A study of gamma shielding

International Nuclear Information System (INIS)

Roogtanakait, N.

1981-01-01

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

Shielding analysis of high level waste water storage facilities using MCNP code

Energy Technology Data Exchange (ETDEWEB)

Yabuta, Naohiro [Mitsubishi Research Inst., Inc., Tokyo (Japan)

2001-01-01

The neutron and gamma-ray transport analysis for the facility as a reprocessing facility with large buildings having thick shielding was made. Radiation shielding analysis consists of a deep transmission calculation for the concrete wall and a skyshine calculation for the space out of the buildings. An efficient analysis with a short running time and high accuracy needs a variance reduction technique suitable for all the calculation regions and structures. In this report, the shielding analysis using MCNP and a discrete ordinate transport code is explained and the idea and procedure of decision of variance reduction parameter is completed. (J.P.N.)

Beta Bremsstrahlung dose in concrete shielding

Energy Technology Data Exchange (ETDEWEB)

Manjunatha, H.C., E-mail: manjunatha@rediffmail.com [Department of Physics, Government college for women, Kolar 563101, Karnataka (India); Chandrika, B.M. [Shravana, 592, Ist Cross, Behind St.Anne s School, PC Extension, Kolar 563101, Karnataka (India); Rudraswamy, B. [Department of Physics, Bangalore University, Bangalore 560056, Karnataka (India); Sankarshan, B.M. [Shravana, 592, Ist Cross, Behind St.Anne s School, PC Extension, Kolar 563101, Karnataka (India)

2012-05-11

In a nuclear reactor, beta nuclides are released during nuclear reactions. These betas interact with shielding concrete and produces external Bremsstrahlung (EB) radiation. To estimate Bremsstrahlung dose and shield efficiency in concrete, it is essential to know Bremsstrahlung distribution or spectra. The present work formulated a new method to evaluate the EB spectrum and hence Bremsstrahlung dose of beta nuclides ({sup 32}P, {sup 89}Sr, {sup 90}Sr-{sup 90}Y, {sup 90}Y, {sup 91}Y, {sup 208}Tl, {sup 210}Bi, {sup 234}Pa and {sup 40}K) in concrete. The Bremsstrahlung yield of these beta nuclides in concrete is also estimated. The Bremsstrahlung yield in concrete due to {sup 90}Sr-{sup 90}Y is higher than those of other given nuclides. This estimated spectrum is accurate because it is based on more accurate modified atomic number (Z{sub mod}) and Seltzer's data, where an electron-electron interaction is also included. Presented data in concrete provide a quick and convenient reference for radiation protection. The present methodology can be used to calculate the Bremsstrahlung dose in nuclear shielding materials. It can be quickly employed to give a first pass dose estimate prior to a more detailed experimental study. - Highlights: Black-Right-Pointing-Pointer Betas released in a nuclear reactor interact with shielding concrete and produces Bremsstrahlung. Black-Right-Pointing-Pointer The present work formulated a new method to evaluate the Bremsstrahlung spectrum and dose in concrete. Black-Right-Pointing-Pointer Presented data in concrete provide a quick and convenient reference for radiation protection.

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

Experimentation with a prototype incinerator for beta-gamma waste

International Nuclear Information System (INIS)

Farber, M.G.; Lewandowski, K.E.; Becker, G.W.

1982-01-01

A test facility for the incineration of suspect and low-level beta-gamma waste has been built and operated at the Savannah River Laboratory. The processing steps include waste feeding, incineration, ash residue packaging, and off-gas cleanup. Demonstration of the full-scale (180 kg/hr) facility with nonradioactive, simulated waste is currently in progress. At the present time, over nine metric tons of material including rubber, polyethylene, and cellulose have been incinerated during three burning campaigns. A comprehensive test program of solid and liquid waste incineration is being implemented. The data from the research program is providing the technical basis for a phase of testing with low-level beta-gamma waste generated at the Savannah River Plant

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

International Nuclear Information System (INIS)

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

2016-01-01

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

TRIGA out of core gamma irradiation facility

International Nuclear Information System (INIS)

Rant, J.; Pregl, G.

1988-01-01

A possibility to irradiate extended objects in a gamma field inside the shielding water tank and above the core of operating TRIGA Mark II Reactor has been investigated. The irradiation cask is shielded with Cd cover to filter out thermal neutrons. The dose rate of the gamma field strongly depends on the distance of the irradiation position above the core. At 25 cm above the core, the gamma dose rate is 2.2 Gy/s and epithermal neutron flux is ∼ 8.10 6 ncm -2 s -1 ∼ 3 as measured by TLD (CaF 2 : Mn) dosimeters and Au foils respectively. Tentative applications of the gamma irradiation facility are in the studies of radiation induced accelerated aging and within the Nuclear Power Plant Equipment Qualification Program (EQP). A complete characterization of the neutron spectrum and optimization of the 7 radiation field within the cask has still to be performed. (author)

Shielding of the GERDA experiment against external gamma background

International Nuclear Information System (INIS)

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

2009-01-01

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

Poster: The EURISOL Beta-beam facility

CERN Document Server

The beta-beam concept for the generation of an electron (anti-)neutrino beam was proposed by Piero Zucchelli (CERN) in 2002. A first study of the possibility of using the existing CERN machines for the acceleration for radioactive ions to a relativistic gamma of roughly 100, for later storage in a new decay ring of approximately the size of SPS, was made in 2002. The results from this very first short study were very encouraging.In 2004 it was decided to incorporate a design study for the beta-beam within the EURISOL DS proposal. EURISOL is a project name for a next-generation radioactive beam facility based on the ISOL method for the production of intense radioactive beams for nuclear physics, astrophysics and other applications. The proposal was accepted with the beta-beam task as an integral part. The design study officially started 1 February 2005 and will run for 4 years resulting in a conceptual design report for a beta-beam facility.

Simultaneous beta and gamma spectroscopy

Science.gov (United States)

Farsoni, Abdollah T.; Hamby, David M.

2010-03-23

A phoswich radiation detector for simultaneous spectroscopy of beta rays and gamma rays includes three scintillators with different decay time characteristics. Two of the three scintillators are used for beta detection and the third scintillator is used for gamma detection. A pulse induced by an interaction of radiation with the detector is digitally analyzed to classify the type of event as beta, gamma, or unknown. A pulse is classified as a beta event if the pulse originated from just the first scintillator alone or from just the first and the second scintillator. A pulse from just the third scintillator is recorded as gamma event. Other pulses are rejected as unknown events.

Beta-gamma contaminated solid waste incinerator facility

International Nuclear Information System (INIS)

Hootman, H.E.

1979-10-01

This technical data summary outlines a reference process to provide a 2-stage, 400 lb/hour incinerator to reduce the storage volume of combustible process waste contaminated with low-level beta-gamma emitters in response to DOE Manual 0511. This waste, amounting to more than 200,000 ft 3 per year, is presently buried in trenches in the burial ground. The anticipated storage volume reduction from incineration will be a factor of 20. The incinerator will also dispose of 150,000 gallons of degraded solvent from the chemical separations areas and 5000 gallons per year of miscellaneous nonradioactive solvents which are presently being drummed for storage

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.

External dosimetry sources and shielding

International Nuclear Information System (INIS)

Calisto, Washington

1994-01-01

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

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)

Selective Beta and Gamma-ray Discrimination by CdWO{sub 4} and PlasticScintillator

Energy Technology Data Exchange (ETDEWEB)

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

2017-04-15

Radiation monitoring technique has been used for monitoring of decommissioning site of nuclear facility, radioactive waste disposal site, or in case of radioactivity accident. For rapid measurement of gamma-ray and beta-ray, many portable radiation detectors were developed but they are sensitive to specific radiation type. For example, portable detectors using NaI(Tl) or high purity germanium (HPGe) are suitable to detect gamma-ray. Otherwise, Geiger-müller (GM) tube or ionization chamber are suitable to detect all-types of radiation but it is hard to determine which particle is detected in the detector. In this reason, phoswich detectors for discrimination of beta-ray and gamma-ray were developed by using pulse shape discrimination. In this study, another approach to discriminate the beta-ray and gamma-ray is carried out. Two scintillators are used, cadmium tungstate (CdWO{sub 4}) and plastic scintillator. They have huge difference in their effective atomic number and mass density, thus they have huge difference in their gamma-ray sensitivity while the sensitivity of beta-ray is similar. The characterization of beta-ray and gamma-ray discrimination by using this characteristics is include. A technique of discrimination between beta-ray and gamma-ray was suggested. The method was verified by Monte Carlo simulation and experiment. This work showed feasibility on in field measurement of radiation with discrimination of beta-ray and gamma-ray.

Beta induced Bremsstrahlung dose rate in concrete shielding

International Nuclear Information System (INIS)

Manjunatha, H.C.

2013-01-01

Dosimetric study of beta-induced Bremsstrahlung in concrete is importance in the field of radiation protection. The efficiency, intensity and dose rate of beta induced Bremsstrahlung by 113 pure beta nuclides in concrete shielding is computed. The Bremsstrahlung dosimetric parameters such as the efficiency (yield), Intensity and dose rate of Bremsstrahlung are low for 199 Au and high for 104 Tc in concrete. The efficiency, Intensity and dose rate of Bremsstrahlung increases with maximum energy of beta nuclide (Emax) and modified atomic number (Zmod) of the target. The estimated Bremsstrahlung efficiency, Intensity and dose rate are useful in the calculations photon track-length distributions. These parameters are useful to determine the quality and quantity of the radiation (known as the source term). Precise estimation of this source term is very important in planning of radiation shielding. (author)

Beta and Gamma Gradients

DEFF Research Database (Denmark)

Løvborg, Leif; Gaffney, C. F.; Clark, P. A.

1985-01-01

Experimental and/or theoretical estimates are presented concerning, (i) attenuation within the sample of beta and gamma radiation from the soil, (ii) the gamma dose within the sample due to its own radioactivity, and (iii) the soil gamma dose in the proximity of boundaries between regions...... of differing radioactivity. It is confirmed that removal of the outer 2 mm of sample is adequate to remove influence from soil beta dose and estimates are made of the error introduced by non-removal. Other evaluations include variation of the soil gamma dose near the ground surface and it appears...... that the present practice of avoiding samples above a depth of 0.3 m may be over-cautious...

Actively shielded low level gamma - spectrometric system

International Nuclear Information System (INIS)

Mrdja, D.; Bikit, I.; Forkapic, S.; Slivka, J.; Veskovic, M.

2005-01-01

The results of the adjusting and testing of the actively shielded low level gamma-spectrometry system are presented. The veto action of the shield reduces the background in the energy region of 50 keV to the 2800 keV for about 3 times. (author) [sr

Shielding experiments for accelerator facilities

Energy Technology Data Exchange (ETDEWEB)

Nakashima, Hiroshi; Tanaka, Susumu; Sakamoto, Yukio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment] [and others

2000-06-01

A series of shielding experiments was carried out by using AVF cyclotron accelerator of TIARA at JAERI in order to validate shielding design methods for accelerator facilities in intermediate energy region. In this paper neutron transmission experiment through thick shields and radiation streaming experiment through a labyrinth are reported. (author)

Shielding experiments for accelerator facilities

International Nuclear Information System (INIS)

Nakashima, Hiroshi; Tanaka, Susumu; Sakamoto, Yukio

2000-01-01

A series of shielding experiments was carried out by using AVF cyclotron accelerator of TIARA at JAERI in order to validate shielding design methods for accelerator facilities in intermediate energy region. In this paper neutron transmission experiment through thick shields and radiation streaming experiment through a labyrinth are reported. (author)

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

International Nuclear Information System (INIS)

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

2012-03-01

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

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

International Nuclear Information System (INIS)

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

1993-01-01

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

Scintillation counter, segmented shield

International Nuclear Information System (INIS)

Olson, R.E.; Thumim, A.D.

1975-01-01

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

Gamma Irradiation Facility at Sandia National Laboratories, Albuquerque, New Mexico. Final environmental assessment

International Nuclear Information System (INIS)

1995-11-01

The US Department of Energy (DOE) has prepared an environmental assessment (EA) on the proposed construction and operation of a new Gamma Irradiation Facility (GIF) at Sandia National Laboratories/New Mexico (SNL/NM). This facility is needed to: enhance capabilities to assure technical excellence in nuclear weapon radiation environments testing, component development, and certification; comply with all applicable ES and H safeguards, standards, policies, and regulations; reduce personnel radiological exposure to comply with ALARA limits in accordance with DOE orders and standards; consolidate major gamma ray sources into a central, secured area; and reduce operational risks associated with operation of the GIF and LICA in their present locations. This proposed action provides for the design, construction, and operation of a new GIF located within TA V and the removal of the existing GIF and Low Intensity Cobalt Array (LICA). The proposed action includes potential demolition of the gamma shield walls and removal of equipment in the existing GIF and LICA. The shielding pool used by the existing GIF will remain as part of the ACRR facility. Transportation of the existing 60 Co sources from the existing LICA and GIF to the new facility is also included in the proposed action. Relocation of the gamma sources to the new GIF will be accomplished by similar techniques to those used to install the sources originally

LOFT shield tank steady state temperatures with addition of gamma and neutron shielding

International Nuclear Information System (INIS)

Kyllingstad, G.

1977-01-01

The effect of introducing a neutron and gamma shield into the annulus between the reactor vessel and the shield tank is analyzed. This addition has been proposed in order to intercept neutron streaming up the annulus during nuclear operations. Its installation will require removal of approximately 20- 1 / 2 inches of stainless steel foil insulation at the top of the annulus. The resulting conduction path is believed to result in increased water temperatures within the shield tank, possibly beyond the 150 0 F limit, and/or cooling of the reactor vessel nozzles such that adverse thermal stresses would be generated. A two dimensional thermal analysis using the finite element code COUPLE/MOD2 was done for the shield tank system illustrated in the figure (1). The reactor was assumed to be at full power, 55 MW (th), with a loop flow rate of 2.15 x 10 6 lbm/hr (268.4 kg/s) at 2250 psi (15.51 MPa). Calculations indicate a steady state shield tank water temperature of 140 0 F (60 0 C). This is below the 150 0 F (65.56 0 C) limit. Also, no significant changes in thermal gradients within the nozzle or reactor vessel wall are generated. A spacer between the gamma shield and the shield tank is recommended, however, in order to ensure free air circulation through the annulus

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.

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 radiation shielding analysis of lead-flyash concretes

International Nuclear Information System (INIS)

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

2015-01-01

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

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

Hot Cell Window Shielding Analysis Using MCNP

International Nuclear Information System (INIS)

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

2009-01-01

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

Safety guide data on radiation shielding in a reprocessing facility

International Nuclear Information System (INIS)

Sekiguchi, Noboru; Naito, Yoshitaka

1986-04-01

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

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

Science.gov (United States)

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

2018-05-01

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

A {beta} - {gamma} coincidence; Metodo de coincidencias {beta} - {gamma}

Energy Technology Data Exchange (ETDEWEB)

Agullo, F

1960-07-01

A {beta} - {gamma} coincidence method for absolute counting is given. The fundamental principles are revised and the experimental part is detailed. The results from {sup 1}98 Au irradiated in the JEN 1 Swimming pool reactor are given. The maximal accuracy is 1 per cent. (Author) 11 refs.

A conceptual gamma shield design using the DRP model computation

Energy Technology Data Exchange (ETDEWEB)

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

1997-12-31

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

Top-Down Beta Enhances Bottom-Up Gamma.

Science.gov (United States)

Richter, Craig G; Thompson, William H; Bosman, Conrado A; Fries, Pascal

2017-07-12

Several recent studies have demonstrated that the bottom-up signaling of a visual stimulus is subserved by interareal gamma-band synchronization, whereas top-down influences are mediated by alpha-beta band synchronization. These processes may implement top-down control of stimulus processing if top-down and bottom-up mediating rhythms are coupled via cross-frequency interaction. To test this possibility, we investigated Granger-causal influences among awake macaque primary visual area V1, higher visual area V4, and parietal control area 7a during attentional task performance. Top-down 7a-to-V1 beta-band influences enhanced visually driven V1-to-V4 gamma-band influences. This enhancement was spatially specific and largest when beta-band activity preceded gamma-band activity by ∼0.1 s, suggesting a causal effect of top-down processes on bottom-up processes. We propose that this cross-frequency interaction mechanistically subserves the attentional control of stimulus selection. SIGNIFICANCE STATEMENT Contemporary research indicates that the alpha-beta frequency band underlies top-down control, whereas the gamma-band mediates bottom-up stimulus processing. This arrangement inspires an attractive hypothesis, which posits that top-down beta-band influences directly modulate bottom-up gamma band influences via cross-frequency interaction. We evaluate this hypothesis determining that beta-band top-down influences from parietal area 7a to visual area V1 are correlated with bottom-up gamma frequency influences from V1 to area V4, in a spatially specific manner, and that this correlation is maximal when top-down activity precedes bottom-up activity. These results show that for top-down processes such as spatial attention, elevated top-down beta-band influences directly enhance feedforward stimulus-induced gamma-band processing, leading to enhancement of the selected stimulus. Copyright © 2017 Richter, Thompson et al.

Shielding design for positron emission tomography facility

International Nuclear Information System (INIS)

Abdallah, I.I.

2007-01-01

With the recent advent of readily available tracer isotopes, there has been marked increase in the number of hospital-based and free-standing positron emission tomography (PET) clinics. PET facilities employ relatively large activities of high-energy photon emitting isotopes, which can be dangerous to the health of humans and animals. This coupled with the current dose limits for radiation worker and members of the public can result in shielding requirements. This research contributes to the calculation of the appropriate shielding to keep the level of radiation within an acceptable recommended limit. Two different methods were used including measurements made at selected points of an operating PET facility and computer simulations by using Monte Carlo Transport Code. The measurements mainly concerned the radiation exposure at different points around facility using the survey meter detectors and Thermoluminescent Dosimeters (TLD). Then the set of manual calculation procedures were used to estimate the shielding requirements for a newly built PEF facility. The results from the measurement and the computer simulation were compared to the results obtained from the set manual calculation procedure. In general, the estimated weekly dose at the points of interest is lower than the regulatory limits for the little company of Mary Hospital. Furthermore, the density and the HVL for normal strength concrete and clay bricks are almost similar. In conclusion, PET facilities present somewhat different design requirements and are more likely to require additional radiation shielding. Therefore, existing shields at the little Company of Mary Hospital are in general found to be adequate and satisfactory and additional shielding was found necessary at the new PET facility in the department of Nuclear Medicine of the Dr. George Mukhari Hospital. By use of appropriate design, by implying specific shielding requirements and by maintaining good operating practices, radiation doses to

Low-level multicounter {beta}/{gamma} systems with external guards in surface and shallow underground laboratories

Energy Technology Data Exchange (ETDEWEB)

Theodorsson, P [Iceland Univ. (Iceland). Science Inst.

1997-03-01

When weak samples are measured it is important that they can be given ample counting time in order to obtain satisfactory accuracy and that the background count rate can be checked well. This calls for a high counting capacity, which multidetectors can bring us. I will discuss development possibilities of low-level {beta}/{gamma} multidetector systems with an external anticosmic shield that will in many cases be operated in underground laboratories. These simple and low-cost system can frequently help us in increasing the number of detectors. Three concepts are combined in these systems: (1) multidetectors, (2) an external anticosmic (or guard) detector arrangement and (3) overburden shielding. (orig.)

Maternal plasma concentrations of beta-lipotrophin, beta-endorphin and gamma-lipotrophin throughout pregnancy.

Science.gov (United States)

Browning, A J; Butt, W R; Lynch, S S; Shakespear, R A

1983-12-01

Plasma beta-LPH, beta-EP and gamma-LPH concentrations were measured by radioimmunoassay in 10 pregnant women from 12 weeks gestation until term and in nine women in the early follicular phase of the cycle. There was a progressive and significant rise in the concentration of all three peptides throughout pregnancy and by 32 weeks the concentrations of beta-LPH and beta-EP were greater than the corresponding concentrations in the follicular phase: gamma-LPH was greater than in the follicular phase by the end of pregnancy in those women who were delivered after 40 weeks. The ratio of beta-LPH to gamma-LPH did not change significantly throughout pregnancy, but there was a progressive fall in the beta-LPH/beta-EP ratio. The possible presence of a 'big LPH' to explain this finding is discussed.

Shield design for the Fusion Materials Irradiation Test facility

International Nuclear Information System (INIS)

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

1983-03-01

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

Separate assessment of natural beta and gamma dose-rates with TL from α-Al2O3:C single-crystal chips

International Nuclear Information System (INIS)

Kalchgruber, R.; Wagner, G.A.

2006-01-01

A measurement procedure was developed for fast and separate assessment of beta and gamma dose-rates in natural sediments using highly sensitive α-Al 2 O 3 :C single-crystal chips. The dosemeters were buried for periods from two days to three weeks in sediments with different layer structure and homogeneity. For each measurement, a pair of dosemeters was buried, in order to assess beta and gamma dose-rates separately. One dosemeter was wrapped only in thin plastic foil to shield it from alpha radiation, thus measuring beta + gamma components. The second one, used for the gamma component only, was packed additionally in a 1mm copper container for absorption of beta radiation. For calibration, another set of dosemeters was buried in reference soil with a well-known dose-rate and similar content of radioactive nuclides. By comparing the thermally stimulated luminescence signals from the dosemeters the gamma dose-rate and also, by subtraction, the beta dose-rate in the unknown soil could be determined. The calculated uncertainties were 5-7% and 10%, respectively. The resulting dose-rates for homogeneous and inhomogeneous media were compared with the results obtained by Ge- and on-site NaI-gamma-ray spectrometry. An agreement within 2σ-error limits was found for homogeneous media after only few days of exposure

New radiation protection calibration facility at CERN.

Science.gov (United States)

Brugger, Markus; Carbonez, Pierre; Pozzi, Fabio; Silari, Marco; Vincke, Helmut

2014-10-01

The CERN radiation protection group has designed a new state-of-the-art calibration laboratory to replace the present facility, which is >20 y old. The new laboratory, presently under construction, will be equipped with neutron and gamma sources, as well as an X-ray generator and a beta irradiator. The present work describes the project to design the facility, including the facility placement criteria, the 'point-zero' measurements and the shielding study performed via FLUKA Monte Carlo simulations. © The Author 2013. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

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

International Nuclear Information System (INIS)

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

2009-01-01

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

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

International Nuclear Information System (INIS)

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

2000-01-01

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

Determination of gamma ray shielding parameters of rocks and concrete

Science.gov (United States)

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

2018-03-01

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

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

International Nuclear Information System (INIS)

Ofori, K.

2011-07-01

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

Shielding technology for high energy radiation production facility

International Nuclear Information System (INIS)

Lee, Byung Chul; Kim, Heon Il

2004-06-01

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

Shielding Design and Radiation Shielding Evaluation for LSDS System Facility

International Nuclear Information System (INIS)

Kim, Younggook; Kim, Jeongdong; Lee, Yongdeok

2015-01-01

As the system characteristics, the target in the spectrometer emits approximately 1012 neutrons/s. To efficiently shield the neutron, the shielding door designs are proposed for the LSDS system through a comparison of the direct shield and maze designs. Hence, to guarantee the radiation safety for the facility, the door design is a compulsory course of the development of the LSDS system. To improve the shielding rates, 250x250 covering structure was added as a subsidiary around the spectrometer. In this study, the evaluations of the suggested shielding designs were conducted using MCNP code. The suggested door design and covering structures can shield the neutron efficiently, thus all evaluations of all conditions are satisfied within the public dose limits. From the Monte Carlo code simulation, Resin(Indoor type) and Tungsten(Outdoor type) were selected as the shielding door materials. From a comparative evaluation of the door thickness, In and Out door thickness was selected 50 cm

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

Response-surface models for deterministic effects of localized irradiation of the skin by discrete {beta}/{gamma} -emitting sources

Energy Technology Data Exchange (ETDEWEB)

Scott, B.R.

1995-12-01

Individuals who work at nuclear reactor facilities can be at risk for deterministic effects in the skin from exposure to discrete {Beta}- and {gamma}-emitting ({Beta}{gamma}E) sources (e.g., {Beta}{gamma}E hot particles) on the skin or clothing. Deterministic effects are non-cancer effects that have a threshold and increase in severity as dose increases (e.g., ulcer in skin). Hot {Beta}{gamma}E particles are {sup 60}Co- or nuclear fuel-derived particles with diameters > 10 {mu}m and < 3 mm and contain at least 3.7 kBq (0.1 {mu}Ci) of radioactivity. For such {Beta}{gamma}E sources on the skin, it is the beta component of the dose that is most important. To develop exposure limitation systems that adequately control exposure of workers to discrete {Beta}{gamma}E sources, models are needed for systems that adequately control exposure of workers to discrete {Beta}{gamma}E sources, models are needed for evaluating the risk of deterministic effects of localized {Beta} irradiation of the skin. The purpose of this study was to develop dose-rate and irradiated-area dependent, response-surface models for evaluating risks of significant deterministic effects of localized irradiation of the skin by discrete {Beta}{gamma}E sources and to use modeling results to recommend approaches to limiting occupational exposure to such sources. The significance of the research results as follows: (1) response-surface models are now available for evaluating the risk of specific deterministic effects of localized irradiation of the skin; (2) modeling results have been used to recommend approaches to limiting occupational exposure of workers to {Beta} radiation from {Beta}{gamma}E sources on the skin or on clothing; and (3) the generic irradiated-volume, weighting-factor approach to limiting exposure can be applied to other organs including the eye, the ear, and organs of the respiratory or gastrointestinal tract and can be used for both deterministic and stochastic effects.

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

Energy Technology Data Exchange (ETDEWEB)

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

2000-03-01

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

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

EFFECTS OF INTERFACES ON GAMMA SHIELDING

Energy Technology Data Exchange (ETDEWEB)

Clifford, C. E.

1963-06-15

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

Analytical methods and laboratory facility for the Defense Waste Processing Facility

International Nuclear Information System (INIS)

Coleman, C.J.; Dewberry, R.A.; Lethco, A.J.; Denard, C.D.

1985-01-01

This paper describes the analytical methods, instruments, and laboratory that will support vitrification of defense waste. The Defense Waste Processing Facility (DWPF) is now being constructed at Savannah River Plant (SRP). Beginning in 1989, SRP high-level defense waste will be immobilized in borosilicate glass for disposal in a federal repository. The DWPF will contain an analytical laboratory for performing process control analyses. Additional analyses will be performed for process history and process diagnostics. The DWPF analytical facility will consist of a large shielded sampling cell, three shielded analytical cells, a laboratory for instrumental analysis and chemical separations, and a counting room. Special instrumentation is being designed for use in the analytical cells, including microwave drying/dissolution apparatus, and remote pipetting devices. The instrumentation laboratory will contain inductively coupled plasma, atomic absorption, Moessbauer spectrometers, a carbon analyzer, and ion chromatography equipment. Counting equipment will include intrinsic germanium detectors, scintillation counters, Phoswich alpha, beta, gamma detectors, and a low-energy photon detector

Shielding of Medical Facilities. Shielding Design Considerations for PET-CT Facilities

International Nuclear Information System (INIS)

Cruzate, J.A.; Discacciatti, A.P.

2011-01-01

The radiological evaluation of a Positron Emission Tomography (PET) 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 evaluation embraces the distributions of rooms, the thickness and physical material of walls, floors and ceilings. This work detail the methodology used for making the assessment of a PET facility design taking into account only radioprotection aspects. The assessment results must be compared to the design requirements established by national regulations in order to determine whether or not, the facility complies with those requirements, both for workers and for members of the public. The analysis presented is useful for both, facility designers and regulators. In addition, some guidelines for improving the shielding design and working procedures are presented in order to help facility designer's job. (authors)

Gamma ray and neutron shielding properties of some concrete materials

International Nuclear Information System (INIS)

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

2011-01-01

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

Liquid argon as active shielding and coolant for bare germanium detectors. A novel background suppression method for the GERDA 0{nu}{beta}{beta} experiment

Energy Technology Data Exchange (ETDEWEB)

Peiffer, J.P.

2007-07-25

Two of the most important open questions in particle physics are whether neutrinos are their own anti-particles (Majorana particles) as required by most extensions of the StandardModel and the absolute values of the neutrino masses. The neutrinoless double beta (0{nu}{beta}{beta}) decay, which can be investigated using {sup 76}Ge (a double beta isotope), is the most sensitive probe for these properties. There is a claim for an evidence for the 0{nu}{beta}{beta} decay in the Heidelberg-Moscow (HdM) {sup 76}Ge experiment by a part of the HdM collaboration. The new {sup 76}Ge experiment Gerda aims to check this claim within one year with 15 kg.y of statistics in Phase I at a background level of {<=}10{sup -2} events/(kg.keV.y) and to go to higher sensitivity with 100 kg.y of statistics in Phase II at a background level of {<=}10{sup -3} events/(kg.keV.y). In Gerda bare germanium semiconductor detectors (enriched in {sup 76}Ge) will be operated in liquid argon (LAr). LAr serves as cryogenic coolant and as high purity shielding against external background. To reach the background level for Phase II, new methods are required to suppress the cosmogenic background of the diodes. The background from cosmogenically produced {sup 60}Co is expected to be {proportional_to}2.5.10{sup -3} events/(kg.keV.y). LAr scintillates in UV ({lambda}=128 nm) and a novel concept is to use this scintillation light as anti-coincidence signal for background suppression. In this work the efficiency of such a LAr scintillation veto was investigated for the first time. In a setup with 19 kg active LAr mass a suppression of a factor 3 has been achieved for {sup 60}Co and a factor 17 for {sup 232}Th around Q{sub {beta}}{sub {beta}} = 2039 keV. This suppression will further increase for a one ton active volume (factor O(100) for {sup 232}Th and {sup 60}Co). LAr scintillation can also be used as a powerful tool for background diagnostics. For this purpose a new, very stable and robust wavelength

New information on the $\\beta$-decay of $^{11}$Li from Doppler broadened $\\gamma$-lines

CERN Document Server

Fynbo, H O U; Cederkäll, J; Courtin, S; Dessagne, P; Jonson, B; Le Scornet, G; Nilsson, T; Nyman, G H; Poirier, E; Riisager, K; Tengblad, O; Wilhelmsen, K

2004-01-01

The $\\gamma$-ray spectrum following $\\beta$-decay of $^{11}$Li has been remeasured at the ISOLDE facility at CERN. Two new transitions were observed through the use of large Ge-detectors. Most $\\gamma$-decays will follow P-delayed neutron emission. Information on the energy of the neutron is derived from analysis of the $\\gamma$-line-shape and used to construct a partial decay scheme for $^{11}$Li. Lifetime values for the 1$^{-}$ and 2$^{-}$ levels in $^{10}$Be are also derived. A new partial decay scheme is presented. (43 refs).

