaction. The detectors where the DXTRAN spheres were used are indicated in the table. The computation was continued until electrons reached a total of 1x10{sup +8} histories.
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.)
Energy Technology Data Exchange (ETDEWEB)
Kaur, Kulwinder; Singh, K.J., E-mail: kanwarjitsingh@yahoo.com; Anand, Vikas
2015-04-15
Highlights: • Transparent glass samples of the system 55PbO{sub x}BaO(45 − x)P{sub 2}O{sub 5} (x = 1 up to 5) have been prepared in the laboratory. • Gamma ray shielding properties improve with the addition of BaO. • Number of non-bridging oxygens decrease with the increase in the content of BaO. • Investigated glass system can be potential candidate as an alternate to conventional radiation shielding ‘concrete’. - Abstract: The presented work has been undertaken to evaluate the applicability of BaO doped PbO-P{sub 2}O{sub 5} glass system as gamma ray shielding material in terms of mass attenuation coefficient and half value layer at photon energies 662, 1173 and1332 keV. A meaningful comparison of their radiation shielding properties has been made in terms of their mass attenuation coefficient and HVL parameters with standard radiation shielding concrete ‘barite’. The density, molar volume, XRD, FTIR, Raman and UV–visible techniques and mechanical properties (by Yamane and Mackenzie's procedure) have been used to study the structural properties of the prepared glass system in order to check the possibility of their commercial utility as alternate to conventional concrete for gamma ray shielding applications.
International Nuclear Information System (INIS)
Marques, Andre Luis Ferreira; Ting, Daniel Kao Sun; Mendonca, Arlindo Gilson
1996-01-01
A calculation methodology of Flux Depression, Self-Shielding and Cadmium Factors is presented, using the ANISN code, for experiments conducted at the IPEN/MB-01 Research Reactor. The correction factors were determined considering thermal neutron flux and 0.125 e 0.250 mm diameter of 197 Au wires. (author)
International Nuclear Information System (INIS)
Kramer, S. L.; Ghosh, V. J.; Breitfeller, M.; Wahl, W.
2016-01-01
Third generation high brightness light sources are designed to have low emittance and high current beams, which contribute to higher beam loss rates that will be compensated by Top-Off injection. Shielding for these higher loss rates will be critical to protect the projected higher occupancy factors for the users. Top-Off injection requires a full energy injector, which will demand greater consideration of the potential abnormal beam miss-steering and localized losses that could occur. The high energy electron injection beam produces significantly higher neutron component dose to the experimental floor than a lower energy beam injection and ramped operations. Minimizing this dose will require adequate knowledge of where the miss-steered beam can occur and sufficient EM shielding close to the loss point, in order to attenuate the energy of the particles in the EM shower below the neutron production threshold (<10 MeV), which will spread the incident energy on the bulk shield walls and thereby the dose penetrating the shield walls. Designing supplemental shielding near the loss point using the analytic shielding model is shown to be inadequate because of its lack of geometry specification for the EM shower process. To predict the dose rates outside the tunnel requires detailed description of the geometry and materials that the beam losses will encounter inside the tunnel. Modern radiation shielding Monte-Carlo codes, like FLUKA, can handle this geometric description of the radiation transport process in sufficient detail, allowing accurate predictions of the dose rates expected and the ability to show weaknesses in the design before a high radiation incident occurs. The effort required to adequately define the accelerator geometry for these codes has been greatly reduced with the implementation of the graphical interface of FLAIR to FLUKA. This made the effective shielding process for NSLS-II quite accurate and reliable. Lastly, the principles used to provide
International Nuclear Information System (INIS)
Larcher, A.M.; Bonet Duran, S.M.
1998-01-01
Full text: Medical electron accelerators operating above 10 MeV produce radiation beams that are contaminated with neutrons. Therefore, shielding design for high energy accelerator rooms must consider the neutron component of the radiation field. In this paper a semiempirical method is presented to calculate doses due to neutrons and capture gamma rays inside the room and the maze. The calculation method is based on the knowledge of the neutron yield Q (neutrons/Gy of photons at isocenter) and the average energy of the primary beam of neutrons Eo (MeV). The method constitutes an appropriate tool for shielding facilities evaluation. The accuracy of the method has been contrasted with data obtained from the literature and an excellent correlation among the calculations and the measured values was achieved. In addition, the method has been used in the verification of experimental data corresponding to a 15 MeV linear accelerator installed in the country with similar results. (author) [es
International Nuclear Information System (INIS)
Kim, Do Hyun; Shin, Chang Ho; Kim, Song Hyun
2014-01-01
It uses the deterministic method to calculate adjoint fluxes for the decision of the parameters used in the variance reductions. This is called as hybrid Monte Carlo method. The CADIS method, however, has a limitation to reduce the stochastic errors of all responses. The Forward Weighted CADIS (FW-CADIS) was introduced to solve this problem. To reduce the overall stochastic errors of the responses, the forward flux is used. In the previous study, the Multi-Response CADIS (MR-CAIDS) method was derived for minimizing sum of each squared relative error. In this study, the characteristic of the MR-CADIS method was evaluated and compared with the FW-CADIS method. In this study, how the CADIS, FW-CADIS, and MR-CADIS methods are applied to optimize and decide the parameters used in the variance reduction techniques was analyzed. The MR-CADIS Method uses a technique that the sum of squared relative error in each tally region was minimized to achieve uniform uncertainty. To compare the simulation efficiency of the methods, a simple shielding problem was evaluated. Using FW-CADIS method, it was evaluated that the average of the relative errors was minimized; however, MR-CADIS method gives a lowest variance of the relative errors. Analysis shows that, MR-CADIS method can efficiently and uniformly reduce the relative error of the plural response problem than FW-CADIS method
International Nuclear Information System (INIS)
Nasrabadi, M.N.; Jalali, M.; Mohammadi, A.
2007-01-01
In this work thermal neutron self-shielding in aqueous bulk samples containing neutron absorbing materials is studied using bulk sample prompt gamma neutron activation analysis (BSPGNAA) with the MCNP code. The code was used to perform three dimensional simulations of a neutron source, neutron detector and sample of various material compositions. The MCNP model was validated against experimental measurements of the neutron flux performed using a BF 3 detector. Simulations were performed to predict thermal neutron self-shielding in aqueous bulk samples containing neutron absorbing solutes. In practice, the MCNP calculations are combined with experimental measurements of the relative thermal neutron flux over the sample's surface, with respect to a reference water sample, to derive the thermal neutron self-shielding within the sample. The proposed methodology can be used for the determination of the elemental concentration of unknown aqueous samples by BSPGNAA where knowledge of the average thermal neutron flux within the sample volume is required
Calculation of shielding parameters
International Nuclear Information System (INIS)
Montoya Z, J.
1994-01-01
With the propose of reduce the hazard to radiation, exist three basic factors: a) time, the time to exposition to working person inside to area, from exist determined speed the doses, is proportional of the time permanence; b) distance, the reduce to doses is inverse square of the distance to exposition point; c) building, consist to interpose between source and exposition point to material. The main aspect development to the analysis of parameters distance and building. The analysis consist to development of the mathematical implicit, in the model of source radioactive, beginning with the geometry to source, distance to exposition source, and configuration building. In the final part was realize one comparative studied to calculus of parameters to blinding, employs two codes CPBGAM and MICROSHIELD, the first made as part to work thesis. The point source its a good approximation to any one real source, but in the majority of the time to propose analysis the spatial distribution of the source must realized in explicit way. The buildings calculus in volumetry's source can be approximate begin's of plan as source adaptations. It's important to have present that not only the building exist the exposition to the radiation, and the parameters time and distance plays an important paper too. (Author)
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
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
Software for X-Ray Images Calculation of Hydrogen Compression Device in Megabar Pressure Range
Egorov, Nikolay; Bykov, Alexander; Pavlov, Valery
2007-06-01
Software for x-ray images simulation is described. The software is a part of x-ray method used for investigation of an equation of state of hydrogen in a megabar pressure range. A graphical interface that clearly and simply allows users to input data for x-ray image calculation: properties of the studied device, parameters of the x-ray radiation source, parameters of the x-ray radiation recorder, the experiment geometry; to represent the calculation results and efficiently transmit them to other software for processing. The calculation time is minimized. This makes it possible to perform calculations in a dialogue regime. The software is written in ``MATLAB'' system.
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)
International Nuclear Information System (INIS)
Bush, B.W.; Nix, J.R.