Gamma dose from activation of internal shields in IRIS reactor.

Science.gov (United States)

Agosteo, Stefano; Cammi, Antonio; Garlati, Luisella; Lombardi, Carlo; Padovani, Enrico

2005-01-01

The International Reactor Innovative and Secure is a modular pressurised water reactor with an integral design. This means that all the primary system components, such as the steam generators, pumps, pressuriser and control rod drive mechanisms, are located inside the reactor vessel, which requires a large diameter. For the sake of better reliability and safety, it is desirable to achieve the reduction of vessel embrittlement as well as the lowering of the dose beyond the vessel. The former can be easily accomplished by the presence of a wide downcomer, filled with water, which surrounds the core region, while the latter needs the presence of additional internal shields. An optimal shielding configuration is under investigation, for reducing the ex-vessel dose due to activated internals and for limiting the amount of the biological shielding. MCNP 4C calculations were performed to evaluate the neutron and the gamma dose during operation and the 60Co activation of various shields configurations. The gamma dose beyond the vessel from activation of its structural components was estimated in a shutdown condition, with the Monte Carlo code FLUKA 2002 and the MicroShield software. The results of the two codes are in agreement and show that the dose is sufficiently low, even without an additional shield.

Gamma dose from activation of internal shields in IRIS reactor

International Nuclear Information System (INIS)

Agosteo, S.; Cammi, A.; Garlati, L.; Lombardi, C.; Padovani, E.

2005-01-01

The International Reactor Innovative and Secure is a modular pressurised water reactor with an integral design. This means that all the primary system components, such as the steam generators, pumps, pressurizer and control rod drive mechanisms, are located inside the reactor vessel, which requires a large diameter. For the sake of better reliability and safety, it is desirable to achieve the reduction of vessel embrittlement as well as the lowering of the dose beyond the vessel. The former can be easily accomplished by the presence of a wide downcomer, filled with water, which surrounds the core region, while the latter needs the presence of additional internal shields. An optimal shielding configuration is under investigation, for reducing the ex-vessel dose due to activated internals and for limiting the amount of the biological shielding. MCNP 4C calculations were performed to evaluate the neutron and the gamma dose during operation and the 60 Co activation of various shields configurations. The gamma dose beyond the vessel from activation of its structural components was estimated in a shutdown condition, with the Monte Carlo code FLUKA 2002 and the MicroShield software. The results of the two codes are in agreement and show that the dose is sufficiently low, even without an additional shield. (authors)

Analysis of radionuclide mixtures by {alpha}-{gamma} and {beta}-{gamma} coincidences using a simple device; Analyse de melanges de radionucleides par un dispositif simple de coincidences {alpha}-{gamma} et {beta}-{gamma}

Energy Technology Data Exchange (ETDEWEB)

Pottier, R; Berger, R [Commissariat a l' Energie Atomique, Centre d' Etudes Nucleaires de Fontenay-aux-Roses, 92 (France)

1966-06-01

A procedure is described for the qualitative and quantitative spectrographic analysis of radioactive sources containing two alpha-gamma emitters having the same alpha energy or two beta-gamma emitters having the same gamma energy. The main apparatus is a multichannel pulse-height analyzer including a coincidence circuit. The principle of the method, the synoptic scheme, the electronic device, the type of sources, and the precautions to be taken or the corrections to take into account are reported. The results obtained in solving the three following problems are discussed as examples of applications of the method: analysis of {sup 241}Am in alpha-gamma sources containing {sup 238}Pu; analysis of {sup 237}Np in beta-gamma sources containing {sup 239}Pu; and analysis of {sup 106}Ru-{sup 106}Rh in beta-gamma sources containing {sup 95}Zr-{sup 95}Nb. (authors) [French] Dans ce. rapport, on presente une methode d'analyse spectrographique qualitative et quantitative de sources radioactives contenant deux emetteurs alpha-gamma de meme energie alpha et deux emetteurs beta-gamma de meme energie gamma. L'organe principal est un analyseur d'amplitude a 400 canaux comprenant un circuit de coincidence. On decrit le principe de la methode, le schema synoptique, l'appareillage, le type des sources, les precautions a prendre ou les corrections a faire. On discute les resultats obtenus dans la solution des trois problemes suivants traites a titre d'application de la methode: 1. analyse d'americium 241 en presence de plutonium 238; 2. analyse de neptunium 237 en presence de plutonium 239; 3. analyse de ruthenium 106-rhodium 106 en presence de zirconium 95-niobium 95. (auteurs)

Gamma rays shielding parameters for white metal alloys

Science.gov (United States)

Kaur, Taranjot; Sharma, Jeewan; Singh, Tejbir

2018-05-01

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

Beta/gamma test problems for ITS

International Nuclear Information System (INIS)

Mei, G.T.

1993-01-01

The Integrated Tiger Series of Coupled Electron/Photon Monte Carlo Transport Codes (ITS 3.0, PC Version) was used at Oak Ridge National Laboratory (ORNL) to compare with and extend the experimental findings of the beta/gamma response of selected health physics instruments. In order to assure that ITS gives correct results, several beta/gamma problems have been tested. ITS was used to simulate these problems numerically, and results for each were compared to the problem's experimental or analytical results. ITS successfully predicted the experimental or analytical results of all tested problems within the statistical uncertainty inherent in the Monte Carlo method

Microscopic beta and gamma data for decay-heat needs

International Nuclear Information System (INIS)

Dickens, J.K.

1983-01-01

Microscopic beta and gamma data for decay-heat needs are defined as absolute-intensity spectral distributions of beta and gamma rays following radioactive decay of radionuclides created by, or following, the fission process. Four well-known evaluated data files, namely the US ENDF/B-V, the UK UKFPDD-2, the French BDN (for fission products), and the Japanese JNDC Nuclear Data Library, are reviewed. Comments regarding the analyses of experimental data (particularly gamma-ray data) are given; the need for complete beta-ray spectral measurements is emphasized. Suggestions on goals for near-term future experimental measurements are presented. 34 references

Upgrade of the muon veto and current status of the Dortmund low background HPGe facility

Energy Technology Data Exchange (ETDEWEB)

Goessling, Claus; Kroeninger, Kevin; Neddermann, Till; Nitsch, Christian; Quante, Thomas [TU Dortmund, Physik EIV, D-44221 Dortmund (Germany)

2015-07-01

The Dortmund Low Background HPGe Facility (DLB) is a germanium facility with heavy shielding located above ground. It's primary task is to provide material screening support for the COBRA experiment which was built to search for neutrinoless double beta decay. Germanium detectors used for low background gamma spectroscopy are usually operated under either a fairly low overburden (O(1m) water equivalent (mwe)) or high overburden, e.g. in specialised underground laboratories (O(>100 mwe)). In between, only a few facilities exist, such as the DLB. The artificial overburden of 10 mwe already shields the weak component of cosmic rays. The lead castle with a state-of-the-art neutron shielding as well as the active anti-cosmics veto detector enable low background gamma spectrometry with the advantage of good accessibility on the university campus. Throughout the last years improvements have been made especially on the cosmics veto and the MC simulation leading to an remarkable low integral background counting rate (40-2700 keV) of about 2.5228(52) counts/kg/min. The talk summarises the completed tasks and presents the current status.

The angular gamma flux in an iron slab shield

International Nuclear Information System (INIS)

Penkuhn, H.

1975-08-01

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

Development of special radiation shielding concretes using natural local materials and evaluation of their shielding characteristics

International Nuclear Information System (INIS)

Kharita, M. H.; Takeyeddin, M.; Al-Nassri, M.; Yousef, S.

2008-01-01

Concrete is one of the most important materials used for radiation shielding in facilities containing radioactive sources and radiation generating machines. The concrete shielding properties may vary depending on the composite of the concrete. Aggregates is the largest constituent (about 70-80% of the total weight of normal concrete). The aim of this work is to develop special concrete with good shielding properties for gamma and neutrons, using natural local materials. For this reason two types of typical concrete widely used in Syria (in Damascus and Aleppo) and four other types of concrete, using aggregates from different regions, have been prepared. The shielding properties of these six types were studied for gamma ray (from Cs-137 and Co-60 sources)and for neutrons (from am-Be source). A reduction of about 10% in the HVL was obtained for the concrete from Damascus in comparison with that from Aleppo, for both neutrons and gammas. One of the other four types of concrete (from Rajo site, mostly Hematite), was found to further reduce the HVL by about 10% for both neutrons and gamma rays.(author)

{beta} {gamma} porch detector; Detecteur portique {beta} {gamma}

Energy Technology Data Exchange (ETDEWEB)

Roulet, R [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

1963-07-01

This device is to be placed at the outside of reactors, hot laboratories and others where radioactive products are treated; it is intended to give the alarm when someone, passing through the porch is greatly contaminated, or carries, without his knowing, a radioactive substance. Being to be used in places where there might be an important ground noise, this device is provided with an automatic offset of this noise; an adjusting system of sensitivity allows to obtain a 15 {mu}Ci in {gamma} and 10 {mu}Ci in {beta} radioactive source, passing through the porch at the normal speed at which man is walking. A battery, set in buffer, allows working of the device, even when current is off. (author) [French] Cet appareil est destine a etre place a la sortie des reacteurs, laboratoires chauds ou autres laboratoires travaillant sur des produits radioactifs; son but est de donner une alarme lorsque quelqu'un, passant sous le portique, presente une forte contamination, ou surtout transporte par inadvertance un corps radioactif. Cet appareil devant etre utilise dans les lieux ou peut regner un bruit de fond important, possede une compensation automatique de ce bruit de fond; un reglage de la sensibilite permet d'obtenir au mieux un declenchement pour une source. de 15 {mu}Ci en {gamma} et 10 {mu}Ci en {beta} passant sous le portique a la vitesse normale d'un homme qui marche. Une batterie montee en tampon permet a l'appareil de fonctionner meme en cas de coupure de courant. (auteur)

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

CERN Document Server

Hong, J; Hailey, C J

2000-01-01

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

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.

Using natural local materials for developing special radiation shielding concretes, and deduction of its shielding characteristics

International Nuclear Information System (INIS)

Kharita, M. H.; Takeyeddin, M.; Al-Nassar, M.; Yousef, S.

2006-06-01

Concrete is considered as the most important material to be used for radiation shielding in facilities contain radioactive sources and radiation generating machines. The concrete shielding properties may vary depending on the construction of the concrete, which is highly relative to the composing aggregates i.e. aggregates consist about 70 - 80% of the total weight of normal concrete. In this project tow types of concrete used in Syria (in Damascus and Aleppo) had been studied and their shielding properties were defined for gamma ray from Cs-137 and Co-60 sources, and for neutrons from Am-Be source. About 10% reduction in HVL was found in the comparison between the tow concrete types for both neutrons and gammas. Some other types of concrete were studied using aggregates from different regions in Syria, to improve the shielding properties of concrete, and another 10% of reduction was achieved in comparison with Damascene concrete (20% in comparison with the concrete from Aleppo) for both neutrons and gamma rays. (author)

Where have the neutrons gone: A history of the Tower Shielding Facility

International Nuclear Information System (INIS)

Muckenthaler, F.J.

1992-01-01

In the early 1950's, the concept of the unit shield for the nuclear powered aircraft reactor changed to one of the divided shield concept where the reactor and crew compartment shared the shielding load. Design calculations for the divided shield were being made based on data obtained in studies for the, unit shield. It was believed that these divided shield designs were subject to error, the magnitude of which could not be estimated. This belief led to the design of the Tower Shielding Facility where divided-shield-type measurements could be made without interference from ground or structural scattering. This paper discusses that facility, its reactors, and some chosen experiments from the list of many that were performed at that facility during the past 38 years

beta. and. gamma. -comparative dose estimates on Enewetak Atoll

Energy Technology Data Exchange (ETDEWEB)

Crase, K.W.; Gudiksen, P.H.; Robison, W.L. (California Univ., Livermore (USA). Lawrence Livermore National Lab.)

1982-05-01

Enewetak Atoll in the Pacific is used for atmospheric testing of U.S. nuclear weapons. Beta dose and ..gamma..-ray exposure measurements were made on two islands of the Enewetak Atoll during July-August 1976 to determine the ..beta.. and low energy ..gamma..-contribution to the total external radiation doses to the returning Marshallese. Measurements were made at numerous locations with thermoluminescent dosimeters (TLD), pressurized ionization chambers, portable NaI detectors, and thin-window pancake GM probes. Results of the TLD measurements with and without a ..beta..-attenuator indicate that approx. 29% of the total dose rate at 1 m in air is due to ..beta..- or low energy ..gamma..-contribution. The contribution at any particular site, however, is reduced by vegetation. Integral 30-yr external shallow dose estimates for future inhabitants were made and compared with external dose estimates of a previous large scale radiological survey. Integral 30-yr shallow external dose estimates are 25-50% higher than whole body estimates. Due to the low penetrating ability of the ..beta..'s or low energy ..gamma..'s, however, several remedial actions can be taken to reduce the shallow dose contribution to the total external dose.

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)

Numerical simulation of a reinforced concrete shield around a nuclear reactor

International Nuclear Information System (INIS)

Mahama, Mumuni Salifu

1996-02-01

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

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.)

The Tower Shielding Facility: Its glorious past

Energy Technology Data Exchange (ETDEWEB)

Muckenthaler, F.J.

1997-05-07

The Tower Shielding Facility (TSF) is the only reactor facility in the US that was designed and built for radiation-shielding studies in which both the reactor source and shield samples could be raised into the air to allow measurements to be made without interference from ground scattering or other spurious effects. The TSF proved its usefulness as many different programs were successfully completed. It became active in work for the Defense Atomic Support Agency (DASA) Space Nuclear Auxiliary Power, Defense Nuclear Agency, Liquid Metal Fast Breeder Reactor Program, the Gas-Cooled and High-Temperature Gas-Cooled Reactor programs, and the Japanese-American Shielding Program of Experimental Research, just to mention a few of the more extensive ones. The history of the TSF as presented in this report describes the various experiments that were performed using the different reactors. The experiments are categorized as to the programs which they supported and placed in corresponding chapters. The experiments are described in modest detail, along with their purpose when appropriate. Discussion of the results is minimal, but references are given to more extensive topical reports.

Ford motor company NDE facility shielding design

International Nuclear Information System (INIS)

Metzger, R. L.; Van Riper, K. A.; Jones, M. H.

2005-01-01

Ford Motor Company proposed the construction of a large non-destructive evaluation laboratory for radiography of automotive power train components. The authors were commissioned to design the shielding and to survey the completed facility for compliance with radiation doses for occupationally and non-occupationally exposed personnel. The two X-ray sources are Varian Linatron 3000 accelerators operating at 9-11 MV. One performs computed tomography of automotive transmissions, while the other does real-time radiography of operating engines and transmissions. The shield thickness for the primary barrier and all secondary barriers were determined by point-kernel techniques. Point-kernel techniques did not work well for skyshine calculations and locations where multiple sources (e.g. tube head leakage and various scatter fields) impacted doses. Shielding for these areas was determined using transport calculations. A number of MCNP [Briesmeister, J. F. MCNPCA general Monte Carlo N-particle transport code version 4B. Los Alamos National Laboratory Manual (1997)] calculations focused on skyshine estimates and the office areas. Measurements on the operational facility confirmed the shielding calculations. (authors)

The Tower Shielding Facility: Its glorious past

International Nuclear Information System (INIS)

Muckenthaler, F.J.

1997-01-01

The Tower Shielding Facility (TSF) is the only reactor facility in the US that was designed and built for radiation-shielding studies in which both the reactor source and shield samples could be raised into the air to allow measurements to be made without interference from ground scattering or other spurious effects. The TSF proved its usefulness as many different programs were successfully completed. It became active in work for the Defense Atomic Support Agency (DASA) Space Nuclear Auxiliary Power, Defense Nuclear Agency, Liquid Metal Fast Breeder Reactor Program, the Gas-Cooled and High-Temperature Gas-Cooled Reactor programs, and the Japanese-American Shielding Program of Experimental Research, just to mention a few of the more extensive ones. The history of the TSF as presented in this report describes the various experiments that were performed using the different reactors. The experiments are categorized as to the programs which they supported and placed in corresponding chapters. The experiments are described in modest detail, along with their purpose when appropriate. Discussion of the results is minimal, but references are given to more extensive topical reports

Ford Motor Company NDE facility shielding design.

Science.gov (United States)

Metzger, Robert L; Van Riper, Kenneth A; Jones, Martin H

2005-01-01

Ford Motor Company proposed the construction of a large non-destructive evaluation laboratory for radiography of automotive power train components. The authors were commissioned to design the shielding and to survey the completed facility for compliance with radiation doses for occupationally and non-occupationally exposed personnel. The two X-ray sources are Varian Linatron 3000 accelerators operating at 9-11 MV. One performs computed tomography of automotive transmissions, while the other does real-time radiography of operating engines and transmissions. The shield thickness for the primary barrier and all secondary barriers were determined by point-kernel techniques. Point-kernel techniques did not work well for skyshine calculations and locations where multiple sources (e.g. tube head leakage and various scatter fields) impacted doses. Shielding for these areas was determined using transport calculations. A number of MCNP [Briesmeister, J. F. MCNPCA general Monte Carlo N-particle transport code version 4B. Los Alamos National Laboratory Manual (1997)] calculations focused on skyshine estimates and the office areas. Measurements on the operational facility confirmed the shielding calculations.

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

Data.gov (United States)

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

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

Data.gov (United States)

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

Dosimetry and Shielding of X and Gamma Radiation

International Nuclear Information System (INIS)

Oncescu, M.; Panaitescu, I.

1992-01-01

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

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

International Nuclear Information System (INIS)

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

1999-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Joenemalm, C; Malen, K

1966-10-15

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

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

International Nuclear Information System (INIS)

Joenemalm, C.; Malen, K

1966-10-01

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

Scintillation counter, maximum gamma aspect

International Nuclear Information System (INIS)

Thumim, A.D.

1975-01-01

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

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

Science.gov (United States)

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

2018-04-01

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

Hard beta and gamma emissions of 124I. Impact on occupational dose in PET/CT.

Science.gov (United States)

Kemerink, G J; Franssen, R; Visser, M G W; Urbach, C J A; Halders, S G E A; Frantzen, M J; Brans, B; Teule, G J J; Mottaghy, F M

2011-01-01

The hard beta and gamma radiation of 124I can cause high doses to PET/CT workers. In this study we tried to quantify this occupational exposure and to optimize radioprotection. Thin MCP-Ns thermoluminescent dosimeters suitable for measuring beta and gamma radiation were used for extremity dosimetry, active personal dosimeters for whole-body dosimetry. Extremity doses were determined during dispensing of 124I and oral administration of the activity to the patient, the body dose during all phases of the PET/CT procedure. In addition, dose rates of vials and syringes as used in clinical practice were measured. The procedure for dispensing 124I was optimized using newly developed shielding. Skin dose rates up to 100 mSv/min were measured when in contact with the manufacturer's vial containing 370 MBq of 124I. For an unshielded 5 ml syringe the positron skin dose was about seven times the gamma dose. Before optimization of the preparation of 124I, using an already reasonably safe technique, the highest mean skin dose caused by handling 370 MBq was 1.9 mSv (max. 4.4 mSv). After optimization the skin dose was below 0.2 mSv. The highly energetic positrons emitted by 124I can cause high skin doses if radioprotection is poor. Under optimized conditions occupational doses are acceptable. Education of workers is of paramount importance.

Using support vector machine to predict beta- and gamma-turns in proteins.

Science.gov (United States)

Hu, Xiuzhen; Li, Qianzhong

2008-09-01

By using the composite vector with increment of diversity, position conservation scoring function, and predictive secondary structures to express the information of sequence, a support vector machine (SVM) algorithm for predicting beta- and gamma-turns in the proteins is proposed. The 426 and 320 nonhomologous protein chains described by Guruprasad and Rajkumar (Guruprasad and Rajkumar J. Biosci 2000, 25,143) are used for training and testing the predictive model of the beta- and gamma-turns, respectively. The overall prediction accuracy and the Matthews correlation coefficient in 7-fold cross-validation are 79.8% and 0.47, respectively, for the beta-turns. The overall prediction accuracy in 5-fold cross-validation is 61.0% for the gamma-turns. These results are significantly higher than the other algorithms in the prediction of beta- and gamma-turns using the same datasets. In addition, the 547 and 823 nonhomologous protein chains described by Fuchs and Alix (Fuchs and Alix Proteins: Struct Funct Bioinform 2005, 59, 828) are used for training and testing the predictive model of the beta- and gamma-turns, and better results are obtained. This algorithm may be helpful to improve the performance of protein turns' prediction. To ensure the ability of the SVM method to correctly classify beta-turn and non-beta-turn (gamma-turn and non-gamma-turn), the receiver operating characteristic threshold independent measure curves are provided. (c) 2008 Wiley Periodicals, Inc.

beta- and gamma-Comparative dose estimates on Eniwetok Atoll

Energy Technology Data Exchange (ETDEWEB)

Crase, K.W.; Gudiksen, P.H.; Robison, W.L.

1982-05-01

Eniwetok Atoll is one of the Pacific atolls used for atmospheric testing of U.S. nuclear weapons. Beta dose and gamma-ray exposure measurements were made on two islands of the Eniwetok Atoll during July-August 1976 to determine the beta and low energy gamma-contribution to the total external radiation doses to the returning Marshallese. Measurements were made at numerous locations with thermoluminescent dosimeters (TLD), pressurized ionization chambers, portable NaI detectors, and thin-window pancake GM probes. Results of the TLD measurements with and without a beta-attenuator indicate that approx. 29% of the total dose rate at 1 m in air is due to beta- or low energy gamma-contribution. The contribution at any particular site, however, is somewhat dependent on ground cover, since a minimal amount of vegetation will reduce it significantly from that over bare soil, but thick stands of vegetation have little effect on any further reductions. Integral 30-yr external shallow dose estimates for future inhabitants were made and compared with external dose estimates of a previous large scale radiological survey (En73). Integral 30-yr shallow external dose estimates are 25-50% higher than whole body estimates. Due to the low penetrating ability of the beta's or low energy gamma's, however, several remedial actions can be taken to reduce the shallow dose contribution to the total external dose.

Study of local Agregate for Gamma radiation concrete shield

International Nuclear Information System (INIS)

Tochrul-Binowo; Endro-Kismolo; Darsono

1996-01-01

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

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

International Nuclear Information System (INIS)

Yuliati, Helfi; Akhadi, Mukhlis

2000-01-01

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

The Prompt Gamma Neutron Activation Analysis Facility at ICN-Pitesti

International Nuclear Information System (INIS)

Barbos, D.; Paunoiu, C.; Mladin, M.; Cosma, C.

2008-01-01

PGNAA is a very widely applicable technique for determining the presence and amount of many elements simultaneously in samples ranging in size from micrograms to many grams. PGNAA is characterized by its capability for nondestructive multi-elemental analysis and its ability to analyse elements that cannot be determined by INAA. By means of this PGNAA method we are able to increase the performance of INAA method. A facility has been developed at Institute for Nuclear Research-Pitesti so that the unique features of prompt gamma-ray neutron activation analysis can be used to measure trace and major elements in samples. The facility is linked at the radial neutron beam tube at ACPR-TRIGA reactor. During the PGNAA-facility is in use the ACPR reactor will be operated in steady-state mode at 250 KW maximum power. The facility consists of a radial beam-port, external sample position with shielding, and induced prompt gamma-ray counting system.Thermal neutron flux with energy lower than cadmium cut-off at the sample position was measured using thin gold foil is: φ scd = 1.10 6 n/cm 2 /s with a cadmium ratio of:80.The gamma-ray detection system consist of an HpGe detector of 16% efficiency (detector model GC1518) with 1.85 keV resolution capability. The HpGe is mounted with its axis at 90 deg. with respect to the incident neutron beam at distance about 200mm from the sample position. To establish the performance capabilities of the facility, irradiation of pure element or sample compound standards were performed to identify the gama-ray energies from each element and their count rates

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

International Nuclear Information System (INIS)

Dunley, Leonardo Souza

2002-01-01

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

Comparative study of tungsten and lead as gamma ray shielding material

International Nuclear Information System (INIS)

Wang Jian; Zou Shuliang

2011-01-01

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

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

International Nuclear Information System (INIS)

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

2016-01-01

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

Gamma sensitivity of the Eberline PCM-1

International Nuclear Information System (INIS)

Blanton, J.D.

1988-01-01

This paper reports that normally, alarm setpoints for the Eberline PCM-1 series personnel contamination monitors are calculated based upon efficiencies measured using 100-cm 2 or larger beta or mixed beta-gamma sources. This simulates the type of contamination most frequently encountered on personnel and clothing--low-level, distributed beta-gamma emitters. In most circumstances, the PCM-1's sensitivity to the other type of contamination encountered in nuclear plant work--hot particles--would be expected to be the same as, or better than, its sensitivity to distributed contamination. However, particles that are deposited on skin or clothing can be partially shielded from the view of the PCM-1's detectors. In these situations, the PCm-1's sensitivity to gamma radiation may be more relevant than its sensitivity to betas

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

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

Science.gov (United States)

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

2018-05-01

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

SU-E-T-400: Evaluation of Shielding and Activation at Two Pencil Beam Scanning Proton Facilities

International Nuclear Information System (INIS)

Remmes, N; Mundy, D; Classic, K; Beltran, C; Kruse, J; Herman, M; Stoker, J; Nelson, K; Bues, M

2015-01-01

Purpose: To verify acceptably low dose levels around two newly constructed identical pencil beam scanning proton therapy facilities and to evaluate accuracy of pre-construction shielding calculations. Methods: Dose measurements were taken at select points of interest using a WENDI-2 style wide-energy neutron detector. Measurements were compared to pre-construction shielding calculations. Radiation badges with neutron dose measurement capabilities were worn by personnel and also placed at points throughout the facilities. Seven neutron and gamma detectors were permanently installed throughout the facility, continuously logging data. Potential activation hazards have also been investigated. Dose rates near water tanks immediately after prolonged irradiation have been measured. Equipment inside the treatment room and accelerator vault has been surveyed and/or wipe tested. Air filters from air handling units, sticky mats placed outside of the accelerator vault, and water samples from the magnet cooling water loops have also been tested. Results: All radiation badges have been returned with readings below the reporting minimum. Measurements of mats, air filters, cooling water, wipe tests and surveys of equipment that has not been placed in the beam have all come back at background levels. All survey measurements show the analytical shielding calculations to be conservative by at least a factor of 2. No anomalous events have been identified by the building radiation monitoring system. Measurements of dose rates close to scanning water tanks have shown dose rates of approximately 10 mrem/hr with a half-life less than 5 minutes. Measurements around the accelerator show some areas with dose rates slightly higher than 10 mrem/hr. Conclusion: The shielding design is shown to be adequate. Measured dose rates are below those predicted by shielding calculations. Activation hazards are minimal except in certain very well defined areas within the accelerator vault and for objects

High Sensitivity Detection of Xe Isotopes Via Beta-Gamma Coincidence Counting

International Nuclear Information System (INIS)

Bowyer, Ted W.; McIntyre, Justin I.; Reeder, Paul L.

1999-01-01

Measurement of xenon fission product isotopes is a key element in the global network being established to monitor the Comprehensive Nuclear-Test-Ban Treaty. Pacific Northwest National Laboratory has developed an automated system for separating Xe from air which includes a beta-gamma counting system for 131mXe, 133mXe, 133Xe, and 135Xe. Betas and conversion electrons are detected in a plastic scintillation cell containing the Xe sample. The counting geometry is nearly 100% for beta and conversion electrons. The resolution in the pulse height spectrum from the plastic scintillator is sufficient to observe distinct peaks for specific conversion electrons. Gamma and X-rays are detected in a NaI(Tl) scintillation detector which surrounds the plastic scintillator sample cell. Two-dimensional pulse height spectra of gamma energy versus beta energy are obtained. Each of the four xenon isotopes has a distinctive signature in the two-dimensional energy array. The details of the counting system, examples of two-dimensional beta-gamma data, and operational experience with this counting system will be described

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

International Nuclear Information System (INIS)

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

2015-01-01

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

Design and operation of a prototype incinerator for beta-gamma waste

International Nuclear Information System (INIS)

Farber, M.G.; Hootman, H.E.; Becker, G.W. Jr.; Makohon, P.A.

1981-01-01

A full-scale test incinerator has been built at the Savannah River Laboratory to provide a design basis for a radioactive facility that will burn low-level beta-gamma contaminated waste. The processing steps include waste feed loading, incineration, ash residue packaging, and off-gas cleanup. Both solid and liquid waste will be incinerated during the test program. The components of the solid waste are cellulose, latex, polyethylene, and PVC; the solvent is composed of n-paraffin and TBP. A research program will confirm the feasibility of the design and determine the operating parameters

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)

Field measurement and interpretation of beta doses and dose rates

International Nuclear Information System (INIS)

Selby, J.M.; Swinth, K.L.; Hooker, C.D.; Kenoyer, J.L.