1993-01-01
The interaction of high-energy cosmic rays with nuclei in spacecraft shielding and the human body is important for manned interplanetary missions and is not well understood either experimentally or theoretically. We present a new theoretical approach to this problem based on classical hadrodynamics for extended nucleons, which treats nucleons of finite size interacting with massive meson fields. This theory represents the classical analogue of the quantum hadrodynamics of Serot and Walecka without the assumptions of the mean-field approximation and point nucleons. It provides a natural covariant microscopic approach to collisions between cosmic rays and nuclei that automatically includes space-time non-locality and retardation, nonequilibrium phenomena, interactions among all nucleons, and particle production. Unlike previous models, this approach is manifestly Lorentz covariant and satisfies a priori the basic conditions that are present when cosmic rays collide with nuclei, namely an interaction time that is extremely short and a nucleon mean-free path, force range, and internucleon separation that are all comparable in size. We review the history of classical meson-field theory and derive the classical relativistic equations of motion for nucleons of finite size interacting with massive scalar and vector meson fields
International Nuclear Information System (INIS)
Almeida, A. T. Jr.; Nogueira, M.S.; Santos, M.A.P.; Campos, L.L.; Araújo, F. G. S.
2015-01-01
The attenuation coefficient depends on the incident photon energy and the nature of the materials. In order to minimize exposure to individuals. Barite concrete has been largely used as a shielding material in installations housing gamma radiation sources as well as X-ray generating equipment. This study was conducted to evaluate the efficacy of different mixtures of barite concrete for shielding in diagnostic X-ray rooms. The mass attenuation coefficient (μ/ρ). The mass attenuation coefficients have been measured by employing the CdTe detector model XR-100T. The distance between the source and the exposed surface of all samples was measured by SSD light indicator of machine which was 350 cm. The slope of the linear plot of the intensity transmitted versus specimen thickness would yield the attenuation coefficient. The mass attenuation coefficients (μ/ρ) were compared with the tabulations based upon the results of the XCOM program. The rectangular barite concrete blocks in different thicknesses from were used for the radiation attenuation test. The experimental values were compared with theoretical values WinXcom. The plots of the logarithm of transmitted intensity versus specimen thickness were linear for all the samples and the µ/ρ was obtained from the plots by linear regression over the 25%-2% transmission range, under good geometrical condition. There is a good agreement between theoretical and experimental values, within the 9%. In fact over the entire transmission range of 25-2% the experimental and theoretical values agree well for both the energies. (authors)
Mahmoud, Mohamed E.; El-Khatib, Ahmed M.; Badawi, Mohamed S.; Rashad, Amal R.; El-Sharkawy, Rehab M.; Thabet, Abouzeid A.
2018-04-01
Polymer composites of high-density polyethylene (HD-PE) filled with powdered lead oxide nanoparticles (PbO NPs) and bulk lead oxide (PbO Blk) were prepared with filler weight fraction [10% and 50%]. These polymer composites were investigated for radiation-shielding of gamma-rays emitted from radioactive point sources [241Am, 133Ba, 137Cs, and 60Co]. The polymer was found to decrease the heaviness of the shielding material and increase the flexibility while the metal oxide fillers acted as principle radiation attenuators in the polymer composite. The prepared composites were characterized by Fourier transform infrared spectrophotometer (FT-IR), X-ray diffraction (XRD), thermogravimetric analysis (TGA), scanning electron microscope (SEM), Brunauer-Emmett-Teller surface area (BET) and field emission transmission electron microscope (FE-TEM). The morphological analysis of the assembled composites showed that, PbO NPs and PbO Blk materials exhibited homogenous dispersion in the polymer-matrix. Thermogravimetric analysis (TGA) demonstrated that the thermal-stability of HD-PE was enhanced in the presence of both PbO Blk and PbO NPs. The results declared that, the density of polymer composites was increase with the percentage of filler contents. The highest density value was identified as 1.652 g cm-3 for 50 wt% of PbO NPs. Linear attenuation coefficients (μ) have been estimated from the use of XCOM code and measured results. Reasonable agreement was attended between theoretical and experimental results. These composites were also found to display excellent percentage of heaviness with respect to other conventional materials.
Energy Technology Data Exchange (ETDEWEB)
Toreti, Dalila; Xavier, Clarice; Moura, Fabio, E-mail: clarice.xavier@rem.ind.b, E-mail: fabio.moura@rem.ind.b [REM Industria e Comercio Ltda., Sao Paulo, SP (Brazil)
2011-10-26
The CyberKnife robotic system uses a manipulator with six grade of freedom for positioning a 6 MV Linac accelerator for treatment of lesions. This paper presents calculations for a standard room, with 200 cm of thickness walls primary, build for a CyberKnife system, and calculations for a room originally designed for a Linac conventional (with gantry), with secondary barriers of 107 cm thickness. After the realization of shielding for both rooms, the results shown that walls of standard room with 200 cm thickness are adequate for the secondary shield, and for a room with a conventional Linac, from all six evaluated points, two would require additional shielding of nine cm and four cm of concrete with 2.4 g/cubic cm. This shows that the CyberKnife system can be installed in a originally designed room for a conventional Linac with neither restrict nor any shielding, since no incidence of beams on the secondary barriers is existent
Studying the effect of nano lead compounds additives on the concrete shielding properties for γ-rays
International Nuclear Information System (INIS)
Hassan, H.E.; Badran, H.M.; Aydarous, A.; Sharshar, T.
2015-01-01
In the present work the effect of concrete incorporation with two types of nano-lead compounds on its γ-ray shielding characteristics is investigated. The concrete samples were prepared according to the local standards of building materials and doped by different percentages of PbO and PbTiO_3 nano powders which were prepared using co-precipitation and oxalate precursor techniques, respectively. In addition, commercial PbO_2 powder additive was used to check the effect of particle size on concrete attenuation properties. The phase composition and particle size of all the lead-oxide additives were confirmed by XRD and TEM imaging. The γ-rays attenuation coefficients were measured as a function of the additive percentage of lead compounds for γ-ray energies of 662, 1173 and 1332 keV using "1"3"7Cs and "6"0Co sources. The microstructure changes occurred in the concrete samples doped with Pb compounds additives were probed using the positron annihilation spectroscopy (PAS) and the results were compared with that for normal concrete. The obtained data revealed that the overall defect density of the investigated samples, as seen by the positrons, decreases with increasing the nano-PbO contents which is in agreement with the determined values of the samples apparent densities. It was found that the γ-ray attenuation coefficient of concrete doped by nano-PbO is improved. The results are explained in the view of the fine structure enhanced modification and its impact on the γ-ray interaction probability at different energies.
International Nuclear Information System (INIS)
Spalding, J.F.; Archuleta, R.F.; London, J.E.; Prine, J.R.
1977-02-01
Two mouse strains (RF/J and C57B1/6J) were exposed to x-ray doses totaling 400, 800, or 1200 rad. Total doses were given in 200-rad fractions at 7-day intervals to the whole body, gut only, or gut shielded. Animals treated as above (conditioned) were divided into 2 groups to form a two-part investigation. X-ray-conditioned and control mice were subjected to a continuous gamma-ray stress (challenge exposure) 28 days after the last x-ray dose. Delayed injury was measured as a reduction in mean after-survival (MAS) time and was observed in whole-body, gut-conditioned, and gut-shielded groups. The cause of death was attributed to hemopoietic hypoplasia in all groups. MAS reduction in all conditioned groups in both strains was linear with dose within the dose range used. Delayed injury per volume dose (measured as a reduction in MAS) was independent of the tissue initially conditioned with an acute dose of x rays. Thus, delayed injury per unit weight of gut tissue exposed was equal to that of either whole-body or gut-shielded radiation injury. Comparative weight loss observations during the continuous gamma-ray challenge exposure revealed a decrement in metabolic processes associated with body weight maintenance. This decrement was seen in all x-ray-conditioned groups
Design and applications of an anticoincidence shielded low background gamma-ray spectrometer
Energy Technology Data Exchange (ETDEWEB)
Petri, H [Forschungszentrum Juelich GmbH (Germany). Zentralabteilung fuer Chemische Analysen
1997-03-01
A low background gamma-ray spectrometer has been constructed for measuring artificial and natural radioative isotopes. The design of the spectrometer, its properties and the application to the determination of natural radioactivity of dental ceramics are described. (orig.)