1983-01-01

A large number of portable survey instruments employing G.M., ionization chamber, and scintillation detectors used for gamma measurements are also used for monitoring in beta fields by using removable shields to separate the beta and gamma components of the radiation field. The difference does not correspond to an absorbed dose rate for the beta field due to a variety of factors. Among these factors are the dependence on beta energy, source-detector geometries, mixed fields and variable ambient conditions. Attempting to use such measurements directly can lead to errors as high as a factor of 100. Appropriate calibrations and correction factors can be used to reduce the errors in beta measurements to a tolerable level

Configuration Design of Detector Shielding for Gamma Prompt Analysis

International Nuclear Information System (INIS)

Elin-Nuraini; Darsono; Elisabeth

2000-01-01

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

Effects of increased shielding on gamma-radiation levels within spacecraft

Science.gov (United States)

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

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

Maternal and cord plasma concentrations of beta-lipotrophin, beta-endorphin and gamma-lipotrophin at delivery; effect of analgesia.

Science.gov (United States)

Browning, A J; Butt, W R; Lynch, S S; Shakespear, R A; Crawford, J S

1983-12-01

Maternal venous plasma concentrations of beta-LPH, beta-EP and gamma-LPH were compared in (i) patients undergoing vaginal delivery, 11 with an epidural block and 13 with pethidine and nitrous oxide or no analgesics; (ii) patients delivered by caesarean section, 7 under epidural block and 8 under general anaesthesia. Patients delivered by either method under epidural block had significantly lower levels of all three peptides than those receiving no epidural. There were significant negative correlations between umbilical vein beta-LPH, beta-EP and gamma-LPH concentrations and umbilical artery pH and positive correlations between beta-LPH and beta-EP but not gamma-LPH and cord PCO2 in 29 patients. There was no relation between cord levels of any of the three peptides and the method of analgesia or the route of delivery. Although concentrations of all three peptides were closely correlated to one another in either maternal or cord plasma, there was no relationship between maternal and fetal levels.

Work for radiation shielding concrete in large-scaled radiation facilities

International Nuclear Information System (INIS)

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

1980-01-01

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

Rat beta-LPH, gamma-LPH and beta-endorphin biosynthesized by isolated cells of pars intermedia and pars distalis

International Nuclear Information System (INIS)

Gianoulakis, C.; Seidah, N.G.; Routhier, R.; Chretien, M.

1980-01-01

Rats pars intermedia cells were incubated for 3 h with the following amino-acids: a) 35 S-methionine and 3 H-phenylalamine; b) 3 H-valine; and c) 3 H-valine and 3 H-lysine. Radioactive gamma-lipotropin, beta-lipotropin and beta-endorphin were purified on carboxy- methyl-cellulose and characterized by polyacrylamide disc gel electrophoresis af pH 4.5, molecular weight estimation and microsequencing. Rat gamma-lipotropin was shown to differ slightly from ovine gamma-lipotropin in its NH 2 -terminal amino acid sequence, in containing no methionine and having phenylalanine at position 6, valine at positions 13 and 27, and lysine at position 20. The same variations were observed in the sequence of rat beta-lipotropin, while rat beta-endorphin was shown to be identical to the ovine beta-endorphin. Following a 3-h pulse of rat pars distalis, the cells were extracted with care to avoid beta-lipotropin degradation by proteolytic enzymes. A peptide was purified and identified to be rat beta-endorphin, thus demonstrating that beta-endorphin is biosynthesized in pars distalis and is not an extraction artifact. (author)

Advanced facilities for radiochemistry at Harwell

International Nuclear Information System (INIS)

1985-01-01

The leaflets in this folder describe the latest addition to Harwell's active handling capability. This is a high level alpha, beta, gamma facility designed specifically for undertaking chemical research and development work. It is based on using high integrity containment boxes which are housed in concrete shielded enclosures. The active boxes can be removed and transferred remotely to a support area where they, and any associated equipment, can be decontaminated and serviced whilst a new fully commissioned box can be readily brought into service. The facility fulfills the principle of ALARA and is sufficiently flexible to accommodate a wide range of active handling requirements. It is supported by a suite of medium active handling cells, radiochemical laboratories and, as necessary, facilities of other scientific and engineering disciplines. The leaflets are: report on conceptual aspects; Techsheet 'Remote handling facility - Salient information'; Techsheet 'Project capabilities'; and 4 sheets of diagrams showing details of the facility. (U.K.)

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

Science.gov (United States)

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

2018-04-01

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

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

International Nuclear Information System (INIS)

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

1999-01-01

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

Radiation shielding design for a hot repair facility

International Nuclear Information System (INIS)

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

1991-01-01

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

New facility shield design criteria

International Nuclear Information System (INIS)

Howell, W.P.

1981-07-01

The purpose of the criteria presented here is to provide standard guidance for the design of nuclear radiation shields thoughout new facilities. These criteria are required to assure a consistent and integrated design that can be operated safely and economically within the DOE standards. The scope of this report is confined to the consideration of radiation shielding for contained sources. The whole body dose limit established by the DOE applies to all doses which are generally distributed throughout the trunk of the body. Therefore, where the whole body is the critical organ for an internally deposited radionuclide, the whole body dose limit applies to the sum of doses received must assure control of the concentration of radionuclides in the building atmosphere and thereby limit the dose from internal sources

PC based alpha, beta and gamma counting system

International Nuclear Information System (INIS)

Bhattacharya, S.; Gopalakrishnan, K.R.; Bairi, B.R.

1992-01-01

In the field of radiation protection and safety a large number of samples have to be assayed for estimation of their radioactivity at various stages in nuclear fuel. A PC based alpha, beta and gamma counting system has been designed for the above application. The system is fully automatic and requires minimum of operator intervention. The main units in the system are detector and pulse processing electronics, sample changer and assembly and IBM PC interface and supporting software. The main function of interface is to control the sample changer which is basically loading and unloading of sample, data acquisition and further processing by software. The pulses from detector are analysed in a pulse shape discriminator. A pulse identified as an alpha, beta or gamma event is converted to digital pulse. These digital pulses are accumulated in the three channels on PC interface each corresponding to alpha, beta and gamma. The sample movements are controlled by interface depending upon sample position. The software has been developed so as to maintain user friendliness and convenience of the operator. Various selection modes for parameters and operation of system provide lot of flexibility in operation of the system. (author). 1 fig

Evaluation of RayXpert® for shielding design of medical facilities

Science.gov (United States)

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.

Arylation of beta, gamma-unsaturated lactones by a Heck-Matsuda reaction: an unexpected route to aryldiazene butenolides and pyridazinones

Energy Technology Data Exchange (ETDEWEB)

Taylor, Jason G.; Correia, Carlos Roque D., E-mail: roque@iqm.unicamp.b [Universidade Estadual de Campinas (UNICAMP), SP (Brazil). Inst. de Quimica

2010-07-01

The palladium catalysed coupling of aryldiazonium salts with {beta}-{gamma}-unsaturated lactones under basic conditions has been investigated. Both (3H)-furanone and {alpha}-angelicalactone were evaluated as substrates in the Heck Matsuda reaction but both failed to afford the desired arylated butenolides. Under basic conditions, {beta}-{gamma}-unsaturated lactones generate highly nucleophilic enolates that preferentially undergo azo coupling reactions with arenediazonium salts to afford aryldiazene butenolides. The electronic and steric effect of the substituents on the aryldiazonium salt in the azo coupling reaction is described. Aryldiazene-lactone derivatives were obtained in good yields from a highly facile and straightforward procedure. An aminoisomaleimide was formed from (3H)-furanone and cyclized to the corresponding pyridazinones in modest yield. (author)

Shield evaluation and validation for design and operation of facility for treatment of legacy Intermediate Level Radioactive Liquid Waste (ILW)

International Nuclear Information System (INIS)

Deepa, A.; Jakhete, A.P.; Rathish, K.R.; Saroj, S.K.; Patel, H.S.; Gopalakrishnan, R.K.; Gangadharan, Anand; Singh, Neelima

2014-01-01

An ion exchange treatment facility has been commissioned at PRIX facility, for the treatment of legacy ILW generated at reprocessing plant, Trombay. The treatment system is based on the deployment of selective sorbents for removal of cesium and strontium from ILW. Activity concentration due to beta emitters likely to be processed is of the order of 111-1850 MBq/l. Dose rates in different areas of the facility were evaluated using shielding code and design input. Present work give details of the comparison of dose rates estimated and dose rates measured at various stages of the processing of ILW. At PRIX, the ILW treatment system comprises of shielded IX columns (two cesium and one strontium) housed in a MS cubicle the process lines inlet and outlet of IX treatment system and effluent storage tanks. The MS cubicle, prefilter and piping are housed in a process cell of 500 mm concrete shielding. Effluent storage tanks are outside processing building. Theoretical assessment of expected dose rates were carried out prior to installation of various systems in different areas of PRIX. Dose rate on IX column and MS cubicle for a maximum inventory of 3.7x10 7 MBq of 137 Cs and its contribution in operating gallery was estimated

Gamma radiation shielding materials improved with burning resistance

International Nuclear Information System (INIS)

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

1985-01-01

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

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

Science.gov (United States)

Ardiyati, Tanti; Rozali, Bang; Kasmudin

2018-02-01

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

Design and Shielding of Radiotherapy Treatment Facilities; IPEM Report 75, 2nd Edition

Science.gov (United States)

Horton, Patrick; Eaton, David

2017-07-01

Design and Shielding of Radiotherapy Treatment Facilities provides readers with a single point of reference for protection advice to the construction and modification of radiotherapy facilities. The book assembles a faculty of national and international experts on all modalities including megavoltage and kilovoltage photons, brachytherapy and high-energy particles, and on conventional and Monte Carlo shielding calculations. This book is a comprehensive reference for qualified experts and radiation-shielding designers in radiation physics and also useful to anyone involved in the design of radiotherapy facilities.

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

Alpha/beta(gamma ray) discrimination and spillover quantification with a BaF2 scintillator

International Nuclear Information System (INIS)

DeVol, T.A.; Fjeld, R.A.

1994-01-01

A simple pulse shape discrimination technique was used to separate alpha and beta(gamma ray) interactions in a BaF 2 scintillator. The separation was not ideal, resulting in a 5.1% spillover of alpha interactions into the beta(gamma ray) channel and 11.9% spillover of beta(gamma ray) interactions into the alpha channel for a set pulse shape discriminator. The misclassification of events was reduced by post-processing the data using either a simple analytical technique or a more complex linear least squares technique. Both techniques typically reduced the difference between the expected and calculated interaction rates to <10% when the ratio of beta(gamma ray) to alpha count rate was less than 100 : 1. ((orig.))

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

Science.gov (United States)

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

2017-10-01

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

Y-formalism and curved {beta}-{gamma} systems

Energy Technology Data Exchange (ETDEWEB)

Grassi, Pietro Antonio [DISTA, Universita del Piemonte Orientale, via Bellini 25/g, 15100 Alessandria (Italy); INFN - Sezione di Torino (Italy)], E-mail: antonio.pietro.grassi@cern.ch; Oda, Ichiro [Department of Physics, Faculty of Science, University of the Ryukyus, Nishihara, Okinawa 903-0213 (Japan); Tonin, Mario [Dipartimento di Fisica, Universita degli Studi di Padova, INFN, Sezionedi Padova, Via F. Marzolo 8, 35131 Padova (Italy)

2009-01-01

We adopt the Y-formalism to study {beta}-{gamma} systems on hypersurfaces. We compute the operator product expansions of gauge-invariant currents and we discuss some applications of the Y-formalism to model on Calabi-Yau spaces.

Gamma ampersand beta-gamma storm water monitor operability

International Nuclear Information System (INIS)

Tshiskiku, E.M.

1993-01-01

High Level Waste (HLW) facilities have nine storm water monitors that monitor storm water run off from different process areas for Cesium 137, a Gamma emitter. F - Area has three monitors: 907-2F, 907-3F and 907-4F while H - Area has six monitors: 907-2H, 907-3H, 907-4H, 907-5H, 907-6H and 907-7H (See attachments number-sign 1, number-sign 2 and number-sign 3 for location). In addition to monitoring for Cesium, 907-6H and 907-7H monitor for Strontium-90, a Beta emitter. Each monitor is associated with one of the following diversion gate encasements 907-1H, 241-15H, 241-51H, 907-1F or 241-23F. Normal flow of storm water from these diversion gate encasements is to the Four Mile Creek. When a storm water monitor detects radioactivity at a level exceeding the Four Mile Creek discharge limit, the monitor causes repositioning of the associated diversion gate to discharge to the H - Area retention basin 281-8H or the F - Area retention basin 281-8F. In response to recent OSR interpretation of storm water monitor calibration requirements, this report is provided to document operability and accuracy of radiation detection

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

Energy Technology Data Exchange (ETDEWEB)

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

2004-07-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-11

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

Support vector machines for prediction and analysis of beta and gamma-turns in proteins.

Science.gov (United States)

Pham, Tho Hoan; Satou, Kenji; Ho, Tu Bao

2005-04-01

Tight turns have long been recognized as one of the three important features of proteins, together with alpha-helix and beta-sheet. Tight turns play an important role in globular proteins from both the structural and functional points of view. More than 90% tight turns are beta-turns and most of the rest are gamma-turns. Analysis and prediction of beta-turns and gamma-turns is very useful for design of new molecules such as drugs, pesticides, and antigens. In this paper we investigated two aspects of applying support vector machine (SVM), a promising machine learning method for bioinformatics, to prediction and analysis of beta-turns and gamma-turns. First, we developed two SVM-based methods, called BTSVM and GTSVM, which predict beta-turns and gamma-turns in a protein from its sequence. When compared with other methods, BTSVM has a superior performance and GTSVM is competitive. Second, we used SVMs with a linear kernel to estimate the support of amino acids for the formation of beta-turns and gamma-turns depending on their position in a protein. Our analysis results are more comprehensive and easier to use than the previous results in designing turns in proteins.

Facility target insert shielding assessment

Energy Technology Data Exchange (ETDEWEB)

Mocko, Michal [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2015-10-06

Main objective of this report is to assess the basic shielding requirements for the vertical target insert and retrieval port. We used the baseline design for the vertical target insert in our calculations. The insert sits in the 12”-diameter cylindrical shaft extending from the service alley in the top floor of the facility all the way down to the target location. The target retrieval mechanism is a long rod with the target assembly attached and running the entire length of the vertical shaft. The insert also houses the helium cooling supply and return lines each with 2” diameter. In the present study we focused on calculating the neutron and photon dose rate fields on top of the target insert/retrieval mechanism in the service alley. Additionally, we studied a few prototypical configurations of the shielding layers in the vertical insert as well as on the top.

Development of Simultaneous Beta-and-Coincidence-Gamma Imager for Plant Imaging Research

Energy Technology Data Exchange (ETDEWEB)

Tai, Yuan-Chuan [Washington Univ., St. Louis, MO (United States). School of Medicine

2016-09-30

The goal of this project is to develop a novel imaging system that can simultaneously acquire beta and coincidence gamma images of positron sources in thin objects such as leaves of plants. This hybrid imager can be used to measure carbon assimilation in plants quantitatively and in real-time after C-11 labeled carbon-dioxide is administered. A better understanding of carbon assimilation, particularly under the increasingly elevated atmospheric CO₂ level, is extremely critical for plant scientists who study food crop and biofuel production. Phase 1 of this project is focused on the technology development with 3 specific aims: (1) develop a hybrid detector that can detect beta and gamma rays simultaneously; (2) develop an imaging system that can differentiate these two types of radiation and acquire beta and coincidence gamma images in real-time; (3) develop techniques to quantify radiotracer distribution using beta and gamma images. Phase 2 of this project is to apply technologies developed in phase 1 to study plants using positron-emitting radionuclide such as ¹¹C to study carbon assimilation in biofuel plants.

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.

Accelerator shield design of KIPT neutron source facility

International Nuclear Information System (INIS)

Zhong, Z.; Gohar, Y.

2013-01-01

Argonne National Laboratory (ANL) of the United States and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the design development of a neutron source facility at KIPT utilizing an electron-accelerator-driven subcritical assembly. Electron beam power is 100 kW, using 100 MeV electrons. The facility is designed to perform basic and applied nuclear research, produce medical isotopes, and train young nuclear specialists. The biological shield of the accelerator building is designed to reduce the biological dose to less than 0.5-mrem/hr during operation. The main source of the biological dose is the photons and the neutrons generated by interactions of leaked electrons from the electron gun and accelerator sections with the surrounding concrete and accelerator materials. The Monte Carlo code MCNPX serves as the calculation tool for the shield design, due to its capability to transport electrons, photons, and neutrons coupled problems. The direct photon dose can be tallied by MCNPX calculation, starting with the leaked electrons. However, it is difficult to accurately tally the neutron dose directly from the leaked electrons. The neutron yield per electron from the interactions with the surrounding components is less than 0.01 neutron per electron. This causes difficulties for Monte Carlo analyses and consumes tremendous computation time for tallying with acceptable statistics the neutron dose outside the shield boundary. To avoid these difficulties, the SOURCE and TALLYX user subroutines of MCNPX were developed for the study. The generated neutrons are banked, together with all related parameters, for a subsequent MCNPX calculation to obtain the neutron and secondary photon doses. The weight windows variance reduction technique is utilized for both neutron and photon dose calculations. Two shielding materials, i.e., heavy concrete and ordinary concrete, were considered for the shield design. The main goal is to maintain the total

Shielding for a tandem accelerator coupled to linac booster

International Nuclear Information System (INIS)

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

1996-01-01

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

Quantitative portable gamma spectroscopy sample analysis for non-standard sample geometries

International Nuclear Information System (INIS)

Enghauser, M.W.; Ebara, S.B.

1997-01-01

Utilizing a portable spectroscopy system, a quantitative method for analysis of samples containing a mixture of fission and activation products in nonstandard geometries was developed. The method can be used with various sample and shielding configurations where analysis on a laboratory based gamma spectroscopy system is impractical. The portable gamma spectroscopy method involves calibration of the detector and modeling of the sample and shielding to identify and quantify the radionuclides present in the sample. The method utilizes the intrinsic efficiency of the detector and the unattenuated gamma fluence rate at the detector surface per unit activity from the sample to calculate the nuclide activity and Minimum Detectable Activity (MDA). For a complex geometry, a computer code written for shielding applications (MICROSHIELD) is utilized to determine the unattenuated gamma fluence rate per unit activity at the detector surface. Lastly, the method is only applicable to nuclides which emit gamma rays and cannot be used for pure beta emitters. In addition, if sample self absorption and shielding is significant, the attenuation will result in high MDA's for nuclides which solely emit low energy gamma rays. The following presents the analysis technique and presents verification results demonstrating the accuracy of the method

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

Determination of beta and gamma radioactivity on the coal

International Nuclear Information System (INIS)

Suhardi; Mulyono; Sutanto WW; Rosidi

2013-01-01

Radioactivity in coal, botton ash and fly ash. This determination was carried out to know each the radioactivity of β gross, γ gross on the coal, botton ash and fly ash, which accommodation of environment data the present in PLTU Paiton Probolinggo. Samples taken preparation and analysis based on the procedures of environmental radioactivity analysis. The radioactivity on the PLTU Paiton Probolinggo detected by beta gross Spectrometer with Geiger Muller (GM) and gamma Spectrometer with Ge(Li) detector. The result indicates that radioactivity concentration of beta and gamma gross are (95,57-308,26) Bq/kg and (1,374 - 31,677) Bq/kg respectively. (author)

Early test facilities and analytic methods for radiation shielding: Proceedings

International Nuclear Information System (INIS)

Ingersoll, D.T.; Ingersoll, J.K.

1992-11-01

This report represents a compilation of eight papers presented at the 1992 American Nuclear Society/European Nuclear Society International Meeting. The meeting is of special significance since it commemorates the fiftieth anniversary of the first controlled nuclear chain reaction. The papers contained in this report were presented in a special session organized by the Radiation Protection and Shielding Division in keeping with the historical theme of the meeting. The paper titles are good indicators of their content and are: (1) The origin of radiation shielding research: The Oak Ridge experience, (2) Shielding research at the hanford site, (3) Aircraft shielding experiments at General Dynamics Fort Worth, 1950-1962, (4) Where have the neutrons gone?, a history of the tower shielding facility, (5) History and evolution of buildup factors, (6) Early shielding research at Bettis atomic power laboratory, (7) UK reactor shielding: then and now, (8) A very personal view of the development of radiation shielding theory

Optimization of plastic scintillator thicknesses for online beta/gamma detection

Directory of Open Access Journals (Sweden)

Pourtangestani K.

2012-04-01

Full Text Available For efficient beta detection in a mixed beta gamma field, Monte Carlo simulation models have been built to optimize the thickness of a plastic scintillator, used in a whole body monitor. The simulation has been performed using the MCNP/X code for different thicknesses of plastic scintillator from 150 μm to 600 μm. The relationship between the thickness of the scintillator and the efficiency of the detector has been analyzed. For 150 μm thickness, an experimental investigation has been conducted with different beta sources at different positions on the scintillator and the counting efficiency of the unit has been measured. Evaluated data along with experimental ones have been discussed. A thickness of 300 μm to 500 μm has been found to be the optimum thickness for high efficiency beta detection in the presence of low energy gamma-rays.

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)

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-02-15

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

Determination of $\\gamma$ and $-2\\beta_s$ from charmless two-body decays of beauty mesons

CERN Document Server

Aaij, Roel; Adeva, Bernardo; Adinolfi, Marco; Affolder, Anthony; Ajaltouni, Ziad; Akar, Simon; Albrecht, Johannes; Alessio, Federico; Alexander, Michael; Ali, Suvayu; Alkhazov, Georgy; Alvarez Cartelle, Paula; Alves Jr, Antonio Augusto; Amato, Sandra; Amerio, Silvia; Amhis, Yasmine; An, Liupan; Anderlini, Lucio; Anderson, Jonathan; Andreassen, Rolf; Andreotti, Mirco; Andrews, Jason; Appleby, Robert; Aquines Gutierrez, Osvaldo; Archilli, Flavio; Artamonov, Alexander; Artuso, Marina; Aslanides, Elie; Auriemma, Giulio; Baalouch, Marouen; Bachmann, Sebastian; Back, John; Badalov, Alexey; Baesso, Clarissa; Baldini, Wander; Barlow, Roger; Barschel, Colin; Barsuk, Sergey; Barter, William; Batozskaya, Varvara; Battista, Vincenzo; Bay, Aurelio; Beaucourt, Leo; Beddow, John; Bedeschi, Franco; Bediaga, Ignacio; Belogurov, Sergey; Belous, Konstantin; Belyaev, Ivan; Ben-Haim, Eli; Bencivenni, Giovanni; Benson, Sean; Benton, Jack; Berezhnoy, Alexander; Bernet, Roland; Bettler, Marc-Olivier; van Beuzekom, Martinus; Bien, Alexander; Bifani, Simone; Bird, Thomas; Bizzeti, Andrea; Bjørnstad, Pål Marius; Blake, Thomas; Blanc, Frédéric; Blouw, Johan; Blusk, Steven; Bocci, Valerio; Bondar, Alexander; Bondar, Nikolay; Bonivento, Walter; Borghi, Silvia; Borgia, Alessandra; Borsato, Martino; Bowcock, Themistocles; Bowen, Espen Eie; Bozzi, Concezio; Brambach, Tobias; Bressieux, Joël; Brett, David; Britsch, Markward; Britton, Thomas; Brodzicka, Jolanta; Brook, Nicholas; Brown, Henry; Bursche, Albert; Busetto, Giovanni; Buytaert, Jan; Cadeddu, Sandro; Calabrese, Roberto; Calvi, Marta; Calvo Gomez, Miriam; Campana, Pierluigi; Campora Perez, Daniel; Carbone, Angelo; Carboni, Giovanni; Cardinale, Roberta; Cardini, Alessandro; Carson, Laurence; Carvalho Akiba, Kazuyoshi; Casse, Gianluigi; Cassina, Lorenzo; Castillo Garcia, Lucia; Cattaneo, Marco; Cauet, Christophe; Cenci, Riccardo; Charles, Matthew; Charpentier, Philippe; Chefdeville, Maximilien; Chen, Shanzhen; Cheung, Shu-Faye; Chiapolini, Nicola; Chrzaszcz, Marcin; Ciba, Krzystof; Cid Vidal, Xabier; Ciezarek, Gregory; Clarke, Peter; Clemencic, Marco; Cliff, Harry; Closier, Joel; Coco, Victor; Cogan, Julien; Cogneras, Eric; Cojocariu, Lucian; Collins, Paula; Comerma-Montells, Albert; Contu, Andrea; Cook, Andrew; Coombes, Matthew; Coquereau, Samuel; Corti, Gloria; Corvo, Marco; Counts, Ian; Couturier, Benjamin; Cowan, Greig; Craik, Daniel Charles; Cruz Torres, Melissa Maria; Cunliffe, Samuel; Currie, Robert; D'Ambrosio, Carmelo; Dalseno, Jeremy; David, Pascal; David, Pieter; Davis, Adam; De Bruyn, Kristof; De Capua, Stefano; De Cian, Michel; De Miranda, Jussara; De Paula, Leandro; De Silva, Weeraddana; De Simone, Patrizia; Decamp, Daniel; Deckenhoff, Mirko; Del Buono, Luigi; Déléage, Nicolas; Derkach, Denis; Deschamps, Olivier; Dettori, Francesco; Di Canto, Angelo; Dijkstra, Hans; Donleavy, Stephanie; Dordei, Francesca; Dorigo, Mirco; Dosil Suárez, Alvaro; Dossett, David; Dovbnya, Anatoliy; Dreimanis, Karlis; Dujany, Giulio; Dupertuis, Frederic; Durante, Paolo; Dzhelyadin, Rustem; Dziurda, Agnieszka; Dzyuba, Alexey; Easo, Sajan; Egorychev, Victor; Eidelman, Semen; Eisenhardt, Stephan; Eitschberger, Ulrich; Ekelhof, Robert; Eklund, Lars; El Rifai, Ibrahim; Elsasser, Christian; Ely, Scott; Esen, Sevda; Evans, Hannah Mary; Evans, Timothy; Falabella, Antonio; Färber, Christian; Farinelli, Chiara; Farley, Nathanael; Farry, Stephen; Fay, Robert; Ferguson, Dianne; Fernandez Albor, Victor; Ferreira Rodrigues, Fernando; Ferro-Luzzi, Massimiliano; Filippov, Sergey; Fiore, Marco; Fiorini, Massimiliano; Firlej, Miroslaw; Fitzpatrick, Conor; Fiutowski, Tomasz; Fol, Philip; Fontana, Marianna; Fontanelli, Flavio; Forty, Roger; Francisco, Oscar; Frank, Markus; Frei, Christoph; Frosini, Maddalena; Fu, Jinlin; Furfaro, Emiliano; Gallas Torreira, Abraham; Galli, Domenico; Gallorini, Stefano; Gambetta, Silvia; Gandelman, Miriam; Gandini, Paolo; Gao, Yuanning; García Pardiñas, Julián; Garofoli, Justin; Garra Tico, Jordi; Garrido, Lluis; Gaspar, Clara; Gauld, Rhorry; Gavardi, Laura; Gavrilov, Gennadii; Geraci, Angelo; Gersabeck, Evelina; Gersabeck, Marco; Gershon, Timothy; Ghez, Philippe; Gianelle, Alessio; Gianì, Sebastiana; Gibson, Valerie; Giubega, Lavinia-Helena; Gligorov, Vladimir; Göbel, Carla; Golubkov, Dmitry; Golutvin, Andrey; Gomes, Alvaro; Gotti, Claudio; Grabalosa Gándara, Marc; Graciani Diaz, Ricardo; Granado Cardoso, Luis Alberto; Graugés, Eugeni; Graziani, Giacomo; Grecu, Alexandru; Greening, Edward; Gregson, Sam; Griffith, Peter; Grillo, Lucia; Grünberg, Oliver; Gui, Bin; Gushchin, Evgeny; Guz, Yury; Gys, Thierry; Hadjivasiliou, Christos; Haefeli, Guido; Haen, Christophe; Haines, Susan; Hall, Samuel; Hamilton, Brian; Hampson, Thomas; Han, Xiaoxue; Hansmann-Menzemer, Stephanie; Harnew, Neville; Harnew, Samuel; Harrison, Jonathan; He, Jibo; Head, Timothy; Heijne, Veerle; Hennessy, Karol; Henrard, Pierre; Henry, Louis; Hernando Morata, Jose Angel; van Herwijnen, Eric; Heß, Miriam; Hicheur, Adlène; Hill, Donal; Hoballah, Mostafa; Hombach, Christoph; Hulsbergen, Wouter; Hunt, Philip; Hussain, Nazim; Hutchcroft, David; Hynds, Daniel; Idzik, Marek; Ilten, Philip; Jacobsson, Richard; Jaeger, Andreas; Jalocha, Pawel; Jans, Eddy; Jaton, Pierre; Jawahery, Abolhassan; Jing, Fanfan; John, Malcolm; Johnson, Daniel; Jones, Christopher; Joram, Christian; Jost, Beat; Jurik, Nathan; Kaballo, Michael; Kandybei, Sergii; Kanso, Walaa; Karacson, Matthias; Karbach, Moritz; Karodia, Sarah; Kelsey, Matthew; Kenyon, Ian; Ketel, Tjeerd; Khanji, Basem; Khurewathanakul, Chitsanu; Klaver, Suzanne; Klimaszewski, Konrad; Kochebina, Olga; Kolpin, Michael; Komarov, Ilya; Koopman, Rose; Koppenburg, Patrick; Korolev, Mikhail; Kozlinskiy, Alexandr; Kravchuk, Leonid; Kreplin, Katharina; Kreps, Michal; Krocker, Georg; Krokovny, Pavel; Kruse, Florian; Kucewicz, Wojciech; Kucharczyk, Marcin; Kudryavtsev, Vasily; Kurek, Krzysztof; Kvaratskheliya, Tengiz; La Thi, Viet Nga; Lacarrere, Daniel; Lafferty, George; Lai, Adriano; Lambert, Dean; Lambert, Robert W; Lanfranchi, Gaia; Langenbruch, Christoph; Langhans, Benedikt; Latham, Thomas; Lazzeroni, Cristina; Le Gac, Renaud; van Leerdam, Jeroen; Lees, Jean-Pierre; Lefèvre, Regis; Leflat, Alexander; Lefrançois, Jacques; Leo, Sabato; Leroy, Olivier; Lesiak, Tadeusz; Leverington, Blake; Li, Yiming; Likhomanenko, Tatiana; Liles, Myfanwy; Lindner, Rolf; Linn, Christian; Lionetto, Federica; Liu, Bo; Lohn, Stefan; Longstaff, Iain; Lopes, Jose; Lopez-March, Neus; Lowdon, Peter; Lu, Haiting; Lucchesi, Donatella; Luo, Haofei; Lupato, Anna; Luppi, Eleonora; Lupton, Oliver; Machefert, Frederic; Machikhiliyan, Irina V; Maciuc, Florin; Maev, Oleg; Malde, Sneha; Malinin, Alexander; Manca, Giulia; Mancinelli, Giampiero; Mapelli, Alessandro; Maratas, Jan; Marchand, Jean François; Marconi, Umberto; Marin Benito, Carla; Marino, Pietro; Märki, Raphael; Marks, Jörg; Martellotti, Giuseppe; Martens, Aurelien; Martín Sánchez, Alexandra; Martinelli, Maurizio; Martinez Santos, Diego; Martinez Vidal, Fernando; Martins Tostes, Danielle; Massafferri, André; Matev, Rosen; Mathe, Zoltan; Matteuzzi, Clara; Mazurov, Alexander; McCann, Michael; McCarthy, James; McNab, Andrew; McNulty, Ronan; McSkelly, Ben; Meadows, Brian; Meier, Frank; Meissner, Marco; Merk, Marcel; Milanes, Diego Alejandro; Minard, Marie-Noelle; Moggi, Niccolò; Molina Rodriguez, Josue; Monteil, Stephane; Morandin, Mauro; Morawski, Piotr; Mordà, Alessandro; Morello, Michael Joseph; Moron, Jakub; Morris, Adam Benjamin; Mountain, Raymond; Muheim, Franz; Müller, Katharina; Mussini, Manuel; Muster, Bastien; Naik, Paras; Nakada, Tatsuya; Nandakumar, Raja; Nasteva, Irina; Needham, Matthew; Neri, Nicola; Neubert, Sebastian; Neufeld, Niko; Neuner, Max; Nguyen, Anh Duc; Nguyen, Thi-Dung; Nguyen-Mau, Chung; Nicol, Michelle; Niess, Valentin; Niet, Ramon; Nikitin, Nikolay; Nikodem, Thomas; Novoselov, Alexey; O'Hanlon, Daniel Patrick; Oblakowska-Mucha, Agnieszka; Obraztsov, Vladimir; Oggero, Serena; Ogilvy, Stephen; Okhrimenko, Oleksandr; Oldeman, Rudolf; Onderwater, Gerco; Orlandea, Marius; Otalora Goicochea, Juan Martin; Owen, Patrick; Oyanguren, Maria Arantza; Pal, Bilas Kanti; Palano, Antimo; Palombo, Fernando; Palutan, Matteo; Panman, Jacob; Papanestis, Antonios; Pappagallo, Marco; Pappalardo, Luciano; Parkes, Christopher; Parkinson, Christopher John; Passaleva, Giovanni; Patel, Girish; Patel, Mitesh; Patrignani, Claudia; Pazos Alvarez, Antonio; Pearce, Alex; Pellegrino, Antonio; Pepe Altarelli, Monica; Perazzini, Stefano; Perez Trigo, Eliseo; Perret, Pascal; Perrin-Terrin, Mathieu; Pescatore, Luca; Pesen, Erhan; Petridis, Konstantin; Petrolini, Alessandro; Picatoste Olloqui, Eduardo; Pietrzyk, Boleslaw; Pilař, Tomas; Pinci, Davide; Pistone, Alessandro; Playfer, Stephen; Plo Casasus, Maximo; Polci, Francesco; Poluektov, Anton; Polycarpo, Erica; Popov, Alexander; Popov, Dmitry; Popovici, Bogdan; Potterat, Cédric; Price, Eugenia; Price, Joseph David; Prisciandaro, Jessica; Pritchard, Adrian; Prouve, Claire; Pugatch, Valery; Puig Navarro, Albert; Punzi, Giovanni; Qian, Wenbin; Rachwal, Bartolomiej; Rademacker, Jonas; Rakotomiaramanana, Barinjaka; Rama, Matteo; Rangel, Murilo; Raniuk, Iurii; Rauschmayr, Nathalie; Raven, Gerhard; Redi, Federico; Reichert, Stefanie; Reid, Matthew; dos Reis, Alberto; Ricciardi, Stefania; Richards, Sophie; Rihl, Mariana; Rinnert, Kurt; Rives Molina, Vincente; Robbe, Patrick; Rodrigues, Ana Barbara; Rodrigues, Eduardo; Rodriguez Perez, Pablo; Roiser, Stefan; Romanovsky, Vladimir; Romero Vidal, Antonio; Rotondo, Marcello; Rouvinet, Julien; Ruf, Thomas; Ruiz, Hugo; Ruiz Valls, Pablo; Saborido Silva, Juan Jose; Sagidova, Naylya; Sail, Paul; Saitta, Biagio; Salustino Guimaraes, Valdir; Sanchez Mayordomo, Carlos; Sanmartin Sedes, Brais; Santacesaria, Roberta; Santamarina Rios, Cibran; Santovetti, Emanuele; Sarti, Alessio; Satriano, Celestina; Satta, Alessia; Saunders, Daniel Martin; Savrie, Mauro; Savrina, Darya; Schiller, Manuel; Schindler, Heinrich; Schlupp, Maximilian; Schmelling, Michael; Schmidt, Burkhard; Schneider, Olivier; Schopper, Andreas; Schune, Marie Helene; Schwemmer, Rainer; Sciascia, Barbara; Sciubba, Adalberto; Seco, Marcos; Semennikov, Alexander; Sepp, Indrek; Serra, Nicola; Serrano, Justine; Sestini, Lorenzo; Seyfert, Paul; Shapkin, Mikhail; Shapoval, Illya; Shcheglov, Yury; Shears, Tara; Shekhtman, Lev; Shevchenko, Vladimir; Shires, Alexander; Silva Coutinho, Rafael; Simi, Gabriele; Sirendi, Marek; Skidmore, Nicola; Skwarnicki, Tomasz; Smith, Anthony; Smith, Edmund; Smith, Eluned; Smith, Jackson; Smith, Mark; Snoek, Hella; Sokoloff, Michael; Soler, Paul; Soomro, Fatima; Souza, Daniel; Souza De Paula, Bruno; Spaan, Bernhard; Sparkes, Ailsa; Spradlin, Patrick; Sridharan, Srikanth; Stagni, Federico; Stahl, Marian; Stahl, Sascha; Steinkamp, Olaf; Stenyakin, Oleg; Stevenson, Scott; Stoica, Sabin; Stone, Sheldon; Storaci, Barbara; Stracka, Simone; Straticiuc, Mihai; Straumann, Ulrich; Stroili, Roberto; Subbiah, Vijay Kartik; Sun, Liang; Sutcliffe, William; Swientek, Krzysztof; Swientek, Stefan; Syropoulos, Vasileios; Szczekowski, Marek; Szczypka, Paul; Szilard, Daniela; Szumlak, Tomasz; T'Jampens, Stephane; Teklishyn, Maksym; Tellarini, Giulia; Teubert, Frederic; Thomas, Christopher; Thomas, Eric; van Tilburg, Jeroen; Tisserand, Vincent; Tobin, Mark; Tolk, Siim; Tomassetti, Luca; Topp-Joergensen, Stig; Torr, Nicholas; Tournefier, Edwige; Tourneur, Stephane; Tran, Minh Tâm; Tresch, Marco; Tsaregorodtsev, Andrei; Tsopelas, Panagiotis; Tuning, Niels; Ubeda Garcia, Mario; Ukleja, Artur; Ustyuzhanin, Andrey; Uwer, Ulrich; Vacca, Claudia; Vagnoni, Vincenzo; Valenti, Giovanni; Vallier, Alexis; Vazquez Gomez, Ricardo; Vazquez Regueiro, Pablo; Vázquez Sierra, Carlos; Vecchi, Stefania; Velthuis, Jaap; Veltri, Michele; Veneziano, Giovanni; Vesterinen, Mika; Viaud, Benoit; Vieira, Daniel; Vieites Diaz, Maria; Vilasis-Cardona, Xavier; Vollhardt, Achim; Volyanskyy, Dmytro; Voong, David; Vorobyev, Alexey; Vorobyev, Vitaly; Voß, Christian; Voss, Helge; de Vries, Jacco; Waldi, Roland; Wallace, Charlotte; Wallace, Ronan; Walsh, John; Wandernoth, Sebastian; Wang, Jianchun; Ward, David; Watson, Nigel; Websdale, David; Whitehead, Mark; Wicht, Jean; Wiedner, Dirk; Wilkinson, Guy; Williams, Matthew; Williams, Mike; Wilschut, Hans; Wilson, Fergus; Wimberley, Jack; Wishahi, Julian; Wislicki, Wojciech; Witek, Mariusz; Wormser, Guy; Wotton, Stephen; Wright, Simon; Wyllie, Kenneth; Xie, Yuehong; Xing, Zhou; Xu, Zhirui; Yang, Zhenwei; Yuan, Xuhao; Yushchenko, Oleg; Zangoli, Maria; Zavertyaev, Mikhail; Zhang, Liming; Zhang, Wen Chao; Zhang, Yanxi; Zhelezov, Alexey; Zhokhov, Anatoly; Zhong, Liang; Zvyagin, Alexander

2015-01-01

Using the latest LHCb measurements of time-dependent $C\\!P$ violation in the $B^0_s \\to K^+K^-$ decay, a U-spin relation between the decay amplitudes of $B^0_s \\to K^+K^-$ and $B^0\\to \\pi^+\\pi^-$ decay processes allows constraints to be placed on the angle $\\gamma$ of the unitarity triangle and on the $B^0_s$ mixing phase $-2\\beta_s$. Results from an extended approach, which uses additional inputs on $B^0\\to \\pi^0\\pi^0$ and $B^+\\to \\pi^+\\pi^0$ decays from other experiments and exploits isospin symmetry, are also presented. The dependence of the results on the maximum allowed amount of U-spin breaking is studied. At 68% probability, the value $\\gamma = \\left( 63.5^{\\,+\\, 7.2}_{\\,-\\,6.7} \\right)^\\circ~\\mathrm{modulo}~180^\\circ$ is determined. In an alternative analysis, the value $-2\\beta_s = -0.12 ^{\\,+\\,0.14}_{\\,-\\,0.16}\\,\\,\\mathrm{rad}$ is found. In both measurements, the uncertainties due to U-spin breaking effects up to 50% are included.

Determination of $\\gamma$ and $-2\\beta_s$ from charmless two-body decays of beauty mesons

CERN Document Server

Aaij, Roel; Adeva, Bernardo; Adinolfi, Marco; Affolder, Anthony; Ajaltouni, Ziad; Akar, Simon; Albrecht, Johannes; Alessio, Federico; Alexander, Michael; Ali, Suvayu; Alkhazov, Georgy; Alvarez Cartelle, Paula; Alves Jr, Antonio Augusto; Amato, Sandra; Amerio, Silvia; Amhis, Yasmine; An, Liupan; Anderlini, Lucio; Anderson, Jonathan; Andreassen, Rolf; Andreotti, Mirco; Andrews, Jason; Appleby, Robert; Aquines Gutierrez, Osvaldo; Archilli, Flavio; Artamonov, Alexander; Artuso, Marina; Aslanides, Elie; Auriemma, Giulio; Baalouch, Marouen; Bachmann, Sebastian; Back, John; Badalov, Alexey; Baesso, Clarissa; Baldini, Wander; Barlow, Roger; Barschel, Colin; Barsuk, Sergey; Barter, William; Batozskaya, Varvara; Battista, Vincenzo; Bay, Aurelio; Beaucourt, Leo; Beddow, John; Bedeschi, Franco; Bediaga, Ignacio; Belogurov, Sergey; Belous, Konstantin; Belyaev, Ivan; Ben-Haim, Eli; Bencivenni, Giovanni; Benson, Sean; Benton, Jack; Berezhnoy, Alexander; Bernet, Roland; Bettler, Marc-Olivier; van Beuzekom, Martinus; Bien, Alexander; Bifani, Simone; Bird, Thomas; Bizzeti, Andrea; Bjørnstad, Pål Marius; Blake, Thomas; Blanc, Frédéric; Blouw, Johan; Blusk, Steven; Bocci, Valerio; Bondar, Alexander; Bondar, Nikolay; Bonivento, Walter; Borghi, Silvia; Borgia, Alessandra; Borsato, Martino; Bowcock, Themistocles; Bowen, Espen Eie; Bozzi, Concezio; Brambach, Tobias; Bressieux, Joël; Brett, David; Britsch, Markward; Britton, Thomas; Brodzicka, Jolanta; Brook, Nicholas; Brown, Henry; Bursche, Albert; Busetto, Giovanni; Buytaert, Jan; Cadeddu, Sandro; Calabrese, Roberto; Calvi, Marta; Calvo Gomez, Miriam; Campana, Pierluigi; Campora Perez, Daniel; Carbone, Angelo; Carboni, Giovanni; Cardinale, Roberta; Cardini, Alessandro; Carson, Laurence; Carvalho Akiba, Kazuyoshi; Casse, Gianluigi; Cassina, Lorenzo; Castillo Garcia, Lucia; Cattaneo, Marco; Cauet, Christophe; Cenci, Riccardo; Charles, Matthew; Charpentier, Philippe; Chefdeville, Maximilien; Chen, Shanzhen; Cheung, Shu-Faye; Chiapolini, Nicola; Chrzaszcz, Marcin; 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Dzhelyadin, Rustem; Dziurda, Agnieszka; Dzyuba, Alexey; Easo, Sajan; Egorychev, Victor; Eidelman, Semen; Eisenhardt, Stephan; Eitschberger, Ulrich; Ekelhof, Robert; Eklund, Lars; El Rifai, Ibrahim; Elsasser, Christian; Ely, Scott; Esen, Sevda; Evans, Hannah Mary; Evans, Timothy; Falabella, Antonio; Färber, Christian; Farinelli, Chiara; Farley, Nathanael; Farry, Stephen; Fay, Robert; Ferguson, Dianne; Fernandez Albor, Victor; Ferreira Rodrigues, Fernando; Ferro-Luzzi, Massimiliano; Filippov, Sergey; Fiore, Marco; Fiorini, Massimiliano; Firlej, Miroslaw; Fitzpatrick, Conor; Fiutowski, Tomasz; Fol, Philip; Fontana, Marianna; Fontanelli, Flavio; Forty, Roger; Francisco, Oscar; Frank, Markus; Frei, Christoph; Frosini, Maddalena; Fu, Jinlin; Furfaro, Emiliano; Gallas Torreira, Abraham; Galli, Domenico; Gallorini, Stefano; Gambetta, Silvia; Gandelman, Miriam; Gandini, Paolo; Gao, Yuanning; García Pardiñas, Julián; Garofoli, Justin; Garra Tico, Jordi; Garrido, Lluis; Gaspar, Clara; Gauld, Rhorry; 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Hulsbergen, Wouter; Hunt, Philip; Hussain, Nazim; Hutchcroft, David; Hynds, Daniel; Idzik, Marek; Ilten, Philip; Jacobsson, Richard; Jaeger, Andreas; Jalocha, Pawel; Jans, Eddy; Jaton, Pierre; Jawahery, Abolhassan; Jing, Fanfan; John, Malcolm; Johnson, Daniel; Jones, Christopher; Joram, Christian; Jost, Beat; Jurik, Nathan; Kaballo, Michael; Kandybei, Sergii; Kanso, Walaa; Karacson, Matthias; Karbach, Moritz; Karodia, Sarah; Kelsey, Matthew; Kenyon, Ian; Ketel, Tjeerd; Khanji, Basem; Khurewathanakul, Chitsanu; Klaver, Suzanne; Klimaszewski, Konrad; Kochebina, Olga; Kolpin, Michael; Komarov, Ilya; Koopman, Rose; Koppenburg, Patrick; Korolev, Mikhail; Kozlinskiy, Alexandr; Kravchuk, Leonid; Kreplin, Katharina; Kreps, Michal; Krocker, Georg; Krokovny, Pavel; Kruse, Florian; Kucewicz, Wojciech; Kucharczyk, Marcin; Kudryavtsev, Vasily; Kurek, Krzysztof; Kvaratskheliya, Tengiz; La Thi, Viet Nga; Lacarrere, Daniel; Lafferty, George; Lai, Adriano; Lambert, Dean; Lambert, Robert W; Lanfranchi, Gaia; Langenbruch, Christoph; Langhans, Benedikt; Latham, Thomas; Lazzeroni, Cristina; Le Gac, Renaud; van Leerdam, Jeroen; Lees, Jean-Pierre; Lefèvre, Regis; Leflat, Alexander; Lefrançois, Jacques; Leo, Sabato; Leroy, Olivier; Lesiak, Tadeusz; Leverington, Blake; Li, Yiming; Likhomanenko, Tatiana; Liles, Myfanwy; Lindner, Rolf; Linn, Christian; Lionetto, Federica; Liu, Bo; Lohn, Stefan; Longstaff, Iain; Lopes, Jose; Lopez-March, Neus; Lowdon, Peter; Lu, Haiting; Lucchesi, Donatella; Luo, Haofei; Lupato, Anna; Luppi, Eleonora; Lupton, Oliver; Machefert, Frederic; Machikhiliyan, Irina V; Maciuc, Florin; Maev, Oleg; Malde, Sneha; Malinin, Alexander; Manca, Giulia; Mancinelli, Giampiero; Mapelli, Alessandro; Maratas, Jan; Marchand, Jean François; Marconi, Umberto; Marin Benito, Carla; Marino, Pietro; Märki, Raphael; Marks, Jörg; Martellotti, Giuseppe; Martens, Aurelien; Martín Sánchez, Alexandra; Martinelli, Maurizio; Martinez Santos, Diego; Martinez Vidal, Fernando; Martins Tostes, Danielle; Massafferri, André; Matev, Rosen; Mathe, Zoltan; Matteuzzi, Clara; Mazurov, Alexander; McCann, Michael; McCarthy, James; McNab, Andrew; McNulty, Ronan; McSkelly, Ben; Meadows, Brian; Meier, Frank; Meissner, Marco; Merk, Marcel; Milanes, Diego Alejandro; Minard, Marie-Noelle; Moggi, Niccolò; Molina Rodriguez, Josue; Monteil, Stephane; Morandin, Mauro; Morawski, Piotr; Mordà, Alessandro; Morello, Michael Joseph; Moron, Jakub; Morris, Adam Benjamin; Mountain, Raymond; Muheim, Franz; Müller, Katharina; Mussini, Manuel; Muster, Bastien; Naik, Paras; Nakada, Tatsuya; Nandakumar, Raja; Nasteva, Irina; Needham, Matthew; Neri, Nicola; Neubert, Sebastian; Neufeld, Niko; Neuner, Max; Nguyen, Anh Duc; Nguyen, Thi-Dung; Nguyen-Mau, Chung; Nicol, Michelle; Niess, Valentin; Niet, Ramon; Nikitin, Nikolay; Nikodem, Thomas; Novoselov, Alexey; O'Hanlon, Daniel Patrick; Oblakowska-Mucha, Agnieszka; Obraztsov, Vladimir; Oggero, Serena; Ogilvy, Stephen; Okhrimenko, Oleksandr; Oldeman, Rudolf; Onderwater, Gerco; Orlandea, Marius; Otalora Goicochea, Juan Martin; Owen, Patrick; Oyanguren, Maria Arantza; Pal, Bilas Kanti; Palano, Antimo; Palombo, Fernando; Palutan, Matteo; Panman, Jacob; Papanestis, Antonios; Pappagallo, Marco; Pappalardo, Luciano; Parkes, Christopher; Parkinson, Christopher John; Passaleva, Giovanni; Patel, Girish; Patel, Mitesh; Patrignani, Claudia; Pazos Alvarez, Antonio; Pearce, Alex; Pellegrino, Antonio; Pepe Altarelli, Monica; Perazzini, Stefano; Perez Trigo, Eliseo; Perret, Pascal; Perrin-Terrin, Mathieu; Pescatore, Luca; Pesen, Erhan; Petridis, Konstantin; Petrolini, Alessandro; Picatoste Olloqui, Eduardo; Pietrzyk, Boleslaw; Pilař, Tomas; Pinci, Davide; Pistone, Alessandro; Playfer, Stephen; Plo Casasus, Maximo; Polci, Francesco; Poluektov, Anton; Polycarpo, Erica; Popov, Alexander; Popov, Dmitry; Popovici, Bogdan; Potterat, Cédric; Price, Eugenia; Price, Joseph David; Prisciandaro, Jessica; Pritchard, Adrian; Prouve, Claire; Pugatch, Valery; Puig Navarro, Albert; Punzi, Giovanni; Qian, Wenbin; Rachwal, Bartolomiej; Rademacker, Jonas; Rakotomiaramanana, Barinjaka; Rama, Matteo; Rangel, Murilo; Raniuk, Iurii; Rauschmayr, Nathalie; Raven, Gerhard; Redi, Federico; Reichert, Stefanie; Reid, Matthew; dos Reis, Alberto; Ricciardi, Stefania; Richards, Sophie; Rihl, Mariana; Rinnert, Kurt; Rives Molina, Vincente; Robbe, Patrick; Rodrigues, Ana Barbara; Rodrigues, Eduardo; Rodriguez Perez, Pablo; Roiser, Stefan; Romanovsky, Vladimir; Romero Vidal, Antonio; Rotondo, Marcello; Rouvinet, Julien; Ruf, Thomas; Ruiz, Hugo; Ruiz Valls, Pablo; Saborido Silva, Juan Jose; Sagidova, Naylya; Sail, Paul; Saitta, Biagio; Salustino Guimaraes, Valdir; Sanchez Mayordomo, Carlos; Sanmartin Sedes, Brais; Santacesaria, Roberta; Santamarina Rios, Cibran; Santovetti, Emanuele; Sarti, Alessio; Satriano, Celestina; Satta, Alessia; Saunders, Daniel Martin; Savrie, Mauro; Savrina, Darya; Schiller, Manuel; Schindler, Heinrich; Schlupp, Maximilian; Schmelling, Michael; Schmidt, Burkhard; Schneider, Olivier; Schopper, Andreas; Schune, Marie Helene; Schwemmer, Rainer; Sciascia, Barbara; Sciubba, Adalberto; Seco, Marcos; Semennikov, Alexander; Sepp, Indrek; Serra, Nicola; Serrano, Justine; Sestini, Lorenzo; Seyfert, Paul; Shapkin, Mikhail; Shapoval, Illya; Shcheglov, Yury; Shears, Tara; Shekhtman, Lev; Shevchenko, Vladimir; Shires, Alexander; Silva Coutinho, Rafael; Simi, Gabriele; Sirendi, Marek; Skidmore, Nicola; Skwarnicki, Tomasz; Smith, Anthony; Smith, Edmund; Smith, Eluned; Smith, Jackson; Smith, Mark; Snoek, Hella; Sokoloff, Michael; Soler, Paul; Soomro, Fatima; Souza, Daniel; Souza De Paula, Bruno; Spaan, Bernhard; Sparkes, Ailsa; Spradlin, Patrick; Sridharan, Srikanth; Stagni, Federico; Stahl, Marian; Stahl, Sascha; Steinkamp, Olaf; Stenyakin, Oleg; Stevenson, Scott; Stoica, Sabin; Stone, Sheldon; Storaci, Barbara; Stracka, Simone; Straticiuc, Mihai; Straumann, Ulrich; Stroili, Roberto; Subbiah, Vijay Kartik; Sun, Liang; Sutcliffe, William; Swientek, Krzysztof; Swientek, Stefan; Syropoulos, Vasileios; Szczekowski, Marek; Szczypka, Paul; Szilard, Daniela; Szumlak, Tomasz; T'Jampens, Stephane; Teklishyn, Maksym; Tellarini, Giulia; Teubert, Frederic; Thomas, Christopher; Thomas, Eric; van Tilburg, Jeroen; Tisserand, Vincent; Tobin, Mark; Tolk, Siim; Tomassetti, Luca; Topp-Joergensen, Stig; Torr, Nicholas; Tournefier, Edwige; Tourneur, Stephane; Tran, Minh Tâm; Tresch, Marco; Tsaregorodtsev, Andrei; Tsopelas, Panagiotis; Tuning, Niels; Ubeda Garcia, Mario; Ukleja, Artur; Ustyuzhanin, Andrey; Uwer, Ulrich; Vacca, Claudia; Vagnoni, Vincenzo; Valenti, Giovanni; Vallier, Alexis; Vazquez Gomez, Ricardo; Vazquez Regueiro, Pablo; Vázquez Sierra, Carlos; Vecchi, Stefania; Velthuis, Jaap; Veltri, Michele; Veneziano, Giovanni; Vesterinen, Mika; Viaud, Benoit; Vieira, Daniel; Vieites Diaz, Maria; Vilasis-Cardona, Xavier; Vollhardt, Achim; Volyanskyy, Dmytro; Voong, David; Vorobyev, Alexey; Vorobyev, Vitaly; Voß, Christian; Voss, Helge; de Vries, Jacco; Waldi, Roland; Wallace, Charlotte; Wallace, Ronan; Walsh, John; Wandernoth, Sebastian; Wang, Jianchun; Ward, David; Watson, Nigel; Websdale, David; Whitehead, Mark; Wicht, Jean; Wiedner, Dirk; Wilkinson, Guy; Williams, Matthew; Williams, Mike; Wilschut, Hans; Wilson, Fergus; Wimberley, Jack; Wishahi, Julian; Wislicki, Wojciech; Witek, Mariusz; Wormser, Guy; Wotton, Stephen; Wright, Simon; Wyllie, Kenneth; Xie, Yuehong; Xing, Zhou; Xu, Zhirui; Yang, Zhenwei; Yuan, Xuhao; Yushchenko, Oleg; Zangoli, Maria; Zavertyaev, Mikhail; Zhang, Liming; Zhang, Wen Chao; Zhang, Yanxi; Zhelezov, Alexey; Zhokhov, Anatoly; Zhong, Liang; Zvyagin, Alexander

2015-02-04

Using the latest LHCb measurements of time-dependent $C\\!P$ violation in the $B^0_s \\to K^+K^-$ decay, a U-spin relation between the decay amplitudes of $B^0_s \\to K^+K^-$ and $B^0\\to \\pi^+\\pi^-$ decay processes allows constraints to be placed on the angle $\\gamma$ of the unitarity triangle and on the $B^0_s$ mixing phase $-2\\beta_s$. Results from an extended approach, which uses additional inputs on $B^0\\to \\pi^0\\pi^0$ and $B^+\\to \\pi^+\\pi^0$ decays from other experiments and exploits isospin symmetry, are also presented. The dependence of the results on the maximum allowed amount of U-spin breaking is studied. At 68% probability, the value $\\gamma = \\left( 63.5^{\\,+\\, 7.2}_{\\,-\\,6.7} \\right)^\\circ~\\mathrm{modulo}~180^\\circ$ is determined. In an alternative analysis, the value $-2\\beta_s = -0.12 ^{\\,+\\,0.14}_{\\,-\\,0.16}\\,\\,\\mathrm{rad}$ is found. In both measurements, the uncertainties due to U-spin breaking effects up to 50% are included.

Shielding assessment for the proposed HRIBF upgrade to the National ISOL Facility

International Nuclear Information System (INIS)

Slater, C.O.; Olsen, D.K.; Johnson, J.O.; Lillie, R.A.; Gabriel, T.A.

1997-04-01

An upgrade of the existing ORNL Holifield Radioactive Ion Beam Facility (HRIBF) to the National Radioactive Ion Beam Isotope Separator On Line (RIB ISOL) Facility is being proposed. Part of the upgrade involves increasing the source proton energy and current, resulting in more intense, higher energy radiation. Shielding requirements for the proposed upgrade to the HRIBF have been assessed with respect to weight, space, and dose-rate constraints. Shielding assessments were made for operating, shutdown, and accident conditions. The results indicate reasonable shielding solutions for the target room except for the marginal dose rate on the roof. Shielding requirements in the target room were greatly reduced by decisions to move the target to a more interior room and to direct the proton beam downward into the target. A slightly more difficult shielding problem arises for proton beam extraction losses from the cyclotron. Here, the assumed isotropic beam losses (hence, neutron emissions) mean higher roof dose rates than those over the target room unless substantial localized shielding is placed over the cyclotron. Shutdown dose rates were found to present no problems. While dose rates through the sides of the facility during accident conditions will probably satisfy the accident dose-rate constraints, dose rates above the roof will be well above the constraints unless a solution is devised to shield the locations where beam losses are likely to occur. Ground activation analysis was postponed for this study

Density functional calculations of backbone 15N shielding tensors in beta-sheet and turn residues of protein G

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.

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

International Nuclear Information System (INIS)

Helfi-Yuliati; Mukhlis-Akhadi

2003-01-01

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

MicroShield/ISOCS gamma modeling comparison.

Energy Technology Data Exchange (ETDEWEB)

Sansone, Kenneth R

2013-08-01

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

Simulated progress in double-beta decay

International Nuclear Information System (INIS)

Miley, H.S.; Arthur, R.J.; Avignone, F.T.