International Nuclear Information System (INIS)
Remetti, Romolo; Andreoli, Giulio; Keshishian, Silvina
2012-01-01
Highlights: ► We deal with HTR-10, that is a helium-cooled graphite-moderated pebble bed reactor. ► We carried out Monte Carlo simulation of the core by MCNP5. ► Extensive use of MCNP5 variance reduction methods has been done. ► We calculated the trend of neutron flux within the biological shield. ► We calculated neutron effective dose at the outer surface of biological shield. - Abstract: Research on experimental reactors, such as HTR-10, provide useful data about potentialities of very high temperature gas-cooled reactors (VHTR). The latter is today rated as one of the six nuclear reactor types involved in the Generation-IV International Forum (GIF) Initiative. In this study, the MCNP5 code has been employed to evaluate the neutron radiation trend vs. the biological shield's thickness and to calculate the neutron effective dose rate at the outer surface. The reactor's geometry has been completely modeled by means of lattices and universes provided by MCNP, even though some approximations were required. Monte Carlo calculations have been performed by means of a simple PC and, as a consequence, in order to obtain acceptable run times, it was made an extensive recourse to variance reduction methods.
DEFF Research Database (Denmark)
Kupka, Teobald; Stachów, Michal; Kaminsky, Jakub
2013-01-01
, estimated from calculations with the family of polarizationconsistent pcS-n basis sets is reported. This dependence was also supported by inspection of profiles of deviation between CBS estimated nuclear shieldings and obtained with significantly smaller basis sets pcS-2 and aug-cc-pVTZ-J for the selected......A linear correlation between isotropic nuclear magnetic shielding constants for seven model molecules (CH2O, H2O, HF, F2, HCN, SiH4 and H2S) calculated with 37 methods (34 density functionals, RHF, MP2 and CCSD(T) ), with affordable pcS-2 basis set and corresponding complete basis set results...... set of 37 calculation methods. It was possible to formulate a practical approach of estimating the values of isotropic nuclear magnetic shielding constants at the CCSD(T)/CBS and MP2/CBS levels from affordable CCSD(T)/pcS-2, MP2/pcS-2 and DFT/CBS calculations with pcS-n basis sets. The proposed method...
Energy Technology Data Exchange (ETDEWEB)
Lava, Deise D.; Borges, Diogo da S.; Affonso, Renato R.W.; Guimaraes, Antonio C.F.; Moreira, Maria de L., E-mail: deise_dy@hotmail.com, E-mail: diogosb@outlook.com, E-mail: raoniwa@yahoo.com.br, E-mail: tony@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)
2014-07-01
This paper is prepared in order to address calculations of shielding to minimize the interaction of patients with ionizing radiation and / or personnel. The work includes the use of protection report Radiation in Dental Medicine (NCRP-145 or Radiation Protection in Dentistry), which establishes calculations and standards to be adopted to ensure safety to those who may be exposed to ionizing radiation in dental facilities, according to the dose limits established by CNEN-NN-3.1 standard published in September / 2011. The methodology comprises the use of computer language for processing data provided by that report, and a commercial application used for creating residential projects and decoration. The FORTRAN language was adopted as a method for application to a real case. The result is a programming capable of returning data related to the thickness of material, such as steel, lead, wood, glass, plaster, acrylic, acrylic and leaded glass, which can be used for effective shielding against single or continuous pulse beams. Several variables are used to calculate the thickness of the shield, as: number of films used in the week, film load, use factor, occupational factor, distance between the wall and the source, transmission factor, workload, area definition, beam intensity, intraoral and panoramic exam. Before the application of the methodology is made a validation of results with examples provided by NCRP-145. The calculations redone from the examples provide answers consistent with the report.
International Nuclear Information System (INIS)
Sinitsa, V.V.
1984-11-01
The author gives a scheme for the calculation of the self-shielding factors in the unresolved resonance region using the GRUCON applied program package. This package is especially created to be used in the conversion of evaluated neutron cross-section data, as available in existing data libraries, into multigroup microscopic constants. A detailed description of the formulae and algorithms used in the programs is given. Some typical examples of calculation are considered and the results are compared with those of other authors. The calculation accuracy is better than 2%
Shielding concerns at a spallation source
International Nuclear Information System (INIS)
Russell, G.J.; Robinson, H.; Legate, G.L.; Woods, R.