1993-09-01

A Monte Carlo code has been developed to accurately simulate double-beta decay measurements. Coincident gamma rays, beta spectra, and angular correlations have been added to adequately simulate a complete 100 Mo nuclear decay and provide corrections to experimentally determined detector efficiencies. This code has been used to strip certain low-background spectra obtained in the Homestake gold mine in Lead, SD, for the purpose of extremely sensitive materials assay for the construction of new, large, enriched germanium detectors. Assays as low as 9 μBq/g of 210 Pb in lead shielding were obtained

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-03-15

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

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

International Nuclear Information System (INIS)

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

2017-01-01

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

Microcalorimetric and spectrographic studies on host-guest interactions of {alpha}-, {beta}-, {gamma}- and M{beta}-cyclodextrin with resveratrol

Energy Technology Data Exchange (ETDEWEB)

Li, Hui; Xu, Xiangyu; Liu, Min [College of Chemistry and Chemical Engineering, Liaocheng University, Liaocheng 252059, Shandong Province (China); Sun, Dezhi, E-mail: sundezhisdz@163.com [College of Chemistry and Chemical Engineering, Liaocheng University, Liaocheng 252059, Shandong Province (China); Li, Linwei, E-mail: lilinwei@lcu.edu.cn [College of Chemistry and Chemical Engineering, Liaocheng University, Liaocheng 252059, Shandong Province (China)

2010-10-20

Thermal effects of inclusion processes of {alpha}-, {beta}-, {gamma}- and M{beta}-cyclodextrin with resveratrol (RES) in aqueous solutions were determined by isothermal titration calorimetry (ITC) with nanowatt sensitivity at the temperature of 298.15 K. Standard enthalpy changes, stoichiometry and equilibrium constants of the inclusion complexes were derived from the direct calorimetric data utilizing nonlinear simulation. The thermodynamic parameters were discussed in the light of weak interactions between the host and the guest molecules combining with UV spectral message. The results indicate that all of the complexes formed in the aqueous solutions are in 1:1 stoichiometry. The binding processes of {alpha}-, {beta}- and M{beta}-cyclodextrin with the guest are mainly driven by enthalpy, while that of {gamma}-cyclodextrin with the drug is driven by both enthalpy and entropy.

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

Science.gov (United States)

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

2018-05-01

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

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

International Nuclear Information System (INIS)

Penkuhn, H.

1977-04-01

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

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

Directory of Open Access Journals (Sweden)

Reza Bagheri

2017-02-01

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

Beta, but not gamma, band oscillations index visual form-motion integration.

Directory of Open Access Journals (Sweden)

Charles Aissani

Full Text Available Electrophysiological oscillations in different frequency bands co-occur with perceptual, motor and cognitive processes but their function and respective contributions to these processes need further investigations. Here, we recorded MEG signals and seek for percept related modulations of alpha, beta and gamma band activity during a perceptual form/motion integration task. Participants reported their bound or unbound perception of ambiguously moving displays that could either be seen as a whole square-like shape moving along a Lissajou's figure (bound percept or as pairs of bars oscillating independently along cardinal axes (unbound percept. We found that beta (15-25 Hz, but not gamma (55-85 Hz oscillations, index perceptual states at the individual and group level. The gamma band activity found in the occipital lobe, although significantly higher during visual stimulation than during base line, is similar in all perceptual states. Similarly, decreased alpha activity during visual stimulation is not different for the different percepts. Trial-by-trial classification of perceptual reports based on beta band oscillations was significant in most observers, further supporting the view that modulation of beta power reliably index perceptual integration of form/motion stimuli, even at the individual level.

CaSO4: Dy + Teflon dosimetric pellets for X, beta and gamma radiation detection

International Nuclear Information System (INIS)

Campos, L.L.; Lima, M.F.

1987-08-01

CaSO 4 : Dy + TEFLON dosimetric pellets with high sensitivity and low cost for X, beta and gamma radiation monitoring were studied and developed by the Dosimetric Material Production Laboratory of the Radiological Protection Departament and are disposable for sale. The thickness of the pellets are suitable for X, beta and gamma radiation measurements. The dosimetric properties of these pellets were determined and presented in this work. The results show the usefulness of 0,20mm thick pellets for beta radiation monitoring and 0,80mm thick pellets for x and gamma radiation detection. (Author) [pt

Beta and gamma decay heat evaluation for the thermal fission of 235U

International Nuclear Information System (INIS)

Schenter, G.K.; Schmittroth, F.

1979-01-01

Beta and gamma fission product decay heat curves are evaluated for the thermal fission of 235 U. Experimental data that include beta, gamma, and total measurements are combined with summation calculations based on ENDF/B in a consistent evaluation. Least-squares methods are used that take proper account of data uncertainties and correlations. 4 figures, 2 tables

Optimization of the National Ignition Facility primary shield design

International Nuclear Information System (INIS)

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

1993-10-01

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

Effect of water on self-assembled tubules in {beta}-sitosterol + {gamma}-oryzanol-based organogels

Energy Technology Data Exchange (ETDEWEB)

Adel, Ruud den; Heussen, Patricia C M; Bot, Arjen, E-mail: ruud-den.adel@unilever.co [Unilever Research and Development Vlaardingen, Olivier van Noortlaan 120, NL-3133 AT Vlaardingen (Netherlands)

2010-10-01

Mixtures of {beta}-sitosterol and {gamma}-oryzanol form a network in triglyceride oil that may serve as an alternative to the network of small crystallites of triglycerides occurring in regular oil structuring. The present x-ray diffraction study investigates the relation between the crystal forms of the individual compounds and the mixture in oil, water and emulsion. {beta}-Sitosterol and {gamma}-oryzanol form normal crystals in oil, in water, or in emulsions. The crystals are sensitive to the presence of water. The mixture of {beta}-sitosterol + {gamma}-oryzanol forms crystals in water and emulsions that can be traced back to the crystals of the pure compounds. Only in oil, a completely different structure emerges in the mixture of {beta}-sitosterol + {gamma}-oryzanol, which bears no relation to the structures that are formed by both individual compounds, and which can be identified as a self-assembled tubule (diameter 7.2{+-}0.1 nm, wall thickness 0.8{+-}0.2 nm).

Using the Monte Carlo Coupling Technique to Evaluate the Shielding Ability of a Modular Shielding House to Accommodate Spent-Fuel Transportable Storage Casks

International Nuclear Information System (INIS)

Ueki, Kohtaro; Kawakami, Kazuo; Shimizu, Daisuke

2003-01-01

The Monte Carlo coupling technique with the coordinate transformation is used to evaluate the shielding ability of a modular shielding house that accommodates four spent-fuel transportable storage casks for two units. The effective dose rate distributions can be obtained as far as 300 m from the center of the shielding house. The coupling technique is created with the Surface Source Write (SSW) card and the Surface Source Read/Coordinate Transformation (SSR/CRT) card in the MCNP 4C continuous energy Monte Carlo code as the 'SSW-SSR/CRT calculation system'. In the present Monte Carlo coupling calculation, the total effective dose rates 100, 200, and 300 m from the center of the shielding house are estimated to be 1.69, 0.285, and 0.0826 (μSv/yr per four casks), respectively. Accordingly, if the distance between the center of the shielding house and the site boundary of the storage facility is kept at >300 m, approximately 2400 casks are able to be accommodated in the modular shielding houses, under the Japanese severe criterion of 50 μSv/yr at the site boundary. The shielding house alone satisfies not only the technical conditions but also the economic requirements.It became evident that secondary gamma rays account for >60% of the effective total dose rate at all the calculated points around the shielding house, most of which are produced from the water in the steel-water-steel shielding system of the shielding house. The remainder of the dose rate comes mostly from neutrons; the fission product and 60 Co activation gamma rays account for small percentages. Accordingly, reducing the secondary gamma rays is critical to improving not only the shielding ability but also the radiation safety of the shielding house

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-03-15

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

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

Astronaut Ronald Sega with Wake Shield Facility on test stand at JSC

Science.gov (United States)

1991-01-01

The Wake Shield Facility is displayed on a test stand at JSC. Astronaut Ronald M. Sega, mission specialist for STS-60, is seen with the facility during a break in testing in the acoustic and vibration facility at JSC.

Concrete shielding for nuclear ship 'Mutsu'

International Nuclear Information System (INIS)

Nagase, Tetsuo; Nakajima, Tadao; Okumura, Tadahiko; Saito, Tetsuo

1983-01-01

The nuclear ship ''Mutsu'' was constructed in 1970 as the fourth in the world. On September 1, 1974, during the power raising test in the Pacific Ocean, radiation leak was detected. As the result of investigation, it was found that the cause was the fast neutrons streaming through the gap between the reactor pressure vessel and the primary shield. In order to repair the shielding facility, the Japan Nuclear Ship Research Development Agency carried out research and development and shielding design. It was decided to adopt serpentine concrete for the primary shield, which is the excellent moderator of fast neutrons even at high temperature, and heavy concrete for the secondary shield, which is effective for shielding both gamma ray and neutron beam. The repair of shielding was carried out in the Sasebo Shipyard, and completed in August, 1982. The outline of the repair work is reported. The weight increase was about 300 t. The conditions of the shielding design, the method of shielding analysis, the performance required for the shielding concrete, the preliminary experiment on heavy concrete and the construction works of serpentine concrete and heavy concrete are described. (Kako, I.)

{beta}-{gamma} systems and the deformations of the BRST operator

Energy Technology Data Exchange (ETDEWEB)

Zeitlin, Anton M [Department of Mathematics, Yale University, 442 Dunham Lab, 10 Hillhouse Ave, New Haven, CT 06511 (United States)], E-mail: anton.zeitlin@yale.edu

2009-09-04

We describe the relation between simple logarithmic CFTs associated with closed and open strings, and their 'infinite metric' limits, corresponding to the {beta}-{gamma} systems. This relation is studied on the level of the BRST complex: we show that the consideration of metric as a perturbation leads to a certain deformation of the algebraic operations of the Lian-Zuckerman type on the vertex algebra, associated with the {beta}-{gamma} systems. The Maurer-Cartan equations corresponding to this deformed structure in the quasi-classical approximation lead to the nonlinear field equations. As an explicit example, we demonstrate that using this construction, Yang-Mills equations can be derived. This gives rise to a nontrivial relation between the Courant-Dorfman algebroid and homotopy algebras emerging from the gauge theory. We also discuss a possible algebraic approach to the study of beta-functions in sigma-models.

Ice shielding in the large scale GENIUS experiment for double beta decay and dark matter search

International Nuclear Information System (INIS)

Klapdor-Kleingrothaus, H.V.; Zdesenko, Yu.G.

1998-01-01

We suggest here the use of ice as shielding material in the large scale GENIUS experiment for the ultimate sensitive double beta decay and dark matter search. The idea is to pack a working volume of several tons of liquid nitrogen, which contains the ''naked'' Ge detectors, inside an ice shielding. Very thin plastic foil would be used in order to prevent leakage of the liquid nitrogen. Due to the excellent advantages of ice shielding (high purity and low cost, self-supporting ability, thermo-isolation and optical properties, safety) this could be another possible way of realization of the GENIUS project. (orig.)

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

Science.gov (United States)

Bakshi, Jayeesh

2018-04-01

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

High-$\\gamma$ Beta Beams within the LAGUNA design study

CERN Document Server

Orme, Christopher

2010-01-01

Within the LAGUNA design study, seven candidate sites are being assessed for their feasibility to host a next-generation, very large neutrino observatory. Such a detector will be expected to feature within a future European accelerator neutrino programme (Superbeam or Beta Beam), and hence the distance from CERN is of critical importance. In this article, the focus is a $^{18}$Ne and $^{6}$He Beta Beam sourced at CERN and directed towards a 50 kton Liquid Argon detector located at the LAGUNA sites: Slanic (L=1570 km) and Pyh\\"{a}salmi (L=2300 km). To improve sensitivity to the neutrino mass ordering, these baselines are then combined with a concurrent run with the same flux directed towards a large Water \\v{C}erenkov detector located at Canfranc (L=650 km). This degeneracy breaking combination is shown to provide comparable physics reach to the conservative Magic Baseline Beta Beam proposals. For $^{18}$Ne ions boosted to $\\gamma=570$ and $^{6}$He ions boosted to $\\gamma=350$, the correct mass ordering can be...

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)

Investigation of the effect of barium content on the structural and gamma-ray shielding properties of bismuth borate glasses

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)

Development of radiation safety monitoring system at gamma greenhouse gamma facility

International Nuclear Information System (INIS)

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

2009-01-01

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

Alpha-beta-gamma spectrometer as an aid in directing decontamination of soils

International Nuclear Information System (INIS)

Berger, C.D.

1981-01-01

This technique permits rapid assessment of alpha-beta-gamma-emitter contamination in soils at sufficiently low concentrations to direct field operations. Of particular importance is its applicability during initial decommissioning and decontamination surveys when characterization of alpha and beta contamination in the presence of a high gamma background is necessary. This system has not yet been made portable for in-situ use, but it is expected that results willbe favorable when operated as a field instrument, resulting in simplified standard decontamination operation

Procedure for the elaboration of extended sources beta and/or gamma emitting

International Nuclear Information System (INIS)

Tejera R, A.; Cortes P, A.; Becerril V, A.

1991-12-01

In the laboratory of radioactive standards they have been come manufacturing punctual sources gauged gamma emitting during several years. Before the demand of extended radioactive homogeneous sources of beta particles emitting, in particular with nuclides that are simultaneously gamma and beta emitting, it was designed a procedure for it elaboration based on the one that we use at the moment for the elaboration of the punctual gamma sources. This procedure consists on the integration of a compact group of this type of sources on a single extended support, sealed one of its faces with a film of transparent material in satisfactory grade to the beta particles. In this work this procedure is described and it is applied in the elaboration of two sources that its were requested by the Laguna Verde Central (CFE), one with area of 20 cm 2 and the other one of 100 cm 2 . The homogeneity, measure as the dispersion of the activities of the aliquot ones distributed in the active surfaces was inside 2%. The percentage of attenuation of the beta particles was also measured by the film (window) with the one that the sources were sealed. (Author)

Responses of a grassland arthropod community to chronic beta and gamma radiation

International Nuclear Information System (INIS)

Styron, C.E.; Dodson, G.J.; Beauchamp, J.J.; Miller, F.L. Jr.

1976-01-01

A long-term project was initiated in 1968 at Oak Ridge National Laboratory to assess effects of mixed beta and gamma radiation from simulated fallout on a grassland ecosystem. Beta and gamma radiation dose rates in microhabitats of the experimentally contaminated enclosure were measured with LiF thermoluminescent microdosimeters. Extensive statistical analyses of data on numbers of individuals collected for each of 76 arthropod and 2 molluscan taxa have identified no lasting significant changes in similarity or species diversity of experimental versus control communities as the result of the long-term irradiation at low dose rates. Natural fluctuations in community dynamics obscured any possible radiation effects. Thus, the apparent threshold for mixed beta and gamma radiation inducing changes in community structure must be above the exposure rate range of 2.3 to 13 rad/day delivered during the 5 yr of observation. Establishing such a threshold is of importance in assessing the impact of communities subjected to chronic, low level environmental exposure to ionizing radiation

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

Study of gamma radiation shielding properties of ZnO-TeO_2 glasses

International Nuclear Information System (INIS)

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

2017-01-01

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

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

International Nuclear Information System (INIS)

Nonato, Fernanda Beatrice Conceicao

2010-01-01

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

Electron accelerator shielding design of KIPT neutron source facility

Energy Technology Data Exchange (ETDEWEB)

Zhong, Zhao Peng; Gohar, Yousry [Argonne National Laboratory, Argonne (United States)

2016-06-15

The Argonne National Laboratory of the United States and the Kharkov Institute of Physics and Technology of the Ukraine have been collaborating on the design, development and construction of a neutron source facility at Kharkov Institute of Physics and Technology utilizing an electron-accelerator-driven subcritical assembly. The electron beam power is 100 kW using 100-MeV electrons. The facility was designed to perform basic and applied nuclear research, produce medical isotopes, and train nuclear specialists. The biological shield of the accelerator building was designed to reduce the biological dose to less than 5.0e-03 mSv/h during operation. The main source of the biological dose for the accelerator building is the photons and neutrons generated from different interactions of leaked electrons from the electron gun and the accelerator sections with the surrounding components and materials. The Monte Carlo N-particle extended code (MCNPX) was used for the shielding calculations because of its capability to perform electron-, photon-, and neutron-coupled transport simulations. The photon dose was tallied using the MCNPX calculation, starting with the leaked electrons. However, it is difficult to accurately tally the neutron dose directly from the leaked electrons. The neutron yield per electron from the interactions with the surrounding components is very small, ∼0.01 neutron for 100-MeV electron and even smaller for lower-energy electrons. This causes difficulties for the Monte Carlo analyses and consumes tremendous computation resources for tallying the neutron dose outside the shield boundary with an acceptable accuracy. To avoid these difficulties, the SOURCE and TALLYX user subroutines of MCNPX were utilized for this study. The generated neutrons were banked, together with all related parameters, for a subsequent MCNPX calculation to obtain the neutron dose. The weight windows variance reduction technique was also utilized for both neutron and photon dose

A shielding chamber for the Rossendorf whole body counter

International Nuclear Information System (INIS)

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

2016-01-01

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

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

International Nuclear Information System (INIS)

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

1977-01-01

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

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

Directory of Open Access Journals (Sweden)

M.I. Sayyed

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

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

Energy Technology Data Exchange (ETDEWEB)

Leimdoerfer, M

1964-02-15

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

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

International Nuclear Information System (INIS)

Leimdoerfer, M.

1964-02-01

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

Skin dose assessment in routine personnel beta/gamma dosimetry

International Nuclear Information System (INIS)

Christensen, P.

1980-01-01

The International Commission on Radiological Protection (Publication 26) has recommended a tissue depth of 5 to 10 mg.cm -2 for skin dose assessments. This requirement is generally not fulfilled by routine monitoring procedures because of practical difficulties in using very thin dosemeters with low sensitivity and therefore a high minimum detectable dose. Especially for low-energy beta-ray exposures underestimations of the skin dose by a factor of more than ten may occur. Low-transparent graphite-mixed sintered LiF and Li 2 B 4 0 7 : Mn dosemeters were produced which show a skin-equivalent response to beta and gamma exposures over a wide range of energies. These have found wide-spread application for extremity dosimetry but have not yet been generally introduced in routine personnel beta/gamma monitoring. The following adaptations of existing routine monitoring systems for improved skin dose assessments have been investigated: 1) Placement of a supplementary, thin, skin-dose equivalent dosemeter in the TLD badge to give additional information on low-energy exposures. 2) Introduction of a second photomultiplier in the read-out chamber which enables a simultaneous determination of emitted TL from both sides of the dosemeter separately. This method makes use of the selfshielding of the dosemeter to give information on the low-energy dose contribution. 3) By diffusion of Li 2 B 4 0 7 into solid LiF-dosemeters it was possible to produce a surface layer with a new distinct glow-peak at about 340 deg C which is not present in the undiffused part of the LiF chip, and which can be utilized for the assessment of the skin-dose. Data on energy response and accuracy of dose measurement for beta/gamma exposures are given for the three methods and advantages and disadvantages are discussed (H.K.)

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-04-15

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

Mixing of ground-state rotational and gamma and beta vibrational bands in the region A>=228

Energy Technology Data Exchange (ETDEWEB)

Mittal, R; Sahota, H S [Punjabi Univ., Patiala (India). Dept. of Physics

1983-06-21

The mixing of beta, gamma and ground-state bands has been investigated through the experimental determination of mixing parameters Zsub(..gamma..) and Zsub(..beta gamma..). These Zsub(..gamma..) values have been compared with the theoretical calculations of this parameter from the solutions of time-dependent HFB equations on the adiabatic and nonadiabatic assumptions. The experimental values are in better agreement with the results obtained under the nonadiabatic assumption, valid for small deviations from the spherical symmetry.

High and low energy gamma beam dump designs for the gamma beam delivery system at ELI-NP

International Nuclear Information System (INIS)

Yasin, Zafar; Matei, Catalin; Ur, Calin A.; Mitu, Iani-Octavian; Udup, Emil; Petcu, Cristian

2016-01-01

The Extreme Light Infrastructure - Nuclear Physics (ELI-NP) is under construction in Magurele, Bucharest, Romania. The facility will use two 10 PW lasers and a high intensity, narrow bandwidth gamma beam for stand-alone and combined laser-gamma experiments. The accurate estimation of particle doses and their restriction within the limits for both personel and general public is very important in the design phase of any nuclear facility. In the present work, Monte Carlo simulations are performed using FLUKA and MCNPX to design 19.4 and 4 MeV gamma beam dumps along with shielding of experimental areas. Dose rate contour plots from both FLUKA and MCNPX along with numerical values of doses in experimental area E8 of the facility are performed. The calculated doses are within the permissible limits. Furthermore, a reasonable agreement between both codes enhances our confidence in using one or both of them for future calculations in beam dump designs, radiation shielding, radioactive inventory, and other calculations releated to radiation protection. Residual dose rates and residual activity calculations are also performed for high-energy beam dump and their effect is negligible in comparison to contributions from prompt radiation.

High and low energy gamma beam dump designs for the gamma beam delivery system at ELI-NP

Energy Technology Data Exchange (ETDEWEB)

Yasin, Zafar, E-mail: zafar.yasin@eli-np.ro; Matei, Catalin; Ur, Calin A.; Mitu, Iani-Octavian; Udup, Emil; Petcu, Cristian [Extreme Light Infrastructure - Nuclear Physics / Horia Hulubei National Institute for R& D in Physics and Nuclear Engineering, Bucharest-Magurele (Romania)

2016-03-25

The Extreme Light Infrastructure - Nuclear Physics (ELI-NP) is under construction in Magurele, Bucharest, Romania. The facility will use two 10 PW lasers and a high intensity, narrow bandwidth gamma beam for stand-alone and combined laser-gamma experiments. The accurate estimation of particle doses and their restriction within the limits for both personel and general public is very important in the design phase of any nuclear facility. In the present work, Monte Carlo simulations are performed using FLUKA and MCNPX to design 19.4 and 4 MeV gamma beam dumps along with shielding of experimental areas. Dose rate contour plots from both FLUKA and MCNPX along with numerical values of doses in experimental area E8 of the facility are performed. The calculated doses are within the permissible limits. Furthermore, a reasonable agreement between both codes enhances our confidence in using one or both of them for future calculations in beam dump designs, radiation shielding, radioactive inventory, and other calculations releated to radiation protection. Residual dose rates and residual activity calculations are also performed for high-energy beam dump and their effect is negligible in comparison to contributions from prompt radiation.

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)

Differential expression of largemouth bass (Micropterus salmoides) estrogen receptor isotypes alpha, beta, and gamma by estradiol.

Science.gov (United States)

Sabo-Attwood, Tara; Kroll, Kevin J; Denslow, Nancy D

2004-04-15

The expression levels of three estrogen receptor (ER) isotypes alpha, beta, and gamma were quantified in female largemouth bass (Micropterus salmoides) (LMB) liver, ovary, brain, and pituitary tissues. ER alpha and beta expression predominated in the liver, while ERs beta and gamma predominated in the other tissues. Temporally in females, ER alpha was highly up-regulated, ER gamma was slightly up-regulated, and ER beta levels remained unchanged in the liver when plasma 17-beta estradiol (E2) and vitellogenin (Vtg) levels were elevated in the spring. In ovarian tissue from these same fish, all three ERs were maximally expressed in the fall, during early oocyte development and prior to peak plasma E2 levels. When males were injected with E2, ER alpha was highly inducible, ER gamma was moderately up-regulated, and ER beta levels were not affected. None of the ER isotypes were induced by E2 in gonadal tissues. These results combined suggest that the ERs themselves are not regulated in the same manner by E2, and furthermore, do not contribute equally to the transcriptional regulation of genes involved in fish reproduction such as Vtg.

Hanford beta-gamma personnel dosimeter prototypes and evaluation

International Nuclear Information System (INIS)

Fix, J.J.; Holbrook, K.L.; Soldat, K.L.

1983-04-01

Upgraded and modified Hanford dosimeter prototypes were evaluated for possible use at Hanford as a primary beta-gamma dosimeter. All prototypes were compatible with the current dosimeter card and holder design, as well as processing with the automated Hanford readers. Shallow- and deep-dose response was determined for selected prototypes using several beta sources, K-fluorescent x rays and filtered x-ray techniques. All prototypes included a neutron sensitive chip. A progressive evaluation of the performance of each of the upgrades to the current dosimeter is described. In general, the performance of the current dosimeter can be upgraded using individual chip sensitivity factors to improve precision and an improved algorithm to minimize bias. The performance of this dosimeter would be adequate to pass all categories of the ANSI N13.11 performance criteria for dosimeter procesors, provided calibration techniques compatible with irradiations adopted in the standard were conducted. The existing neutron capability of the dosimeter could be retained. Better dosimeter performance to beta-gamma radiation can be achieved by modifying the Hanford dosimeter so that four of the five chip positions are devoted to calculating these doses instead of the currently used two chip positions. A neutron sensitive chip was used in the 5th chip position, but all modified dosimeter prototypes would be incapable of discriminating between thermal and epithermal neutrons. An improved low energy beta response can be achieved for the current dosimeter and all prototypes considered by eliminating the security credential. Further improvement can be obtained by incorporating the 15-mil thick TLD-700 chips

New applications and developments in the neutron shielding

Directory of Open Access Journals (Sweden)

Uğur Fatma Aysun

2017-01-01

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

New applications and developments in the neutron shielding

Science.gov (United States)

Uğur, Fatma Aysun

2017-09-01

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

Photon spectrum behind biological shielding of the LVR-15 research reactor

Energy Technology Data Exchange (ETDEWEB)

Klupak, V.; Viererbl, L.; Lahodova, Z.; Marek, M.; Vins, M. [Research Centre Rez Ltd., Husinec-Rez 130 (Czech Republic)

2011-07-01

The LVR-15 reactor is a light water research reactor situated at the Research Centre Rez, near Prague. It operates as a multipurpose facility with a maximum thermal power of 10 MW. The reactor core usually contains from 28 to 32 fuel assemblies with a total mass of {sup 235}U of about 5 kg. Emitted radiation from the fuel caused by fission is shielded by moderating water, a steel reactor vessel, and heavy concrete. This paper deals with measurement and analysis of the gamma spectrum near the outer surface of the concrete wall, behind biological shielding, mainly in the 3- to 10-MeV energy range. A portable HPGe detector with a portable multichannel analyzer was used to measure gamma spectra. The origin of energy lines in gamma detector spectra was identified. (authors)

Systematics of gamma-ray energy spectra for classification of workplaces around a nuclear facility

International Nuclear Information System (INIS)

Urabe, Itsumasa; Tsujimoto, Tadashi; Katsurayama, Kousuke

1988-01-01

Radiation dosimetry in workplaces has been carried out both for assurance of the doses complying with the acceptable values and for improvement of protection methods to minimise detriments of the exposed population. This means that it is very important not only to determine dosimetric quantities in workplaces but also to know features of radiation levels because information for radiation protection can often be derived from the radiometric quantities. Classification of workplaces based on the feature of gamma-ray energy spectra is one of the practical ways to realise radiation protection being taken into consideration of the radiometric quantities. Furthermore, demarcation of workplaces based on these radiometric quantities may be effective for improvement of radiation protection practice such as estimation of radiation doses, designing of radiation shields and other activities. From these points of view, gamma-ray energy spectra have been determined in various workplaces in nuclear facilities, and systematics of gamma-ray fields were tried for classification of workplaces on the basis of the feature appeared in health physical quantities such as effective dose equivalents and responses of dosemeters

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

International Nuclear Information System (INIS)

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

2017-01-01

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

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

International Nuclear Information System (INIS)

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

2014-01-01

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

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

Analysis method for beta-gamma coincidence spectra from radio-xenon isotopes

International Nuclear Information System (INIS)

Yang Wenjing; Yin Jingpeng; Huang Xiongliang; Cheng Zhiwei; Shen Maoquan; Zhang Yang

2012-01-01

Radio-xenon isotopes monitoring is one important method for the verification of CTBT, what includes the measurement methods of HPGe γ spectrometer and β-γ coincidence. The article describes the analytic flowchart and method of three-dimensional beta-gamma coincidence spectra from β-γ systems, and analyses in detail the principles and methods of the regions of interest of coincidence spectra and subtracting the interference, finally gives the formula of radioactivity of Xenon isotopes and minimum detectable concentrations. Studying on the principles of three-dimensional beta-gamma coincidence spectra, which can supply the foundation for designing the software of β-γ coincidence systems. (authors)

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-04-15

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

CD147 is a regulatory subunit of the gamma-secretase complex inAlzheimer's disease amyloid beta-peptide production

Energy Technology Data Exchange (ETDEWEB)

Zhou, Shuxia; Zhou, Hua; Walian, Peter J.; Jap, Bing K.

2005-04-06

{gamma}-secretase is a membrane protein complex that cleaves the {beta}-amyloid precursor protein (APP) within the transmembrane region, following prior processing by {beta}-secretase, producing amyloid {beta}-peptides (A{beta}{sub 40} and A{beta}{sub 42}). Errant production of A{beta}-peptides that substantially increases A{beta}{sub 42} production has been associated with the formation of amyloid plaques in Alzheimer's disease patients. Biophysical and genetic studies indicate that presenilin-1 (Psn-1), which contains the proteolytic active site, and three other membrane proteins, nicastrin (Nct), APH-1, and PEN-2 are required to form the core of the active {gamma}-secretase complex. Here, we report the purification of the native {gamma}-secretase complexes from HeLa cell membranes and the identification of an additional {gamma}-secretase complex subunit, CD147, a transmembrane glycoprotein with two immunoglobulin-like domains. The presence of this subunit as an integral part of the complex itself was confirmed through co-immunoprecipitation studies of the purified protein from HeLa cells and solubilized complexes from other cell lines such as neural cell HCN-1A and HEK293. Depletion of CD147 by RNA interference was found to increase the production of A{beta} peptides without changing the expression level of the other {gamma}-secretase components or APP substrates while CD147 overexpression had no statistically significant effect on amyloid {beta}-peptide production, other {gamma}-secretase components or APP substrates, indicating that the presence of the CD147 subunit within the {gamma}-secretase complex directly down-modulates the production of A{beta}-peptides. {gamma}-secretase was first recognized through its role in the production of the A{beta} peptides that are pathogenic in Alzheimer's disease (AD) (1). {gamma}-secretase is a membrane protein complex with unusual aspartyl protease activity that cleaves a variety of type I membrane proteins

Formation of hydrogen fluoride by gamma and beta sterilisation in medical devices containing perfluoroheptane

International Nuclear Information System (INIS)

Zuendorf, Josef; Kremer, Stefan; Grueger, Thomas

2008-01-01

Infusion of hexadecafluoroheptane, a liquid perfluorocarbon released from repaired Althane dialysers was found to be the most probable reason for the deaths of 53 dialysis patients reported in the year 2001. This study focuses on toxic decomposition products generated due to gamma and beta sterilisation of hexadecafluoroheptane. The responsible dialysers were sterilised with a maximum dose of 45 kGy gamma irradiation. We investigated the influence of both 20-500 kGy gamma and beta irradiation on perfluoroheptane. Analysis of the irradiated samples verified the decomposition of perfluoroheptane in dependence on the dose of irradiation. Beta irradiation resulted in a higher degree of decomposition than the same dose of gamma irradiation. As decomposition products, hydrogen fluoride, CO 2 , and one saturated fluorinated hydrocarbon which could not be analysed exactly were identified. Even at 20 kGy gamma irradiation hydrogen fluoride was detectable. Our results provide evidence that hydrogen fluoride is generated as a highly toxic decomposition product when perfluoroheptane is sterilised with gamma irradiation as it was applied on the affected dialysers. There is no evidence of other toxic degradation products especially perfluoroisobutylene. Therefore, hydrogen fluoride or the dissociated fluoride ions might act as a toxic agent when medical devices containing liquid perfluorocarbons are sterilised by irradiation

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

Solved problems in analysis as applied to gamma, beta, Legendre and Bessel functions

CERN Document Server

Farrell, Orin J

2013-01-01

Nearly 200 problems, each with a detailed, worked-out solution, deal with the properties and applications of the gamma and beta functions, Legendre polynomials, and Bessel functions. The first two chapters examine gamma and beta functions, including applications to certain geometrical and physical problems such as heat-flow in a straight wire. The following two chapters treat Legendre polynomials, addressing applications to specific series expansions, steady-state heat-flow temperature distribution, gravitational potential of a circular lamina, and application of Gauss's mechanical quadrature

Endogenous PKI gamma limits the duration of the anti-apoptotic effects of PTH and beta-adrenergic agonists in osteoblasts.

Science.gov (United States)

Chen, Xin; Song, In-Hwan; Dennis, James E; Greenfield, Edward M

2007-05-01

PKI gamma knockdown substantially extended the anti-apoptotic effects of PTH and beta-adrenergic agonists, whereas PKI gamma overexpression decreased these effects. Therefore, inhibition of PKI gamma activity may provide a useful co-therapy in combination with intermittent PTH or beta-adrenergic agonists for bone loss in conditions such as osteoporosis. PTH has both catabolic and anabolic effects on bone, which are primarily caused by cAMP/protein kinase A (PKA) signaling and regulation of gene expression. We previously showed that protein kinase inhibitor-gamma (PKI gamma) is required for efficient termination of cAMP/PKA signaling and gene expression after stimulation with PTH or beta-adrenergic agonists. Inhibition of osteoblast apoptosis is thought to be an important, but transient, mechanism partly responsible for the anabolic effects of intermittent PTH. Therefore, we hypothesized that endogenous PKI gamma also terminates the anti-apoptotic effect of PTH. PKI gamma knockdown by antisense transfection or siRNA was used to examine the ability of endogenous PKI gamma to modulate the anti-apoptotic effects of PTH and beta-adrenergic agonists in ROS 17/2.8 cells. Knockdown of PKI gamma substantially extended the anti-apoptotic effects of PTH, whether apoptosis was induced by etoposide or dexamethasone. In contrast, overexpression of PKI gamma decreased the anti-apoptotic effect of PTH pretreatment. This study is also the first demonstration that beta-adrenergic agonists mimic the anti-apoptotic effects of PTH in osteoblasts. Moreover, PKI gamma knockdown also substantially extended this anti-apoptotic effect of beta-adrenergic agonists. Taken together, these results show that endogenous PKI gamma limits the duration of the anti-apoptotic effects of cAMP/PKA signaling in osteoblasts. Because significant individual variability exists in the anabolic responses to PTH therapy in current clinical treatment of osteoporosis, inhibition of PKI gamma activity may provide a

Ornithogalum virens as a plant assay for beta and gamma radiation effects

International Nuclear Information System (INIS)

Herron, V.J.

1979-01-01

The purpose of this study was to determine if the monocotyledonous angiosperm, Ornithogalum virens (Quintanilha and Cabral, 1947), could be used in such a biological assay system. After exposing O. virens plants to acute ( 60 Co) and chronic ( 137 Cs) gamma radiation and internal beta radiation ( 32 P), lethality (LD 50 , LD 100 ), growth inhibition, and chromosome aberrations were investigated. The LD 50 and LD 100 for acute gamma radiation were estimated to be between 0.91 to 1.8 krad and less than 3.6 krad, respectively. Though growth inhibition and abnormal growth were observed in the acute and chronic gamma radiation studies, the changes in the growth of the plants were so variable that these parameters were found to be unreliable measures of radiation effects. Chromosome aberrations were a more reliable measure of radiation damage because linear relationships between total aberrations and dose were found for both gamma and beta radiation

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

International Nuclear Information System (INIS)

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

2004-01-01

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

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

Science.gov (United States)

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

2004-12-01

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

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

International Nuclear Information System (INIS)

Hedemann Jensen, P.

1982-11-01

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

Nucleonic analysis of a preliminary design for the ETF neutral-beam-injector duct shielding

International Nuclear Information System (INIS)

Urban, W.T.; Seed, T.J.; Dudziak, D.J.

1980-01-01

A nucleonic analysis of the Engineering Test Facility Neutral-Beam-Injector duct shielding has been made using a hybrid Monte Carlo/discrete-ordinates method. This method used Monte Carlo to determine internal and external boundary surface sources for a subsequent discrete-ordinates calculation of the neutron and gamma-ray transport through the shield. The analysis also included determination of the energy and angular distribution of neutrons and gamma rays entering the duct from the torus plasma chamber. Confidence in the hybrid method and the results obtained were provided through a comparison with three-dimensional Monte Carlo results

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

International Nuclear Information System (INIS)

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

2000-01-01

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

Vibrational and chiroptical spectroscopic characterization of gamma-turn model cyclic tetrapeptides containing two beta-Ala residues.

Science.gov (United States)

Vass, Elemér; Majer, Zsuzsa; Kohalmy, Krisztina; Hollósi, Miklós

2010-08-01

The optical spectroscopic characterization of gamma-turns in solution is uncertain and their distinction from beta-turns is often difficult. This work reports systematic ECD and vibrational circular dichroism (VCD) spectroscopic studies on gamma-turn model cyclic tetrapeptides cyclo(Ala-beta-Ala-Pro-beta-Ala) (1), cyclo(Pro-beta-Ala-Pro-beta-Ala) (2) and cyclo(Ala-beta-Ala-Ala-beta-Ala) (3). Conformational analysis performed at the 6-31G(d)/B3LYP level of theory using an adequate PCM solvent model predicted one predominant conformer for 1-3, featuring two inverse gamma-turns. The ECD spectra in ACN of 1 and 2 are characterized by a negative n-->pi* band near 230 nm and a positive pi-->pi* band below 200 nm with a long wavelength shoulder. The ECD spectra in TFE of 1-3 show similar spectra with blue-shifted bands. The VCD spectra in ACN-d(3) of 1 and 2 show a +/-/+/- amide I sign pattern resulting from four uncoupled vibrations in the case of 1 and a sequence of two positive couplets in the case of 2. A -/+/+/- amide I VCD pattern was measured for 3 in TFE-d(2). All three peptides give a positive couplet or couplet-like feature (+/-) in the amide II region. VCD spectroscopy, in agreement with theoretical calculations revealed that low frequency amide I vibrations (at approximately 1630 cm(-1) or below) are indicative of a C(7) H-bonded inverse gamma-turns with Pro in position 2, while gamma-turns encompassing Ala absorb at higher frequency (above 1645 cm(-1)). Copyright 2010 Wiley-Liss, Inc.

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

International Nuclear Information System (INIS)

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

1991-01-01

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

Radiation safety of gamma and electron irradiation facilities

International Nuclear Information System (INIS)

1992-01-01

There are currently some 160 gamma irradiation facilities and over 600 electron beam facilities in operation throughout virtually all Member States of the IAEA. The most widespread uses of these facilities are for the sterilization of medical and pharmaceutical products, the preservation of foodstuffs, polymer synthesis and modification, and the eradication of insect infestation. The safety record of this industry has been very good. Nevertheless, there is a potential for accidents with serious consequences. Gamma and electron beam facilities produce very high dose rates during irradiation, so that a person accidentally present in the irradiation chamber can receive a lethal dose within minutes or seconds. Precautions against uncontrolled entry must therefore be taken. Furthermore, gamma irradiation facilities contain large amounts of radioactivity and if the mechanism for retracting the source is damaged, the source may remain exposed, inhibiting direct access to carry out remedial work. Contamination can result from corroded or damaged sources, and decontamination can be very expensive. These aspects clearly indicate the need to achieve a high degree of safety and reliability in the facilities. This can be accomplished by effective quality control together with careful design, manufacture, installation, operation and decommissioning. The guidance in this Safety Series publication is intended for competent authorities responsible for regulating the use of radiation sources as well as the manufacturers, suppliers, installers and users of gamma and electron beam facilities. 20 refs, 6 figs

Lessons from shielding retrofits at the LAMPF/LANSCE/PSR accelerator, beam lines and target facilities

International Nuclear Information System (INIS)

Macek, R.J.

1994-01-01

The experience in the past 7 years to improve the shielding and radiation control systems at the Los Alamos Meson Physics Facility (LAMPF) and the Manuel Lujan Jr. Neutron Scattering Center (LANSCE) provides important lessons for the design of radiation control systems at future, high beam power proton accelerator facilities. Major issues confronted and insight gained in developing shielding criteria and in the use of radiation interlocks are discussed. For accelerators and beam lines requiring hands-on-maintenance, our experience suggests that shielding criteria based on accident scenarios will be more demanding than criteria based on routinely encountered beam losses. Specification and analysis of the appropriate design basis accident become all important. Mitigation by active protection systems of the consequences of potential, but severe, prompt radiation accidents has been advocated as an alternate choice to shielding retrofits for risk management at both facilities. Acceptance of active protection systems has proven elusive primarily because of the difficulty in providing convincing proof that failure of active systems (to mitigate the accident) is incredible. Results from extensive shielding assessment studies are presented including data from experimental beam spill tests, comparisons with model estimates, and evidence bearing on the limitations of line-of-sight attenuation models in complex geometries. The scope and significant characteristics of major shielding retrofit projects at the LAMPF site are illustrated by the project to improve the shielding beneath a road over a multiuse, high-intensity beam line (Line D)

2015 In-Situ Gamma-Ray Assay of the West Cell Line in the 235-F Plutonium Fuel Form Facility

Energy Technology Data Exchange (ETDEWEB)

Brand, A. D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Aucott, T. J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); DiPrete, D. P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2016-02-01

In November and December 2015, scientists from SRNL took a series of in-situ gamma-ray measurements through the windows in front of Cells 6-9 on the west line of the PuFF facility using a shielded, 120% high-purity germanium detector. The detector efficiency was estimated using a combination of MCNP simulations and empirical measurements. Where possible, the distribution of the Pu-238 in the cells was determined using the Germanium Gamma-ray Imager (GeGI). This distribution was then fed into the MCNP model to quantify the Pu-238 in each cell. Data analysis was performed using three gamma rays emitted by Pu-238 (99.85 keV, 152.7 keV, and 766.4 keV) providing three independent estimates of the mass of Pu-238 holdup in each of the cells.

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

International Nuclear Information System (INIS)

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

2014-01-01

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

Neutron streaming analysis for shield design of FMIT Facility

International Nuclear Information System (INIS)

Carter, L.L.

1980-12-01

Applications of the Monte Carlo method have been summarized relevant to neutron streaming problems of interest in the shield design for the FMIT Facility. An improved angular biasing method has been implemented to further optimize the calculation of streaming and this method has been applied to calculate streaming within a double bend pipe

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

International Nuclear Information System (INIS)

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

2008-01-01

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

Resting-state beta and gamma activity in Internet addiction.

Science.gov (United States)

Choi, Jung-Seok; Park, Su Mi; Lee, Jaewon; Hwang, Jae Yeon; Jung, Hee Yeon; Choi, Sam-Wook; Kim, Dai Jin; Oh, Sohee; Lee, Jun-Young

2013-09-01

Internet addiction is the inability to control one's use of the Internet and is related to impulsivity. Although a few studies have examined neurophysiological activity as individuals with Internet addiction engage in cognitive processing, no information on spontaneous EEG activity in the eyes-closed resting-state is available. We investigated resting-state EEG activities in beta and gamma bands and examined their relationships with impulsivity among individuals with Internet addiction and healthy controls. Twenty-one drug-naïve patients with Internet addiction (age: 23.33 ± 3.50 years) and 20 age-, sex-, and IQ-matched healthy controls (age: 22.40 ± 2.33 years) were enrolled in this study. Severity of Internet addiction was identified by the total score on Young's Internet Addiction Test. Impulsivity was measured with the Barratt Impulsiveness Scale-11 and a stop-signal task. Resting-state EEG during eyes closed was recorded, and the absolute/relative power of beta and gamma bands was analyzed. The Internet addiction group showed high impulsivity and impaired inhibitory control. The generalized estimating equation showed that the Internet-addiction group showed lower absolute power on the beta band than did the control group (estimate = -3.370, p Internet-addiction group showed higher absolute power on the gamma band than did the control group (estimate = 0.434, p Internet addiction as well as with the extent of impulsivity. The present study suggests that resting-state fast-wave brain activity is related to the impulsivity characterizing Internet addiction. These differences may be neurobiological markers for the pathophysiology of Internet addiction. Copyright © 2013 Elsevier B.V. All rights reserved.

Operating manual for the Tower Shielding Facility

International Nuclear Information System (INIS)

1985-12-01

This manual provides information necessary to operate and perform maintenance on the reactor systems and all equipment or systems which can affect their operation or the safety of personnel at the Tower Shielding Facility. The first four chapters consist of introductory and descriptive material of benefit to personnel in training, the qualifications required for training, the responsibilities of the personnel in the organization, and the procedures for reviewing proposed experiments. Chapter 8, Emergency Procedures, is also a necessary part of the indoctrination of personnel. The procedures for operation of the Tower Shielding Reactor (TSR-II), its water cooling system, and the main tower hoists are outlined in Chapters 5, 6, and 7. The Technical Specification surveillance requirements for the TSR-II are summarized in Chapter 9. The maintenance and calibration schedule is spelled out in Chapter 10. The procedures for assembly and disassembly of the TSR-II are outlined in Chapter 11

Practical design of gamma irradiation facility

International Nuclear Information System (INIS)

Sugimoto, Sen-ichi

1976-01-01

In this report, it is intended to describe mainly the multi-purpose irradiation facilities which carry out the consigned irradiation for the sterilization of medical apparatuses, which is most of the demand of gamma irradiation in Japan. Gamma irradiation criterion is summed up to that ''Apply the specified dose properly and uniformly to product cases and be economic.'' Though the establishment of the design standard for irradiation facilities is not easy and is not solve simply, the factors to be considered in the design are as follows: (1) mechanism safety, (2) multipurpose irradiation structure, (3) irradiation criteria and practice, (4) efficiency of radiation source utilization and related problems, and (5) economical merit. Irradiation facilities are generally itemized as follows: irradiation equipments, radiation source-storing facility, package carrier, radiation source-driving equipments, facilities for safety and operational management and others. Examples and their characteristics are reported for the facilities of Japan Radio-isotope Irradiation Cooperative Association and Radie Industries Ltd. Expenses for construction, processing and radiation sources are shown on the basis of a few references, and the cost trially calculated under a certain presumptive condition is given. (Wakatsuki, Y.)

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

Energy Technology Data Exchange (ETDEWEB)

Roos, Matts

1959-07-15

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

Thermodynamics in Gliomas: Interactions between the Canonical WNT/Beta-Catenin Pathway and PPAR Gamma

Directory of Open Access Journals (Sweden)

Alexandre Vallée

2017-05-01

Full Text Available Gliomas cells are the site of numerous metabolic and thermodynamics abnormalities with an increasing entropy rate which is characteristic of irreversible processes driven by changes in Gibbs energy, heat production, intracellular acidity, membrane potential gradient, and ionic conductance. We focus our review on the opposing interactions observed in glioma between the canonical WNT/beta-catenin pathway and PPAR gamma and their metabolic and thermodynamic implications. In gliomas, WNT/beta-catenin pathway is upregulated while PPAR gamma is downregulated. Upregulation of WNT/beta-catenin signaling induces changes in key metabolic enzyme that modify their thermodynamics behavior. This leads to activation pyruvate dehydrogenase kinase 1(PDK-1 and monocarboxylate lactate transporter 1 (MCT-1. Consequently, phosphorylation of PDK-1 inhibits pyruvate dehydrogenase complex (PDH. Thus, a large part of pyruvate cannot be converted into acetyl-CoA in mitochondria and in TCA (tricarboxylic acid cycle. This leads to aerobic glycolysis despite the availability of oxygen, named Warburg effect. Cytoplasmic pyruvate is, in major part, converted into lactate. The WNT/beta-catenin pathway induces also the transcription of genes involved in cell proliferation, cell invasiveness, nucleotide synthesis, tumor growth, and angiogenesis, such as c-Myc, cyclin D1, PDK. In addition, in gliomas cells, PPAR gamma is downregulated, leading to a decrease in insulin sensitivity and an increase in neuroinflammation. Moreover, PPAR gamma contributes to regulate some key circadian genes. Abnormalities in the regulation of circadian rhythms and dysregulation in circadian clock genes are observed in gliomas. Circadian rhythms are dissipative structures, which play a key role in far-from-equilibrium thermodynamics through their interactions with WNT/beta-catenin pathway and PPAR gamma. In gliomas, metabolism, thermodynamics, and circadian rhythms are tightly interrelated.

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.

Characterization by Monte Carlo of the dose after a glass shield lead for gamma ray; Caracterizacion por Monte Carlo de la dosis tras un blindaje de vidrio de plomo para rayos gamma

Energy Technology Data Exchange (ETDEWEB)

Esteve Sanchez, S.; Gil Conde, M.; Contreras Gonzalez, J. L.; Rosado, J.; Pazyi, V.

2013-07-01

When a gamma-ray beam crosses the border between two media characterized by atomic number very different is they produce effects on the distribution of doses near the border difficult to predict with simple models. The case of rays gamma affecting a lead glass is particularly interesting for its application to shielding of common use. interested in studying the importance of the residual dose after the shield. (Author)

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

International Nuclear Information System (INIS)

Tukiman

2008-01-01

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

Radiation Safety of Gamma, Electron and X Ray Irradiation Facilities. Specific Safety Guide (Spanish Edition)

International Nuclear Information System (INIS)

2015-01-01

The objective of this Safety Guide is to provide recommendations on how to meet the requirements of the BSS with regard to irradiation facilities. This Safety Guide provides specific, practical recommendations on the safe design and operation of gamma, electron and X ray irradiators for use by operating organizations and the designers of these facilities, and by regulatory bodies. SCOPE. The facilities considered in this publication include five types of irradiator, whether operated on a commercial basis or for research and development purposes. This publication is concerned with radiation safety issues and not with the uses of irradiators, nor does it cover the irradiation of product or its quality management. The five types of irradiator are: - Panoramic dry source storage irradiators; - Underwater irradiators, in which both the source and the product being irradiated are under water; - Panoramic wet source storage irradiators; - Electron beam irradiation facilities, in which irradiation is performed in an area that is potentially accessible to personnel, but that is kept inaccessible during the irradiation process; - X ray irradiation facilities, in which irradiation is performed in an area that is potentially accessible to personnel, but that is kept inaccessible during the irradiation process. Consideration of non-radiation-related risks and of the benefits resulting from the operation of irradiators is outside the scope of this Safety Guide. The practices of radiotherapy and radiography are also outside the scope of this Safety Guide. Category I gamma irradiators (i.e. 'self-shielded' irradiators) are outside the scope of this Safety Guide

Progress on establishing guidelines for National Ignition Facility (NIF) experiments to extend debris shield lifetime

International Nuclear Information System (INIS)

Tobin, M.; Eder, D.; Braun, D.; MacGowan, B.

2002-01-01

The survivability of the debris shields on the National Ignition Facility (NIF) are a key factor for the affordable operation of the facility. The improvements required over Nova debris shields are described. Estimates of debris shield lifetimes in the presence of target emissions with 4-8 J/cm 2 laser fluences indicate lifetimes that may contribute unacceptably to operations costs for NIF. We are developing detailed suggested guidance for target and experiment designers for NIF to assist in minimizing the damage to, and therefore the cost of, maintaining NIF debris shields. The guidance suggests a target mass quantity that as particulate on the debris shields (300 mg) may be within current operating budgets. It also suggests the amount of material that should become shrapnel on a shot (10 mg). Finally, it suggests the level of non-volatile residue (NVR) that would threaten the sol-gel coatings on the debris shields (1 μg/cm 2 ). We review the experimentation on the Nova chamber that included measuring quantities of particulate on debris shields by element and capturing shrapnel pieces in aerogel samples mounted in the chamber. We also describe computations of X-ray emissions from a likely NIF target and the associated ablation expected from this X-ray exposure on supporting target hardware. We describe progress in assessing the benefits of a pre-shield and the possible impact on the guidance for target experiments on NIF. Plans for possible experimentation on Omega and other facilities to improve our understanding of target emissions and their impacts are discussed. Our discussion of planned future work provides a forum to invite possible collaboration with the IFE community

Shielding container for radioactive isotopes

International Nuclear Information System (INIS)

Sumi, Tetsuo; Tosa, Masayoshi; Hatogai, Tatsuaki.

1975-01-01

Object: To effect opening and closing bidirectional radiation used particularly for a gamma densimeter or the like by one operation. Structure: This device comprises a rotatable shielding body for receiving radioactive isotope in the central portion thereof and having at least two radiation openings through which radiation is taken out of the isotope, and a shielding container having openings corresponding to the first mentioned radiation openings, respectively. The radioactive isotope is secured to a rotational shaft of the shielding body, and the shielding body is rotated to register the openings of the shielding container with the openings of the shielding body or to shield the openings, thereby effecting radiation and cut off of gamma ray in the bidirection by one operation. (Kamimura, M.)

Patterns of alpha, beta and gamma diversity of the herpetofauna in Mexico’s Pacific lowlands and adjacent interior valleys

Directory of Open Access Journals (Sweden)

García, A.

2007-12-01

Full Text Available The latitudinal distribution patterns of alpha, beta and gamma diversity of reptiles, amphibians and herpetofauna were analyzed using individual binary models of potential distribution for 301 species predicted by ecological modelling for a grid of 9,932 quadrants of ~25 km2 each. We arranged quadrants in 312 latitudinal bands in which alpha, beta and gamma values were determined. Latitudinal trends of all scales of diversity were similar in all groups. Alpha and gamma responded inversely to latitude whereas beta showed a high latitudinal fluctuation due to the high number of endemic species. Alpha and gamma showed a strong correlation in all groups. Beta diversity is an important component of the herpetofauna distribution patterns as a continuous source of species diversity throughout the region.

Effect of beta and gamma neurofeedback on memory and intelligence in the elderly

NARCIS (Netherlands)

Staufenbiel, S.M.; Brouwer, A.M.; Keizer, A.W.; Wouwe, N.C. van

2014-01-01

Recent research showed a correlation between cognitive decline and a decrease of EEG gamma activity. In the present double-blind randomized control study, we investigated whether gamma and beta neurofeedback protocols, that have been shown to modulate performance on cognitive control and memory in

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.)

Neutronics issues for a laboratory microfusion facility

International Nuclear Information System (INIS)

Tobin, M.T.

1987-01-01

Discussion concerning goals or design of the Laboratory Microfusion Facility (LMF) should include an understanding of the neutronics issues involved. We consider such aspects as first wall shielding requirements, safety standards as they will apply to such an Inertial Confinement Fusion (ICF) facility, and the interior chamber environment. The selection of materials for the first wall, neutron moderator and absorber, and gamma ray shielding is discussed. We conclude that water or carbon are the choices for bulk neutron moderation and boron placed just in front of the first wall the choice for neutron absorber. Selection of the in-chamber materials and diagnostic design will greatly affect the relative hazards after a shot. Lead is the high-Z material of choice and plastic expendables for the diagnostics. Although a poor gamma ray attenuator, carbon is the choice for this function since it also compensates for the direct neutron shine effects and does not itself activate. Electronics may need to be hardened to the prompt gamma and neutron dose

Determination of gross gamma and gross beta activities in liquid effluent samples. Phase I

International Nuclear Information System (INIS)

Curtis, K.E.; Sood, S.P.

1985-08-01

Several inadequacies in the presently used procedures for gross gamma and gross beta measurements in aqueous wastes have been identified. Both the presence of suspended particulate activity and the use of cesium-137 as a calibration standard can cause gross gamma measurements to overestimate the actual activity in the sample. At the same time, sample preparation for the determination of gross beta activities causes large losses of radioiodine before the measurement step and the presence of solid material can cause a serious decrease in the beta counting efficiency. A combination of these errors could result in large discrepancies between the results obtained by the two measurement methods. Improved procedures are required to overcome these problems

Technical Safety Requirements for the Gamma Irradiation Facility (GIF)

CERN Document Server

Mahn, J A E M J G

2003-01-01

This document provides the Technical Safety Requirements (TSR) for the Sandia National Laboratories Gamma Irradiation Facility (GIF). The TSR is a compilation of requirements that define the conditions, the safe boundaries, and the administrative controls necessary to ensure the safe operation of a nuclear facility and to reduce the potential risk to the public and facility workers from uncontrolled releases of radioactive or other hazardous materials. These requirements constitute an agreement between DOE and Sandia National Laboratories management regarding the safe operation of the Gamma Irradiation Facility.

Rearrangement of beta,gamma-unsaturated esters with thallium trinitrate: synthesis of indans bearing a beta-keto ester moiety

Directory of Open Access Journals (Sweden)

Silva Jr. Luiz F.

2006-01-01

Full Text Available The rearrangement of beta,gamma-unsaturated esters, such as 2-(3,4-dihydronaphthalen-1-yl-propionic acid ethyl ester, with thallium trinitrate (TTN in acetic acid leads to 3-indan-1-yl-2-methyl-3-oxo-propionic acid ethyl ester in good yield, through a ring contraction reaction. The new indans thus obtained feature a beta-keto ester moiety, which would be useful for further functionalization.

Quantitative portable gamma-spectroscopy sample analysis for non-standard sample geometries

International Nuclear Information System (INIS)

Ebara, S.B.

1998-01-01

Utilizing a portable spectroscopy system, a quantitative method for analysis of samples containing a mixture of fission and activation products in nonstandard geometries was developed. This method was not developed to replace other methods such as Monte Carlo or Discrete Ordinates but rather to offer an alternative rapid solution. The method can be used with various sample and shielding configurations where analysis on a laboratory based gamma-spectroscopy system is impractical. The portable gamma-spectroscopy method involves calibration of the detector and modeling of the sample and shielding to identify and quantify the radionuclides present in the sample. The method utilizes the intrinsic efficiency of the detector and the unattenuated gamma fluence rate at the detector surface per unit activity from the sample to calculate the nuclide activity and Minimum Detectable Activity (MDA). For a complex geometry, a computer code written for shielding applications (MICROSHIELD) is utilized to determine the unattenuated gamma fluence rate per unit activity at the detector surface. Lastly, the method is only applicable to nuclides which emit gamma-rays and cannot be used for pure beta or alpha emitters. In addition, if sample self absorption and shielding is significant, the attenuation will result in high MDA's for nuclides which solely emit low energy gamma-rays. The following presents the analysis technique and presents verification results using actual experimental data, rather than comparisons to other approximations such as Monte Carlo techniques, to demonstrate the accuracy of the method given a known geometry and source term. (author)

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

Science.gov (United States)

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

2017-09-01

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

Benchmarking study and its application for shielding analysis of large accelerator facilities

Energy Technology Data Exchange (ETDEWEB)

Lee, Hee-Seock; Kim, Dong-hyun; Oranj, Leila Mokhtari; Oh, Joo-Hee; Lee, Arim; Jung, Nam-Suk [POSTECH, Pohang (Korea, Republic of)

2015-10-15

Shielding Analysis is one of subjects which are indispensable to construct large accelerator facility. Several methods, such as the Monte Carlo, discrete ordinate, and simplified calculation, have been used for this purpose. The calculation precision is overcome by increasing the trial (history) numbers. However its accuracy is still a big issue in the shielding analysis. To secure the accuracy in the Monte Carlo calculation, the benchmarking study using experimental data and the code comparison are adopted fundamentally. In this paper, the benchmarking result for electrons, protons, and heavy ions are presented as well as the proper application of the results is discussed. The benchmarking calculations, which are indispensable in the shielding analysis were performed for different particles: proton, heavy ion and electron. Four different multi-particle Monte Carlo codes, MCNPX, FLUKA, PHITS, and MARS, were examined for higher energy range equivalent to large accelerator facility. The degree of agreement between the experimental data including the SINBAD database and the calculated results were estimated in the terms of secondary neutron production and attenuation through the concrete and iron shields. The degree of discrepancy and the features of Monte Carlo codes were investigated and the application way of the benchmarking results are discussed in the view of safety margin and selecting the code for the shielding analysis. In most cases, the tested Monte Carlo codes give proper credible results except of a few limitation of each codes.