1989-01-01
Neutrons produced by 800-MeV proton reactions at the Los Alamos Neutron Scattering Center spallation neutron source cause a variety of challenging shielding problems. We identify several characteristics distinctly different from reactor shielding and compute the dose attenuation through an infinite slab/shield composed of iron (100 cm) and borated polyethylene (15 cm). Our calculations show that (for an incident spallation spectrum characteristic of neutrons leaking from a tungsten target at 90/degree/) the dose through the shield is a complex mixture of neutrons and gamma rays. High-energy (> 20 MeV) neutron production from the target is ≅5% of the total, yet causes ≅68% of the dose at the shield surface. Primary low-energy (< 20 MeV) neutrons from the target contribute negligibly (≅0.5%) to the dose at the shield surface yet cause gamma rays, which contribute ≅31% to the total dose at the shield surface. Low-energy neutrons from spallation reactions behave similarly to neutrons with a fission spectrum distribution. 6 refs., 8 figs., 1 tab
International Nuclear Information System (INIS)
M. Gross
2006-01-01
The overall objective of the work scope covered by this technical work plan (TWP) is to develop new damage abstractions for the seismic scenario class in total system performance assessment (TSPA). The new abstractions will be based on a new set of waste package and drip shield damage calculations in response to vibratory ground motion and fault displacement. The new damage calculations, which are collectively referred to as damage models in this TWP, are required to represent recent changes in waste form packaging and in the regulatory time frame. The new damage models also respond to comments from the Independent Validation Review Team (IVRT) postvalidation review of the draft TSPA model regarding performance of the drip shield and to an Additional Information Need (AIN) from the U.S. Nuclear Regulatory Commission (NRC)
Energy Technology Data Exchange (ETDEWEB)
M. Gross
2006-12-08
The overall objective of the work scope covered by this technical work plan (TWP) is to develop new damage abstractions for the seismic scenario class in total system performance assessment (TSPA). The new abstractions will be based on a new set of waste package and drip shield damage calculations in response to vibratory ground motion and fault displacement. The new damage calculations, which are collectively referred to as damage models in this TWP, are required to represent recent changes in waste form packaging and in the regulatory time frame. The new damage models also respond to comments from the Independent Validation Review Team (IVRT) postvalidation review of the draft TSPA model regarding performance of the drip shield and to an Additional Information Need (AIN) from the U.S. Nuclear Regulatory Commission (NRC).
Development of Monte Carlo decay gamma-ray transport calculation system
Energy Technology Data Exchange (ETDEWEB)
Sato, Satoshi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Kawasaki, Nobuo [Fujitsu Ltd., Tokyo (Japan); Kume, Etsuo [Japan Atomic Energy Research Inst., Center for Promotion of Computational Science and Engineering, Tokai, Ibaraki (Japan)
2001-06-01
In the DT fusion reactor, it is critical concern to evaluate the decay gamma-ray biological dose rates after the reactor shutdown exactly. In order to evaluate the decay gamma-ray biological dose rates exactly, three dimensional Monte Carlo decay gamma-ray transport calculation system have been developed by connecting the three dimensional Monte Carlo particle transport calculation code and the induced activity calculation code. The developed calculation system consists of the following four functions. (1) The operational neutron flux distribution is calculated by the three dimensional Monte Carlo particle transport calculation code. (2) The induced activities are calculated by the induced activity calculation code. (3) The decay gamma-ray source distribution is obtained from the induced activities. (4) The decay gamma-rays are generated by using the decay gamma-ray source distribution, and the decay gamma-ray transport calculation is conducted by the three dimensional Monte Carlo particle transport calculation code. In order to reduce the calculation time drastically, a biasing system for the decay gamma-ray source distribution has been developed, and the function is also included in the present system. In this paper, the outline and the detail of the system, and the execution example are reported. The evaluation for the effect of the biasing system is also reported. (author)
Spectra transmitted by mortar barite in X-Ray qualities applied in diagnostic radiology as shielding
Energy Technology Data Exchange (ETDEWEB)
Almeida Junior, Airton T., E-mail: airton.almeida@fundacentro.gov.br [Fundacaoo Jorge Duprat Figueiredo de Seguranca e Medicina do Trabalho (FUNDACENTRO), Belo Horizonte, MG (Brazil); Araujo, F.G.S. [Universidade Federal de Ouro Preto (REDEMAT/UFOP), MG (Brazil); Nogueira, M.S., E-mail: mnogue@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Santos, M.A.P. [Centro Regional de Ciencias Nucleares do Nordeste (CRCN-NE/CNEN-PE), Recife, PE (Brazil)
2015-07-01
Concrete which contains water, cement and aggregate, is widely used in building construction such as medical hospitals. The CdZnTe spectrometry system was used to acquire the transmitted spectra in the RQR qualities and the stripping procedure was performed by taking into account both the contributions of efficiency and x-ray escape fraction, experimentally determined. The samples were prepared in rectangular plate format with dimensions of (5 x 5) cm with thicknesses varying from 0.2cm to 2cm and exposed to x-ray beams generated. The HVL and the mean energy in this energy range was determined. (author)
Gonad shielding in diagnostic radiology
International Nuclear Information System (INIS)
Anon.