Benchmarking study and its application for shielding analysis of large accelerator facilities

International Nuclear Information System (INIS)

Lee, Hee-Seock; Kim, Dong-hyun; Oranj, Leila Mokhtari; Oh, Joo-Hee; Lee, Arim; Jung, Nam-Suk

2015-01-01

Shielding Analysis is one of subjects which are indispensable to construct large accelerator facility. Several methods, such as the Monte Carlo, discrete ordinate, and simplified calculation, have been used for this purpose. The calculation precision is overcome by increasing the trial (history) numbers. However its accuracy is still a big issue in the shielding analysis. To secure the accuracy in the Monte Carlo calculation, the benchmarking study using experimental data and the code comparison are adopted fundamentally. In this paper, the benchmarking result for electrons, protons, and heavy ions are presented as well as the proper application of the results is discussed. The benchmarking calculations, which are indispensable in the shielding analysis were performed for different particles: proton, heavy ion and electron. Four different multi-particle Monte Carlo codes, MCNPX, FLUKA, PHITS, and MARS, were examined for higher energy range equivalent to large accelerator facility. The degree of agreement between the experimental data including the SINBAD database and the calculated results were estimated in the terms of secondary neutron production and attenuation through the concrete and iron shields. The degree of discrepancy and the features of Monte Carlo codes were investigated and the application way of the benchmarking results are discussed in the view of safety margin and selecting the code for the shielding analysis. In most cases, the tested Monte Carlo codes give proper credible results except of a few limitation of each codes

Evaluation of beta-decay III. The complex gamma function

International Nuclear Information System (INIS)

Wilkinson, D.H.

1993-05-01

Two real, analytical, approximations for the square of the modulus of the complex gamma function as it appears in F(Z, W), the Fermi function for beta-decay, are evaluated; an accuracy bettering 10 -4 % can easily be achieved for all electron energies throughout the periodic table. (author). 3 refs., 1 tab., 7 figs

Scintillation characteristics of phosphich-detector for detection of beta- and gamma-radiations

CERN Document Server

Ananenko, A A; Gavrilyuk, V

2002-01-01

The results of the study on the influence of individual peculiarities of the compound scintillation detector structure on the value and stability of the light yield by the gamma- and beta-radiation combined registration are presented. The phosphich detector is manufactured from the sodium iodide monocrystal, activated by thallium, and the scintillation plastic on the polystyrol basis. The comparison of the experimental results with the mathematical modeling data revealed certain regularities of the process of forming the phosphich detector light signal. The recommendations are worked out by means whereof the following characteristics of the scintillation unit: the light yield and its stability, amplitude resolution and the peak-to-valley ratio by the gamma- and beta-radiation registration were improved

Standardization of low energy beta and beta-gamma complex emitters by the tracer and the efficiency extrapolation methods

International Nuclear Information System (INIS)

Sahagia, M.

1978-01-01

The absolute standardization of radioactive solutions of low energy beta emitters and beta-gamma emitters with a high probability of disintegration to the ground state is described; the tracer and the efficiency extrapolation methods were used. Both types of radionuclides were mathematically and physically treated in an unified manner. The theoretical relations between different beta spectra were calculated according to Williams' model and experimentally verified for: 35 S + 60 Co, 35 S + 95 Nb, 147 Pm + 60 Co, 14 C + 95 Nb and two beta branches of 99 Mo. The optimum range of beta efficiency variation was indicated. The basic supposition that all beta efficieny tend to unity in the same time was experimentally verified, using two 192 Ir beta branches. Four computer programs, written in the FORTRAN IV language, were elaborated, for the adequate processing of the experimental data. Good precision coefficients according to international standards were obtained in the absolute standardization of 35 S, 147 Pm, 99 Mo solutions. (author)

Calibrated radioactive sources - absolute measurements using a 4{pi} {beta}-{gamma} apparatus; Sources etalons de radioactivite - mesures absolues au moyen d'un ensemble 4{pi} {beta}-{gamma}

Energy Technology Data Exchange (ETDEWEB)

Benoit, P; Philis, C [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1965-07-01

In this paper, the principle of the standardization of the radioisotopes by the 4{pi} {beta} - {gamma} coincidence method is reminded. Some theoretical examples are given emphasizing on instrumental coincidences and their corrections. The experimental apparatus is described: one discusses the choice of the experimental conditions for the many isotopes measured. Results are given and discussed. In appendix we describe the preparation of the sources. (authors) [French] Dans ce rapport, nous rappelons d'abord le principe de la methode d'etalonnage 4{pi} {beta} - {gamma} en donnant quelques exemples theoriques et en insistant sur la correction des erreurs instrumentales. Une description critique de l'appareillage utilise est donnee. On discute le choix des conditions experimentales pour divers isotopes mesures et les resultats obtenus. En annexe, nous decrivons la fabrication des sources que nous mesurons.

Counting efficiency for radionuclides decaying by beta and gamma-ray emission; Calculo de la eficiencia de recuento de nucleidos que experimentan desintegracion beta y desexcitacion gamma simple

Energy Technology Data Exchange (ETDEWEB)

Grau, A.; Garcia-Torano, E.

1988-07-01

In this paper, counting efficiency vs figure of merit for beta and gamma-ray emitters has been computed. It is assumed that the decay scheme has only a gamma level and the beta-ray emission may be coincident with the gamma-rays or the internal-conversion electrons. The radionuclides tabulated are: 20 {sub 0}, 20{sub p}, 28{sub A}l, 35{sub p}, 41{sub A}r, 42{sub K}, 47{sub S}e, 62{sub F}e, 66{sub C}u, 81{sub G}e, 86{sub B}b, 108{sub R}u, 112{sub p}d, 121{sub S}n(Ni), 122{sub I}n, 129{sub I}, 141{sub C}e 171{sub T}m, 194{sub O}s, 2O3{sub H}g, 205{sub H}g, 210{sub p}b, 225{sub R}a, 142{sub p}r, 151{sub S}m, 244{sub A}m(m). It has been assumed that the liquid is a toluene based scintillator solution in standard glass vials containing 10 cm''3. (Author) 8 refs.

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

International Nuclear Information System (INIS)

Lee, Yoon Hee

2006-02-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Lee, Yoon Hee

2006-02-15

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

Thermodynamics in Neurodegenerative Diseases: Interplay Between Canonical WNT/Beta-Catenin Pathway-PPAR Gamma, Energy Metabolism and Circadian Rhythms.

Science.gov (United States)

Vallée, Alexandre; Lecarpentier, Yves; Guillevin, Rémy; Vallée, Jean-Noël

2018-03-23

Entropy production rate is increased by several metabolic and thermodynamics abnormalities in neurodegenerative diseases (NDs). Irreversible processes are quantified by changes in the entropy production rate. This review is focused on the opposing interactions observed in NDs between the canonical WNT/beta-catenin pathway and PPAR gamma and their metabolic and thermodynamic implications. In amyotrophic lateral sclerosis and Huntington's disease, WNT/beta-catenin pathway is upregulated, whereas PPAR gamma is downregulated. In Alzheimer's disease and Parkinson's disease, WNT/beta-catenin pathway is downregulated while PPAR gamma is upregulated. The dysregulation of the canonical WNT/beta-catenin pathway is responsible for the modification of thermodynamics behaviors of metabolic enzymes. Upregulation of WNT/beta-catenin pathway leads to aerobic glycolysis, named Warburg effect, through activated enzymes, such as glucose transporter (Glut), pyruvate kinase M2 (PKM2), pyruvate dehydrogenase kinase 1(PDK1), monocarboxylate lactate transporter 1 (MCT-1), lactic dehydrogenase kinase-A (LDH-A) and inactivation of pyruvate dehydrogenase complex (PDH). Downregulation of WNT/beta-catenin pathway leads to oxidative stress and cell death through inactivation of Glut, PKM2, PDK1, MCT-1, LDH-A but activation of PDH. In addition, in NDs, PPAR gamma is dysregulated, whereas it contributes to the regulation of several key circadian genes. NDs show many dysregulation in the mediation of circadian clock genes and so of circadian rhythms. Thermodynamics rhythms operate far-from-equilibrium and partly regulate interactions between WNT/beta-catenin pathway and PPAR gamma. In NDs, metabolism, thermodynamics and circadian rhythms are tightly interrelated.

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

International Nuclear Information System (INIS)

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

2002-01-01

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

Glasses impregnated with lead for radiation shielding

International Nuclear Information System (INIS)

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

2005-01-01

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

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

Science.gov (United States)

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

2017-05-01

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

Shielding Aspects of Accelerators, Targets and Irradiation Facilities - SATIF-11 Workshop Proceedings Report

International Nuclear Information System (INIS)

2013-01-01

Particle accelerators have evolved over the last decades from simple devices to powerful machines. In recent years, new technological and research applications have helped to define requirements while the number of accelerator facilities in operation, being commissioned, designed or planned has grown significantly. Their parameters, which include the beam energy, currents and intensities, and target composition, can vary widely, giving rise to new radiation shielding issues and challenges. Particle accelerators must be operated in safe ways to protect operators, the public and the environment. As the design and use of these facilities evolve, so must the analytical methods used in the safety analyses. These workshop proceedings review the state of the art in radiation shielding of accelerator facilities and irradiation targets. They also evaluate progress in the development of modelling methods used to assess the effectiveness of such shielding as part of safety analyses. The transport of radiation through shielding materials is a major consideration in the safety design studies of nuclear power plants, and the modelling techniques used may be applied to many other types of scientific and technological facilities. Accelerator and irradiation facilities represent a key capability in R and D, medical and industrial infrastructures, and they can be used in a wide range of scientific, medical and industrial applications. High-energy ion accelerators, for example, are now used not only in fundamental research, such as the search for new super-heavy nuclei, but also for therapy as part of cancer treatment. While the energy of the incident particles on the shielding of these facilities may be much higher than those found in nuclear power plants, much of the physics associated with the behaviour of the secondary particles produced is similar, as are the computer modelling techniques used to quantify key safety design parameters, such as radiation dose and activation levels

Circadian rhythms, Wnt/beta-catenin pathway and PPAR alpha/gamma profiles in diseases with primary or secondary cardiac dysfunction

Directory of Open Access Journals (Sweden)

Yves eLecarpentier

2014-11-01

Full Text Available Circadian clock mechanisms are far-from-equilibrium dissipative structures. Peroxisome proliferator-activated receptors (PPAR alpha, beta/delta and gamma play a key role in metabolic regulatory processes, particularly in heart muscle. Links between circadian rhythms (CRs and PPARs have been established. Mammalian CRs involve at least two critical transcription factors, CLOCK and BMAL1 (Gekakis et al., 1998; Hogenesch et al., 1998. PPAR gamma plays a major role in both glucose and lipid metabolisms and presents circadian properties which coordinate the interplay between metabolism and CRs. PPAR gamma is a major component of the vascular clock. Vascular PPAR gamma is a peripheral regulator of cardiovascular rhythms controlling circadian variations in blood pressure and heart rate through BMAL1. We focused our review on diseases with abnormalities of CRs and with primary or secondary cardiac dysfunction. Moreover, these diseases presented changes in the Wnt/beta-catenin pathway and PPARs, according to two opposed profiles. Profile 1 was defined as follows: inactivation of the Wnt/beta-catenin pathway with increased expression of PPAR gamma. Profile 2 was defined as follows: activation of the Wnt/beta-catenin pathway with decreased expression of PPAR gamma. A typical profile 1 disease is arrhythmogenic right ventricular cardiomyopathy, a genetic cardiac disease which presents mutations of the desmosomal proteins and is mainly characterized by fatty acid accumulation in adult cardiomyocytes mainly in the right ventricle. The link between PPAR gamma dysfunction and desmosomal genetic mutations occurs via inactivation of the Wnt/beta-catenin pathway presenting oscillatory properties. A typical profile 2 disease is type 2 diabetes, with activation of the Wnt/beta-catenin pathway and decreased expression of PPAR gamma. CRs abnormalities are present in numerous pathologies such as cardiovascular diseases, sympathetic/parasympathetic dysfunction

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

International Nuclear Information System (INIS)

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

2011-01-01

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

Current segmented gamma-ray scanner technology

International Nuclear Information System (INIS)

Bjork, C.W.

1987-01-01

A new generation of segmented gamma-ray scanners has been developed at Los Alamos for scrap and waste measurements at the Savannah River Plant and the Los Alamos Plutonium Facility. The new designs are highly automated and exhibit special features such as good segmentation and thorough shielding to improve performance

Xe isotope detection and discrimination using beta spectroscopy with coincident gamma spectroscopy

Science.gov (United States)

Reeder, P. L.; Bowyer, T. W.

1998-02-01

Beta spectroscopic techniques show promise of significant improvements for a beta-gamma coincidence counter that is part of a system for analyzing Xe automatically separated from air. The previously developed counting system for 131mXe, 133mXe, 133gXe, and 135gXe can be enhanced to give additional discrimination between these Xe isotopes by using the plastic scintillation sample cell as a beta spectrometer to resolve the conversion electron peaks. The automated system will be a key factor in monitoring the Comprehensive Test Ban Treaty.

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

International Nuclear Information System (INIS)

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

2016-01-01

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

DECOMMISSIONING THE BROOKHAVEN NATIONAL LABORATORY BUILDING 830 GAMMA IRRADIATION FACILITY.

Energy Technology Data Exchange (ETDEWEB)

BOWERMAN, B.S.; SULLIVAN, P.T.

2001-08-13

The Building 830 Gamma Irradiation Facility (GIF) at Brookhaven National Laboratory (BNL) was decommissioned because its design was not in compliance with current hazardous tank standards and its cobalt-60 sources were approaching the end of their useful life. The facility contained 354 stainless steel encapsulated cobalt-60 sources in a pool, which provided shielding. Total cobalt-60 inventory amounted to 24,000 Curies when the sources were shipped for disposal. The decommissioning project included packaging, transport, and disposal of the sources and dismantling and disposing of all other equipment associated with the facility. Worker exposure was a major concern in planning for the packaging and disposal of the sources. These activities were planned carefully according to ALARA (As Low As Reasonably Achievable) principles. As a result, the actual occupational exposures experienced during the work were within the planned levels. Disposal of the pool water required addressing environmental concerns, since the planned method was to discharge the slightly contaminated water to the BNL sewage treatment plant. After the BNL evaluation procedure for discharge to the sewage treatment plant was revised and reviewed by regulators and BNL's Community Advisory Council, the pool water was discharged to the Building 830 sanitary system. Because the sources were sealed and the pool water contamination levels were low, most of the remaining equipment was not contaminated; therefore disposal was straightforward, as scrap metal and construction debris.

Gamma and Beta Oscillations in Human MEG Encode the Contents of Vibrotactile Working Memory

Directory of Open Access Journals (Sweden)

Alexander H. von Lautz

2017-12-01

Full Text Available Ample evidence suggests that oscillations in the beta band represent quantitative information about somatosensory features during stimulus retention. Visual and auditory working memory (WM research, on the other hand, has indicated a predominant role of gamma oscillations for active WM processing. Here we reconciled these findings by recording whole-head magnetoencephalography during a vibrotactile frequency comparison task. A Braille stimulator presented healthy subjects with a vibration to the left fingertip that was retained in WM for comparison with a second stimulus presented after a short delay. During this retention interval spectral power in the beta band from the right intraparietal sulcus and inferior frontal gyrus (IFG monotonically increased with the to-be-remembered vibrotactile frequency. In contrast, induced gamma power showed the inverse of this pattern and decreased with higher stimulus frequency in the right IFG. Together, these results expand the previously established role of beta oscillations for somatosensory WM to the gamma band and give further evidence that quantitative information may be processed in a fronto-parietal network.

Radiological characterization of the TAN-IET facility

Energy Technology Data Exchange (ETDEWEB)

Koeppen, L.D.; Rodriguez, S.V.; Wheeler, O.A.; Cadwell, E.D.; Simpson, O.D.

1982-06-01

The Initial Engine Test (IET) facility is located on the Idaho National Engineering Laboratory (INEL) site at the north end of Test Area North (TAN). The IET facility was constructed and used for the Aircraft Nuclear Propulsion Program during the 1950's and was later used for two other programs: the Space Nuclear Auxiliary Power Transient (SNAP-TRAN) and the Hallam Decontamination and Decommissioning Project. The facility is no longer in use, therefore, a complete radiological characterization was conducted at the IET site. The characterization included measurements of beta-gamma dose rates; beta-gamma and alpha surface contamination; concentrations of selected radionuclides in subsurface storage tanks, surface soil, the exhaust duct, stack and test pad; and a walk-over surface survey of the entire facility. The information contained in this report will be of great value as the IET facility goes through the decommissioning and decontamination process.

Mass measurement project by determination of Q{sub {beta}} for neutron-rich nuclei; Projet de mesure des masses par determination des Q{sub {beta}} pour des noyaux tres riches en neutrons

Energy Technology Data Exchange (ETDEWEB)

Pautrat, M; Lagrange, J M; Petizon, L [Paris-11 Univ., 91 - Orsay (France). Inst. de Physique Nucleaire; Vanhorenbeeck, J; Duhamel, P [Brussels Univ. (Belgium). Inst. d` Astronomie et d` Astrophysique; Binon, F [Universite Libre de Bruxelles (Belgium)

1994-12-31

The aim of the project described hereafter is to collect new data on the exotic neutron rich nuclei of the Fe to Zn region, and in particular to determine their masses, for both nuclear physics and astrophysics purposes. These isotopes will be produced through projectile fragmentation at the GANIL facility and selected by the LISE3 spectrometer. Their half-lives will be measured as well as the energy of their main {gamma} rays; {gamma} - {gamma} coincidences will then allow to build a preliminary level scheme. The analysis of {beta} spectra and {beta} - {gamma} coincidences will finally provide the maximum {beta} decay energies of the studied nuclei leading to their masses. The difficulties arising from the low production rates, the {beta} detection, the data handling are discussed together with the solutions proposed to overcome them. (authors). 17 refs.

Interactions Between the Canonical WNT/Beta-Catenin Pathway and PPAR Gamma on Neuroinflammation, Demyelination, and Remyelination in Multiple Sclerosis.

Science.gov (United States)

Vallée, Alexandre; Vallée, Jean-Noël; Guillevin, Rémy; Lecarpentier, Yves

2018-05-01

Multiple sclerosis (MS) is marked by neuroinflammation and demyelination with loss of oligodendrocytes in the central nervous system. The immune response is regulated by WNT/beta-catenin pathway in MS. Activated NF-kappaB, a major effector of neuroinflammation, and upregulated canonical WNT/beta-catenin pathway positively regulate each other. Demyelinating events present an upregulation of WNT/beta-catenin pathway, whereas proper myelinating phases show a downregulation of WNT/beta-catenin pathway essential for the promotion of oligodendrocytes precursors cells proliferation and differentiation. The activation of WNT/beta-catenin pathway results in differentiation failure and impairment in remyelination. However, PI3K/Akt pathway and TCF7L2, two downstream targets of WNT/beta-catenin pathway, are upregulated and promote proper remyelination. The interactions of these signaling pathways remain unclear. PPAR gamma activation can inhibit NF-kappaB, and can also downregulate the WNT/beta-catenin pathway. PPAR gamma and canonical WNT/beta-catenin pathway act in an opposite manner. PPAR gamma agonists appear as a promising treatment for the inhibition of demyelination and the promotion of proper remyelination through the control of both NF-kappaB activity and canonical WNT/beta-catenin pathway.

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

International Nuclear Information System (INIS)

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

2010-01-01

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

Castor and Pollux - shielded cells for studying fuel treatment processes

International Nuclear Information System (INIS)

Faudot, G.; Bathellier, A.

1969-01-01

CASTOR and POLLUX, two alpha, beta, gamma cells are described in the present paper. They are located in the CEN at Fontenay-aux-Roses (France). They are designed for improvement studies of the various aqueous separation processes used in irradiated fuels reprocessing plants. Located in the same air-tight steel encasement, they arc inter-connected by a pneumatic transfer. These two cells have a similar in-line conception and they include: a gamma shielding in lead of 10 cm of thickness; an inner air-tight box, made with stainless steel and plexiglas, is maintained in lowering in comparison to room pressure. Eleven Hobson model seven master-slave manipulators allow inner manipulations. Then the inner equipment is described briefly. (author) [fr

Neutron/photon/electron shielding study for a laser-fusion facility

International Nuclear Information System (INIS)

Thompson, W.L.

1977-01-01

A Monte Carlo shielding study encompassing neutron, photon, and electron transport has been conducted for the High Energy Gas Laser Facility at the Los Alamos Scientific Laboratory. This paper describes the application of the Monte Carlo technique and several variance reduction schemes to the study. The calculations involve a geometry which is complicated in all three dimensions, a very intense 14 MeV neutron source, skyshine and deep penetrations. The facility design with 1.83 m concrete walls and a 1.52 m concrete roof is based on these calculations

Measurements and FLUKA Simulations of Bismuth, Aluminium and Indium Activation at the upgraded CERN Shielding Benchmark Facility (CSBF)

Science.gov (United States)

Iliopoulou, E.; Bamidis, P.; Brugger, M.; Froeschl, R.; Infantino, A.; Kajimoto, T.; Nakao, N.; Roesler, S.; Sanami, T.; Siountas, A.; Yashima, H.

2018-06-01

The CERN High energy AcceleRator Mixed field (CHARM) facility is situated in the CERN Proton Synchrotron (PS) East Experimental Area. The facility receives a pulsed proton beam from the CERN PS with a beam momentum of 24 GeV/c with 5·1011 protons per pulse with a pulse length of 350 ms and with a maximum average beam intensity of 6.7·1010 protons per second. The extracted proton beam impacts on a cylindrical copper target. The shielding of the CHARM facility includes the CERN Shielding Benchmark Facility (CSBF) situated laterally above the target that allows deep shielding penetration benchmark studies of various shielding materials. This facility has been significantly upgraded during the extended technical stop at the beginning of 2016. It consists now of 40 cm of cast iron shielding, a 200 cm long removable sample holder concrete block with 3 inserts for activation samples, a material test location that is used for the measurement of the attenuation length for different shielding materials as well as for sample activation at different thicknesses of the shielding materials. Activation samples of bismuth, aluminium and indium were placed in the CSBF in September 2016 to characterize the upgraded version of the CSBF. Monte Carlo simulations with the FLUKA code have been performed to estimate the specific production yields of bismuth isotopes (206 Bi, 205 Bi, 204 Bi, 203 Bi, 202 Bi, 201 Bi) from 209 Bi, 24 Na from 27 Al and 115 m I from 115 I for these samples. The production yields estimated by FLUKA Monte Carlo simulations are compared to the production yields obtained from γ-spectroscopy measurements of the samples taking the beam intensity profile into account. The agreement between FLUKA predictions and γ-spectroscopy measurements for the production yields is at a level of a factor of 2.

Beta Beams: an accelerator based facility to explore Neutrino oscillation physics

CERN Document Server

Wildner, E; Hansen, C; De Melo Mendonca, T; Stora, T; Payet, J; Chance, A; Zorin, V; Izotov, I; Rasin, S; Sidorov, A; Skalyga, V; De Angelis, G; Prete, G; Cinausero, M; Kravchuk, VL; Gramegna, F; Marchi, T; Collazuol, G; De Rosa, G; Delbar, T; Loiselet, M; Keutgen, T; Mitrofanov, S; Lamy, T; Latrasse, L; Marie-Jeanne, M; Sortais, P; Thuillier, T; Debray, F; Trophime, C; Hass, M; Hirsh, T; Berkovits, D; Stahl, A

2011-01-01

The discovery that the neutrino changes flavor as it travels through space has implications for the Standard Model of particle physics (SM)[1]. To know the contribution of neutrinos to the SM, needs precise measurements of the parameters governing the neutrino oscillations. This will require a high intensity beam-based neutrino oscillation facility. The EURONu Design Study will review three currently accepted methods of realizing this facility (the so-called Super-Beams, Beta Beams and Neutrino Factories) and perform a cost assessment that, coupled with the physics performance, will give means to the European research authorities to make a decision on the layout and construction of the future European neutrino oscillation facility. ”Beta Beams” produce collimated pure electron neutrino and antineutrino beams by accelerating beta active ions to high energies and letting them decay in a race-track shaped storage ring. EURONu Beta Beams are based on CERNs infrastructure and the fact that some of the already ...

DECONTAMINATION TECHNOLOGIES FOR FACILITY REUSE

International Nuclear Information System (INIS)

Bossart, Steven J.; Blair, Danielle M.

2003-01-01

As nuclear research and production facilities across the U.S. Department of Energy (DOE) nuclear weapons complex are slated for deactivation and decommissioning (D and D), there is a need to decontaminate some facilities for reuse for another mission or continued use for the same mission. Improved technologies available in the commercial sector and tested by the DOE can help solve the DOE's decontamination problems. Decontamination technologies include mechanical methods, such as shaving, scabbling, and blasting; application of chemicals; biological methods; and electrochemical techniques. Materials to be decontaminated are primarily concrete or metal. Concrete materials include walls, floors, ceilings, bio-shields, and fuel pools. Metallic materials include structural steel, valves, pipes, gloveboxes, reactors, and other equipment. Porous materials such as concrete can be contaminated throughout their structure, although contamination in concrete normally resides in the top quarter-inch below the surface. Metals are normally only contaminated on the surface. Contamination includes a variety of alpha, beta, and gamma-emitting radionuclides and can sometimes include heavy metals and organic contamination regulated by the Resource Conservation and Recovery Act (RCRA). This paper describes several advanced mechanical, chemical, and other methods to decontaminate structures, equipment, and materials

Site characterization of ORNL D ampersand D facilities

International Nuclear Information System (INIS)

Kelsey, A.P.; Mandry, G.J.; Haghighi, M.H.

1994-01-01

Site characterization for decontamination and decommissioning (D ampersand D) planning purposes was done for two surplus facilities at Oak Ridge National Laboratory (ORNL) in late 1993 and early 1994. This site characterization includes measurements of radiological and chemical contaminants, assessment of general structural conditions, and investigation of unknown conditions within the buildings. It will serve as input to decisions on D ampersand D engineering, D ampersand D task sequences, radiological and contamination control, and waste management. This paper presents the methods used to investigate these facilities and discusses the preliminary results as they apply to D ampersand D planning. Investigation methods include gross alpha, beta, and gamma surveys; directional gamma surveys; gamma spectroscopy; concrete coring; photography; and collection of soil and miscellaneous samples that are analyzed for radiological and chemical contaminants. Data will be analyzed using radiological models to sort sources and estimate exposure rates and waste volumes due to D ampersand D. The former Waste Evaporator Facility (WEF), consisting of two concrete cells and an operating gallery, once contained a liquid radwaste evaporator. Subsequently it was used for an incinerator experiment and as a dressing area for remediation work on an adjacent tank farm. The building has been partially decontaminated. Figure 1 is a photograph of the WEF. The Fission Product Pilot Plant (FPPP) is a small concrete building containing two cells. It was used to extract isotopes of ruthenium, strontium, cesium, cerium, and other elements from liquid waste. This facility is highly contaminated. In 1960 all doors into FPPP were sealed with concrete block and mortar, and concrete block shielding was added to the external walls making them up to five feet thick. Prior to this study, almost nothing was known about the interior of this building

Benchmark for the qualification of gamma shielding calculation methods for light-water type reactor spent fuels

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

The application of semianalytic method for calculating the thickness of biological shields of nuclear reactors. Part 2. Attenuation of gamma rays. An example of shield's thickness calculation

International Nuclear Information System (INIS)

Lukaszek, W.; Kucypera, S.

1982-01-01

The semianalytic method was used for calculating the attenuation of gamma rays and the thickness of biological shield of graphite moderated reactor. A short description of computer code as well as the exemplary results of calculations are given. (A.S.)

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

International Nuclear Information System (INIS)

Fujimoto, Kenzo; Kobayashi, Sadayoshi

1988-01-01

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

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.

Benchmark experiments of dose distributions in phantom placed behind iron and concrete shields at the TIARA facility

International Nuclear Information System (INIS)

Nakane, Yoshihiro; Sakamoto, Yukio; Tsuda, Shuichi

2004-01-01

To verify the calculation methods used for the evaluations of neutron dose at the radiation shielding design of the high-intensity proton accelerator facility (J-PARC), dose distributions in a plastic phantom of 30x30x30 cm 3 slab placed behind iron and concrete test shields were measured by using a tissue equivalent proportional counter for 65-MeV quasi-monoenergetic neutrons generated from the 7 Li(p,n) reactions with 68-MeV protons at the TIARA facility. Dose distributions in the phantom were calculated by using the MCNPX and the NMTC/JAM-MCNP codes with the flux-to-dose conversion coefficients prepared for the shielding design of the facility. The comparison results show the calculated results were in good agreement with the measured ones within 20%. (author)

MMW [multimegawatt] shielding design and analysis

International Nuclear Information System (INIS)

Olson, A.P.

1988-01-01

Reactor shielding for multimegawatt (MMW) space power must satisfy a mass constraint as well as performance specifications for neutron fluence and gamma dose. A minimum mass shield is helpful in attaining the launch mass goal for the entire vehicle, because the shield comprises about 1% to 2% of the total vehicle mass. In addition, the shield internal heating must produce tolerable temperatures. The analysis of shield performance for neutrons and gamma rays is emphasized. Topics addressed include cross section preparation for multigroup 2D S/sub n/-transport analyses, and the results of parametric design studies on shadow shield performance and mass versus key shield design variables such as cone angle, number, placement, and thickness of layers of tungsten, and shield top radius. Finally, adjoint methods are applied to the shield in order to spatially map its relative contribution to dose reduction, and to provide insight into further design optimization. 7 refs., 2 figs., 3 tabs

Effects of beta/gamma radiation on nuclear waste glasses

Energy Technology Data Exchange (ETDEWEB)

Weber, W.J. [Pacific Northwest National Lab., Richland, WA (United States)

1997-07-01

A key challenge in the disposal of high-level nuclear waste (HLW) in glass waste forms is the development of models of long-term performance based on sound scientific understanding of relevant phenomena. Beta decay of fission products is one source of radiation that can impact the performance of HLW glasses through the interactions of the emitted {beta}-particles and g-rays with the atoms in the glass by ionization processes. Fused silica, alkali silicate glasses, alkali borosilicate glasses, and nuclear waste glasses are all susceptible to radiation effects from ionization. In simple glasses, defects (e.g., non-bridging oxygen and interstitial molecular oxygen) are observed experimentally. In more complex glasses, including nuclear waste glasses, similar defects are expected, and changes in microstructure, such as the formation of bubbles, have been reported. The current state of knowledge regarding the effects of {beta}/{gamma} radiation on the properties and microstructure of nuclear waste glasses are reviewed. (author)

Beta-and gamma-turns in proteins revisited: a new set of amino acid turn-type dependent positional preferences and potentials.