1975-01-01
The use of gonad shielding is an important radiation protection technique, intended to reduce unnecessary x-ray exposure of the gonads of patients from diagnostic x-ray procedures. The types of gonad shields in use are discussed as are the types of diagnostic examinations that should include gonad shielding. It was found that when properly used, most shields provided substantial gonad dose reductions
Monte Carlo shielding analyses using an automated biasing procedure
International Nuclear Information System (INIS)
Tang, J.S.; Hoffman, T.J.
1988-01-01
A systematic and automated approach for biasing Monte Carlo shielding calculations is described. In particular, adjoint fluxes from a one-dimensional discrete ordinates calculation are used to generate biasing parameters for a Monte Carlo calculation. The entire procedure of adjoint calculation, biasing parameters generation, and Monte Carlo calculation has been automated. The automated biasing procedure has been applied to several realistic deep-penetration shipping cask problems. The results obtained for neutron and gamma-ray transport indicate that with the automated biasing procedure Monte Carlo shielding calculations of spent-fuel casks can be easily performed with minimum effort and that accurate results can be obtained at reasonable computing cost
International Nuclear Information System (INIS)
Yohannes Sardjono; Hamidatul Faqqiyyah; Niels Bassler
2014-01-01
The projection of world population growth and increased longevity are leading to a rapid increase in the total number of middle-aged and older adults, with a corresponding increase in the number of deaths caused by non communicable diseases. It is projected that the annual number of deaths due to cardiovascular disease will increase from 17 million in 2008 to 25 million in 2030 with annual cancer deaths increasing from 7.6 million to 13 million. Boron Neutron Capture Therapy is a therapy that utilizes the absorption interaction of Boron-10 with thermal neutron and become He-4 particle and located in cell target and very short half life gamma emission. Studies were carried out to dose distribution in HER-2+ breast cancer therapy by Boron Neutron Capture Therapy (BNCT) using SHIELD Heavy Ion Therapy (HIT12A) T program. The Monte Carlo particle transport code SHIELD-HIT1 is designed to precisely simulate therapeutic beams of protons and ions in biological tissue relevant for ion beam cancer therapy. SHIELD-HIT (Heavy Ion Therapy) evolved from the common SHIELD code that models interactions of hadrons and atomic nuclei in complex extended targets in the energy range up to 1 TeV/nucleon. Through this computer code can be applied to calculate of absorption dose in cell target. (author)
International Nuclear Information System (INIS)
Du Licheng; He Ping; Zhou Yuanlin; Song Kaiping; Yang Kuihua
2012-01-01
Lead tungstate combines the radiation shielding properties of tungsten and lead, and it is quite distinctive to manufacture lead tungstate with ultra-fine granularity to enhance its capacity of radiation shielding. The grain size of lead tungstate has direct impact on the ability of its protection from radioactive materials. the smaller the grain size and more uniform dispersion of lead tungstate, the better protective ability it is going to be. In this paper, soft-template synthesis was introduced to prepare ultra-fine PbWO 4 . Rigorous experiment conditions are settled to ensure the access to obtain ultra-fine, homogeneous lead tungstate product, and it is better than other physical and chemical preparation methods. The surface-active agent for the soft template, with S-60 for the water system W/O microemulsion zone, was used to synthesize successfully ultra-fine PbWO 4 . It was shown that dispersing agent S-60 in the soft template method produced ultra-fine PbWO 4 with uniform granularity distribution. By using orthogonal experimental method, the best experimental conditions were obtained as follows: S-60 as surfactant dispersant with diluted 30 times concentration, solutions with pH9, 0.01 mol/L concentration of reactant, 1300 rpm of stirring speed and slowly adding drops of Na 2 WO 4 solution into Pb (Ac) 2 solution. Based on the optimal experimental conditions, the product of ultra-fine product for the anti-radiation protection filler has been made. The fine packing for the preparation of tungsten the gamma rays on the gloves is an average capacity of 5% or so. (authors)