Science.gov (United States)

Guruprasad, K; Rajkumar, S

2000-06-01

The number of beta-turns in a representative set of 426 protein three-dimensional crystal structures selected from the recent Protein Data Bank has nearly doubled and the number of gamma-turns in a representative set of 320 proteins has increased over seven times since the previous analysis. Beta-turns (7153) and gamma-turns (911) extracted from these proteins were used to derive a revised set of type-dependent amino acid positional preferences and potentials. Compared with previous results, the preference for proline, methionine and tryptophan has increased and the preference for glutamine, valine, glutamic acid and alanine has decreased for beta-turns. Certain new amino acid preferences were observed for both turn types and individual amino acids showed turn-type dependent positional preferences. The rationale for new amino acid preferences are discussed in the light of hydrogen bonds and other interactions involving the turns. Where main-chain hydrogen bonds of the type NH(i + 3) --> CO(i) were not observed for some beta-turns, other main-chain hydrogen bonds or solvent interactions were observed that possibly stabilize such beta-turns. A number of unexpected isolated beta-turns with proline at i + 2 position were also observed. The NH(i + 2) --> CO(i) hydrogen bond was observed for almost all gamma-turns. Nearly 20% classic gamma-turns and 43% inverse gamma-turns are isolated turns.

Period concatenation underlies interactions between gamma and beta rhythms in neocortex

Directory of Open Access Journals (Sweden)

Anita K Roopun

2008-04-01

Full Text Available The neocortex generates rhythmic electrical activity over a frequency range covering many decades. Specific cognitive and motor states are associated with oscillations in discrete frequency bands within this range, but it is not known whether interactions and transitions between distinct frequencies are of functional importance. When coexpressed rhythms have frequencies that differ by a factor of two or more interactions can be seen in terms of phase synchronization. Larger frequency differences can result in interactions in the form of nesting of faster frequencies within slower ones by a process of amplitude modulation. It is not known how coexpressed rhythms, whose frequencies differ by less than a factor of two may interact. Here we show that two frequencies (gamma – 40 Hz and beta2 – 25 Hz, coexpressed in superficial and deep cortical laminae with low temporal interaction, can combine to generate a third frequency (beta1 – 15 Hz showing strong temporal interaction. The process occurs via period concatenation, with basic rhythm-generating microcircuits underlying gamma and beta2 rhythms forming the building blocks of the beta1 rhythm by a process of addition. The mean ratio of adjacent frequency components was a constant – approximately the golden mean – which served to both minimize temporal interactions, and permit multiple transitions, between frequencies. The resulting temporal landscape may provide a framework for multiplexing – parallel information processing on multiple temporal scales.

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

International Nuclear Information System (INIS)

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

1975-12-01

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

On k-Gamma and k-Beta Distributions and Moment Generating Functions

Directory of Open Access Journals (Sweden)

Gauhar Rahman

2014-01-01

Full Text Available The main objective of the present paper is to define k-gamma and k-beta distributions and moments generating function for the said distributions in terms of a new parameter k>0. Also, the authors prove some properties of these newly defined distributions.

SATIF-2 shielding aspects of accelerators, targets and irradiation facilities

International Nuclear Information System (INIS)

1995-01-01

Particle accelerators have evolved over the last 50 years from simple devices to powerful machines, and will continue to have an important impact on research, technology and lifestyle. Today they cover a wide range of applications, from television and computer displays in households to the investigation of the origin and structure of matter. It has become common practice to use them for material science and medical applications. In recent years, requirements from new technological and research applications have emerged, such as increased particle beams intensities, higher flexibility, etc., giving rise to new radiation shielding aspects and problems. These proceedings review recent progress in radiation shielding of accelerator facilities, and evaluate advancements with respect to international co-operation in this field

The p110beta isoform of phosphoinositide 3-kinase signals downstream of G protein-coupled receptors and is functionally redundant with p110gamma.

Science.gov (United States)

Guillermet-Guibert, Julie; Bjorklof, Katja; Salpekar, Ashreena; Gonella, Cristiano; Ramadani, Faruk; Bilancio, Antonio; Meek, Stephen; Smith, Andrew J H; Okkenhaug, Klaus; Vanhaesebroeck, Bart

2008-06-17

The p110 isoforms of phosphoinositide 3-kinase (PI3K) are acutely regulated by extracellular stimuli. The class IA PI3K catalytic subunits (p110alpha, p110beta, and p110delta) occur in complex with a Src homology 2 (SH2) domain-containing p85 regulatory subunit, which has been shown to link p110alpha and p110delta to Tyr kinase signaling pathways. The p84/p101 regulatory subunits of the p110gamma class IB PI3K lack SH2 domains and instead couple p110gamma to G protein-coupled receptors (GPCRs). Here, we show, using small-molecule inhibitors with selectivity for p110beta and cells derived from a p110beta-deficient mouse line, that p110beta is not a major effector of Tyr kinase signaling but couples to GPCRs. In macrophages, both p110beta and p110gamma contributed to Akt activation induced by the GPCR agonist complement 5a, but not by the Tyr kinase ligand colony-stimulating factor-1. In fibroblasts, which express p110beta but not p110gamma, p110beta mediated Akt activation by the GPCR ligands stromal cell-derived factor, sphingosine-1-phosphate, and lysophosphatidic acid but not by the Tyr kinase ligands PDGF, insulin, and insulin-like growth factor 1. Introduction of p110gamma in these cells reduced the contribution of p110beta to GPCR signaling. Taken together, these data show that p110beta and p110gamma can couple redundantly to the same GPCR agonists. p110beta, which shows a much broader tissue distribution than the leukocyte-restricted p110gamma, could thus provide a conduit for GPCR-linked PI3K signaling in the many cell types where p110gamma expression is low or absent.

Beta-gamma counting system for Xe fission products

International Nuclear Information System (INIS)

Reeder, P.L.; Bowyer, T.W.; Perkins, R.W.

1998-01-01

A beta-gamma coincidence counting system has been developed for automated analysis of Xe gas samples separated from air. The Xe gas samples are contained in a cylindrical plastic scintillator cell located between two NaI(Tl) scintillation detectors. The X-ray and gamma spectra gated by coincident events in the plastic scintillator cell are recorded for each NaI(Tl) crystal. The characteristic signatures of the 131m Xe, 133g Xe, 133m Xe, and 135g Xe isotopes of interest for nuclear test-ban verification as well as the procedures and results of absolute efficiency measurements are described. A NaI(Tl) crystal with provision for 4 sample cells has been implemented for the system to be deployed in the field. Examples of data on ambient air samples in New York City obtained with the field prototype are presented. (author)

Evaluation of Shielding Wall Optimization in Lead Slowing Down Spectrometer System

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-15

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

Nuclear Rocket Test Facility Decommissioning Including Controlled Explosive Demolition of a Neutron-Activated Shield Wall

International Nuclear Information System (INIS)

Michael Kruzic

2007-01-01

Located in Area 25 of the Nevada Test Site, the Test Cell A Facility was used in the 1960s for the testing of nuclear rocket engines, as part of the Nuclear Rocket Development Program. The facility was decontaminated and decommissioned (D and D) in 2005 using the Streamlined Approach For Environmental Restoration (SAFER) process, under the Federal Facilities Agreement and Consent Order (FFACO). Utilities and process piping were verified void of contents, hazardous materials were removed, concrete with removable contamination decontaminated, large sections mechanically demolished, and the remaining five-foot, five-inch thick radiologically-activated reinforced concrete shield wall demolished using open-air controlled explosive demolition (CED). CED of the shield wall was closely monitored and resulted in no radiological exposure or atmospheric release

New detection modules for gamma, beta and X-ray cameras

International Nuclear Information System (INIS)

Azman, S.; Bolle, E.; Dang, K.Q.; Dang, W.; Dietzel, K.I.; Froberg, T.; Gaarder, P.E.; Gjaerum, J.A.; Haugen, S.H.; Hellum, G.; Henriksen, J.R.; Johanson, T.M.; Kobbevik, A.; Maehlum, G.; Meier, D.; Mikkelsen, S.; Ninive, I.; Oya, P.; Pavlov, N.; Pettersen, D.M.; Sundal, B.M.; Talebi, J.; Yoshioka, K.

2003-01-01

Full text: Ideas ASA is developing new detection modules for gamma, beta and X-ray cameras. Recent developments focus on modules using various semi-conductor materials (CZT, HgI, Si). The development includes ASIC design, detector module development, and implementation in camera heads. In this presentation we describe the characteristics of important ASICs and its properties in terms of electronic noise, and the modes for measuring signals (switched current modes, sparsified modes, self triggered modes). The ASICs are specific for detectors and applications. We describe recent developments using various semi - conductor materials. We describe important design aspects for medical applications and in life science (SPECT, beta, X-ray cameras)

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)

Beta decay of the fission product 125Sb and a new complete evaluation of absolute gamma ray transition intensities

Science.gov (United States)

Rajput, M. U.; Ali, N.; Hussain, S.; Mujahid, S. A.; MacMahon, D.

2012-04-01

The radionuclide 125Sb is a long-lived fission product, which decays to 125Te by negative beta emission with a half-life of 1008 day. The beta decay is followed by the emission of several gamma radiations, ranging from low to medium energy, that can suitably be used for high-resolution detector calibrations, decay heat calculations and in many other applications. In this work, the beta decay of 125Sb has been studied in detail. The complete published experimental data of relative gamma ray intensities in the beta decay of the radionuclide 125Sb has been compiled. The consistency analysis was performed and discrepancies found at several gamma ray energies. Evaluation of the discrepant data was carried out using Normalized Residual and RAJEVAL methods. The decay scheme balance was carried out using beta branching ratios, internal conversion coefficients, populating and depopulating gamma transitions to 125Te levels. The work has resulted in the consistent conversion factor equal to 29.59(13) %, and determined a new evaluated set of the absolute gamma ray emission probabilities. The work has also shown 22.99% of the delayed intensity fraction as outgoing from the 58 d isomeric 144 keV energy level and 77.01% of the prompt intensity fraction reaching to the ground state from the other excited states. The results are discussed and compared with previous evaluations. The present work includes additional experimental data sets which were not included in the previous evaluations. A new set of recommended relative and absolute gamma ray emission probabilities is presented.

Beta decay of the fission product 125Sb and a new complete evaluation of absolute gamma ray transition intensities

International Nuclear Information System (INIS)

Rajput, M.U.; Ali, N.; Hussain, S.; Mujahid, S.A.; MacMahon, D.

2012-01-01

The radionuclide 125 Sb is a long-lived fission product, which decays to 125 Te by negative beta emission with a half-life of 1008 day. The beta decay is followed by the emission of several gamma radiations, ranging from low to medium energy, that can suitably be used for high-resolution detector calibrations, decay heat calculations and in many other applications. In this work, the beta decay of 125 Sb has been studied in detail. The complete published experimental data of relative gamma ray intensities in the beta decay of the radionuclide 125 Sb has been compiled. The consistency analysis was performed and discrepancies found at several gamma ray energies. Evaluation of the discrepant data was carried out using Normalized Residual and RAJEVAL methods. The decay scheme balance was carried out using beta branching ratios, internal conversion coefficients, populating and depopulating gamma transitions to 125 Te levels. The work has resulted in the consistent conversion factor equal to 29.59(13) %, and determined a new evaluated set of the absolute gamma ray emission probabilities. The work has also shown 22.99% of the delayed intensity fraction as outgoing from the 58 d isomeric 144 keV energy level and 77.01% of the prompt intensity fraction reaching to the ground state from the other excited states. The results are discussed and compared with previous evaluations. The present work includes additional experimental data sets which were not included in the previous evaluations. A new set of recommended relative and absolute gamma ray emission probabilities is presented.

Screening procedures for alpha, beta and gamma emitting radionuclides in urine

International Nuclear Information System (INIS)

Kramer, G.H.; Gardner, S.E.; Johnson, J.R.

1983-08-01

The gross actinide and beta screening methods currently used by the Bioassay Laboratory at Chalk River Nuclear Laboratories have been documented and tested. The gross gamma method has also been documented. The gross actinide method was tested with Th-nat, Pu-239, Np-237, Am-241 and Cm-244 and the average percent recoveries were 90, 99, 45, 72 and 78, respectively. The gross beta method was tested with Sr-90/Y-90, Ce-144/Pr-144, Pm-147 and Ba-133 and the average percent recoveries were 78, 100, 46 and 46, respectively. Detection limits for the gross actinide (alpha) method are found to be about 0.4 mBq and about 39 mBq for the gross beta method

Characteristics of beta detection and dose measurement at Department of Energy facilities

International Nuclear Information System (INIS)

Mulvehill, J.M.; Brackenbush, L.W.

1987-02-01

This report considers the current state of the art of beta dosimetry practices and beta detection methods used by health physicists at US Department of Energy facilities. This information is based on a survey of DOE facilities. Beta measurements are technically difficult and innovative efforts must be expended to improve their accuracy. Perhaps the most pronounced problem is that beta dosimetry and instrumentation in use are highly energy and angular dependent. Many believe that beta exposures are adequately controlled because beta to photon ratios are assumed to be low. This assumption is not always valid as demonstrated by the accident at Three Mile Island (TMI). Significant beta doses exist where personnel are exposed to mixed fission products; for example, chemical reprocessing plants, reactor accidents, or where uranium metals are processed. This report is part of an effort to increase the DOE response to this technically difficult area of health protection. Problem areas are addressed and methods recommended to improve beta dosimetry through a cooperative effort among the various DOE contractors. 34 refs., 2 figs., 16 tabs

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-07-01

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

Important changes in medical x-ray imaging facility shielding design methodology. A brief summary of recommendations in NCRP Report No. 147

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

Design of the magnetized muon shield for the prompt-neutrino facility

International Nuclear Information System (INIS)

Baltay, C.; Bosek, N.; Couch, J.

1982-01-01

The main technical challenge in the design of the prompt neutrino beam is the magnetized muon shield. Two satisfactory alternate designs have been developed for such a shield during this past year and the background muon fluxes have been calculated by three independent programs at Columbia, Fermilab, and MIT. The background muon fluxes have been calculated to be satisfactory in all of the detectors that might use the beam. In Section III of this report we describe in detail the three Monte Carlo programs used in these calculations. In Section IV we give the details of the flux calculations for the E-613 shield and the comparisons with the observed fluxes with various configurations of that shield. In Section V we describe the designs that have been developed for the neutrino area shield. In Section VI we discuss the problem of proton beam transport losses and the associated muon fluxes. Finally, in Section VII a comparison of the two solutions is made which covers cost, effectiveness, schedule and responsiveness to future unknowns. We conclude that there are not overwhelming reasons for the choice of one design over the other. However, for a variety of secondary reasons the superconducting design offers advantages. We therefore propose the construction of the prompt neutrino facility with the superconducting magnet design

3-dimensional shielding design for a spallation neutron source facility in the high-intensity proton accelerator project

Energy Technology Data Exchange (ETDEWEB)

Tamura, Masaya; Maekawa, Fujio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2003-03-01

Evaluation of shielding performance for a 1 MW spallation neutron source facility in the Materials and Life Science Facility being constructed in the High-Intensity Proton Accelerator Project (J-PARC) is important from a viewpoint of radiation safety and optimization of arrangement of components. This report describes evaluated results for the shielding performance with modeling three-dimensionally whole structural components including gaps between them in detail. A Monte Carlo calculation method with MCNPX2.2.6 code and LA-150 library was adopted. Streaming and void effects, optimization of shield for cost reduction and optimization of arrangement of structures such as shutters were investigated. The streaming effects were investigated quantitatively by changing the detailed structure of components and gap widths built into the calculation model. Horizontal required shield thicknesses were ranged from about 6.5 m to 7.5 m as a function of neutron beam line angles. A shutter mechanism for a horizontal neutron reflectometer that was directed downward was devised, and it was shown that the shielding performance of the shutter was acceptable. An optimal biological shield configuration was finally determined according to the calculated results. (author)

Evaluation of bulk shield for the JHP facilities

International Nuclear Information System (INIS)

Uwamino, Yoshitomo; Shibata, Tokushi

1991-01-01

In the Japanese Hadron Project (JHP), a 1-GeV 200-μA proton beam will be handled, and the radiation shield of the facility will be very massive concrete and iron lump. Since the constructing cost is strongly affected by the shielding design, the design must be severely performed. The neutron yield in thin targets and a copper beam dump was calculated by the HETC-KFA-2 Monte Carlo code. For the evaluation of the calculational accuracy, the calculational results were compared with the experimental data by Cierjacks and Raupp. The calculated result of heavy element agreed well with the experiment at a low energy region, E n n >100 MeV) of 90 deg close to the calculated one of about 60 deg in the absolute value. The high energy neutron transport in a 5-m-thick iron slab and in an 8-m-thick ordinary concrete slab was calculated with the HETC code and also with the discrete ordinates transport code, ANISN. In the ANISN calculation, the DLC-87/HILO and the DLC-128/LAHIMAC group cross sections were used. The ANISN calculation with the LAHIMAC cross sections gave strong underestimation compared with the HETC calculation. The difference of the shielding lengths calculated by the HETC code and by the ANISN code used with the HILO cross sections was smaller than 6% for the both iron and concrete cases. (author)

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

International Nuclear Information System (INIS)

Diop, C.M'B.

1984-01-01

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

Radiation shielding activities at IDOM

Energy Technology Data Exchange (ETDEWEB)

Ordóñez, César Hueso; Gurpegui, Unai Cano; Valiente, Yelko Chento; Poveda, Imanol Zamora, E-mail: cesar.hueso@idom.com [IDOM, Consulting, Engineering and Architecture, S.A.U, Vizcaya (Spain)

2017-07-01

When human activities have to be performed under ionising radiation environments the safety of the workers must be guaranteed. Usually three principles are used to accomplish with ALARA (As Low As Reasonably Achievable) requirements: the more distance between the source term and the worker, the better; the less time spent to arrange any task, the better; and, once the previous principles are optimized should the exposure of the workers continues being above the regulatory limits, shielding has to be implemented. Through this paper some different examples of IDOM's shielding design activities are presented. Beginning with the gamma collimators for the Jules Horowitz Reactor, nuclear fuel's behaviour researching facility, where the beam path crosses the reactor's containment walls and is steered up to a gamma detector where the fuel spectrum is analysed and where the beam has to be attenuated several orders of magnitude in a short distance. Later it is shown IDOM’s approach for the shielding of the Emergency Control Management Center of Asociación Nuclear Ascó-Vandellòs-II NPPs, a bunker designed to withstand severe accident conditions and to support the involved staff during 30 days, considering the outside radioactive cloud and the inside source term that filtering units become as they filter the incoming air. And finally, a general approach to this kind of problems is presented, since the study of the source term considering all the possible contributions, passing through the material selection and the thicknesses calculation until the optimization of the materials. (author)

Radiation shielding activities at IDOM

International Nuclear Information System (INIS)

Ordóñez, César Hueso; Gurpegui, Unai Cano; Valiente, Yelko Chento; Poveda, Imanol Zamora

2017-01-01

When human activities have to be performed under ionising radiation environments the safety of the workers must be guaranteed. Usually three principles are used to accomplish with ALARA (As Low As Reasonably Achievable) requirements: the more distance between the source term and the worker, the better; the less time spent to arrange any task, the better; and, once the previous principles are optimized should the exposure of the workers continues being above the regulatory limits, shielding has to be implemented. Through this paper some different examples of IDOM's shielding design activities are presented. Beginning with the gamma collimators for the Jules Horowitz Reactor, nuclear fuel's behaviour researching facility, where the beam path crosses the reactor's containment walls and is steered up to a gamma detector where the fuel spectrum is analysed and where the beam has to be attenuated several orders of magnitude in a short distance. Later it is shown IDOM’s approach for the shielding of the Emergency Control Management Center of Asociación Nuclear Ascó-Vandellòs-II NPPs, a bunker designed to withstand severe accident conditions and to support the involved staff during 30 days, considering the outside radioactive cloud and the inside source term that filtering units become as they filter the incoming air. And finally, a general approach to this kind of problems is presented, since the study of the source term considering all the possible contributions, passing through the material selection and the thicknesses calculation until the optimization of the materials. (author)

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

International Nuclear Information System (INIS)

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

2007-01-01

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

Preliminary Shielding Assessment for the IFF System in the RAON Heavy-ion Facility

International Nuclear Information System (INIS)

Lee, Cheol Woo; Lee, Youngouk; Kim, Jong Won; Kim, Mijung

2014-01-01

A heavy-ion accelerator facility is under a development in Korea to use in the basic science research and various application areas. In this facility, the In-Flight Fragment (IFF) target and isotope separator has been designed to produce various isotopes and transport the interesting isotopes into the experimental rooms. In this work, preliminary radiation shielding assessment was performed for the IFF target room

Proceedings of the International Symposium Advances in alpha, Beta- and Gamma-Ray spectrometry

International Nuclear Information System (INIS)

1997-01-01

The International Committee for Radionuclide Metrology (ICRM) is an association of radionuclide metrology laboratories whose membership is composed of delegates of these laboratories together with other scientists actively engaged in the study and applications of radioactivity. The scientific activities are carried out in the frame of six Working Groups. Two of them, the Alpha-Particle Spectrometry and the Gamma-and Beta-ray Spectrometry Working Groups held a common workshop in Pushkin, St. Petersburg, 18 to 20 September 1996, under the title Advances in Alpha-Beta-and Gamma-Ray Sepectrometry, at the kind invitation of the D.I. Mendeleyev Institute for Metrology. More than 30 people from 14 laboratories attended the meeting, and nineteen oral communications were presented, from which twelve were retained for publication an are included in these proceedings. (Author)

Study of radiation detectors response in standard X, gamma and beta radiation standard beams; Estudo da resposta de monitores de radioprotecao em feixes padronizados de radiacao X, gama e beta

Energy Technology Data Exchange (ETDEWEB)

Nonato, Fernanda Beatrice Conceicao

2010-07-01

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

Study of radiation detectors response in standard X, gamma and beta radiation standard beams; Estudo da resposta de monitores de radioprotecao em feixes padronizados de radiacao X, gama e beta

Energy Technology Data Exchange (ETDEWEB)

Nonato, Fernanda Beatrice Conceicao

2010-07-01

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

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

International Nuclear Information System (INIS)

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

2012-01-01

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

Effect of beta and gamma neurofeedback on memory and intelligence in the elderly.

Science.gov (United States)

Staufenbiel, S M; Brouwer, A-M; Keizer, A W; van Wouwe, N C

2014-01-01

Recent research showed a correlation between cognitive decline and a decrease of EEG gamma activity. In the present double-blind randomized control study, we investigated whether gamma and beta neurofeedback protocols, that have been shown to modulate performance on cognitive control and memory in young adults, also leads to increased brain activity and cognitive performance in elderly. Twenty older adults either performed eight 30-min gamma neurofeedback session or beta neurofeedback session within a period of 21 days. Cognitive performance was determined before and after the training through an IQ and memory task and we added a subjective well-being questionnaire. Both neurofeedback training protocols resulted in a significant increase of the brain activity within each training session, suggesting that the aging brain is still trainable. However, we found no effects on cognitive performance or transfer of the feedback beyond the trainings. We discuss several possible reasons for the lack of training on rest measurements and cognition and ways to improve the feedback protocols for future studies. Copyright © 2013 Elsevier B.V. All rights reserved.

Technical Requirements for Fabrication and Installation of Removable Shield for CNRF in HANARO

Energy Technology Data Exchange (ETDEWEB)

Ryu, Jeong Soo; Cho, Yeong Garp; Lee, Jung Hee; Shin, Jin Won

2008-04-15

This report details the technical requirements for the fabrication and installation of the removable shield for the Cold Neutron Research Facility (CNRF) in HANARO reactor hall. The removable shield is classified as non-nuclear safety (NNS), seismic category II, and quality class T. The main function of the removable shield is to do the biological shielding of neutrons and gamma from the CN port and the guides. The removable shield consists of block type walls and roofs that can be necessarily assembled, disassembled and moveable. These will be installed between the reactor pool wall and the CNS guide bunker in. This report describes technical requirements for the removable shield such as quality assurance, seismic analysis requirements, configuration, concrete compositions, fabrication and installation requirements, test and inspection, shipping, delivery, etc. Appendix is the technical specification of structural design and analysis. Attachments are composed of the technical specification for the fabrication of the removable shield, shielding design drawings and procurement quality requirements. These technical requirements will be provided to a contract for the manufacturing and installation.

Change in the aldolase activity in rats' hearts after irradiation with. gamma. and. beta. rays

Energy Technology Data Exchange (ETDEWEB)

Kukulyanskaya, M F; Borodina, G N

1973-01-01

The activity of aldolase fructoso-1-phosphate (I) and aldolase fructoso-1, 6-diphosphate (II) has been studied at various periods after total ..gamma.. and ..beta.. irradiation of rats at a dose of 42 rads. It has been shown that after ..gamma.. irradiation the activity of I increases in the supernatant liquid of the heart muscle homogenate, but drops sharply in the nuclei. In total, the activity of the homogenate did not change. The activity of II dropped for seven days, and after 1 hour and 30 days it was above the normal level. After ..beta.. radiation the activity of II is slowed down after 1, 7, and 15 days. These changes occur in all subcellular fractions. The authors note that the changes in the activity of aldolases are more sharply demarcated after ..gamma.. irradiation and are more substantial for II. (JPRS)

Standardization of 56Co had been carried out using 4 pi beta-gamma coincidence methods

International Nuclear Information System (INIS)

Wardiyanto, Gatot; Pujadi

2000-01-01

Standardization of exp.56 Co had been carried out using 4 pi beta-gamma coincidence methods. The radionuclide use for calibration of nuclear instruments on range of energy over 1500 keV. The exp.56 Co had been produced by irradiation of proton by using a cyclotron with 15 MeV of energy and 300 mb of cross-section to natural iron target (99,5% of purity) at the Institute for Nuclear Study, University of Tokyo. Source preparation had been done by gravimetry method after the irradiated source was dissolved in 8N HCI solution. The disintegration rate had been measured using 4 pi beta-gamma coincidence apparatus, where the gamm gets sets on 511 and 847 keV gamma-rays. The result measurement is fairly good with the specific activity is 3078 n 15 Bq/mg

Bulk-shield design for the Fusion Materials Irradiation Test facility

International Nuclear Information System (INIS)

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

1982-07-01

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

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

International Nuclear Information System (INIS)

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

2007-01-01

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

A study on radiation shield design of storage facility for low and intermediate level radioactive waste in Bangladesh

International Nuclear Information System (INIS)

Khan, JJahirul Haque

2005-02-01

Bangladesh has no nuclear power reactor but has only one 3 MW TRIGA Mark-II Research Reactor. The Bangladesh Atomic Energy commission (BAEC) operates a 3 MW TRIGA Mark-II Research Reactor and maintains not only the nuclear facilities at its Atomic Energy Research Establishment (AERE) at Savar (near Dhaka) but also the related radiation facilities the whole country. The main sources of radioactive wastes result from the use of sealed and unsealed radiation sources in medicine industry, research, agriculture, etc as well as from operation and maintenance of the nuclear facilities the whole country. As a result radioactive wastes are increasing day by day and these wastes are classified as low and intermediate level radioactive waste (LILW) following the radiation safety philosophy of IAEA recommendations in Bangladesh. Radioactive waste is very sensitive issue to public and environment from the hazardous standpoint of ionizing radiation. Therefore, storage facility of LILW is very essential for safe radioactive waste management in Bangladesh and in parallel: this study is of a great importance due to new installation of this storage facility in future. The basic objective of this study is to recommend the radiation shield design parameters of the installation of storage facility for low and intermediate level radioactive waste from the points of view of radiation safety and sensitivity analysis. The shield design of this installation has been carried out with the Monte Carlo Code MCNP4C and the point Kernel Code Micro Shield 5.05 respectively considering the ICRP-60 (1990) recommendations for occupational exposure limit (10 μ Sv/hr). For more safety purpose every equivalent dose rate at different positions of this installation is considered below 9 μ Sv/hr in this study. The radiation shield design parameters are recommended based on MCNP4C calculated results than those of Micro Shield due to more credible results and these parameters are: (I) 51 cm thickness of

Monte Carlo simulation of {beta}-{gamma} coincidence system using plastic scintillators in 4{pi} geometry

Energy Technology Data Exchange (ETDEWEB)

Dias, M.S. [Instituto de Pesquisas Energeticas e Nucleares: IPEN-CNEN/SP, Av. Prof. Lineu Prestes 2242, 05508-000 Sao Paulo, SP (Brazil)], E-mail: msdias@ipen.br; Piuvezam-Filho, H. [Instituto de Pesquisas Energeticas e Nucleares: IPEN-CNEN/SP, Av. Prof. Lineu Prestes 2242, 05508-000 Sao Paulo, SP (Brazil); Baccarelli, A.M. [Departamento de Fisica-PUC/SP-Rua Marques de Paranagua 111, 01303-050 Sao Paulo, SP (Brazil); Takeda, M.N. [Universidade Santo Amaro, UNISA-Rua Prof. Eneas da Siqueira Neto 340, 04829-300 Sao Paulo, SP (Brazil); Koskinas, M.F. [Instituto de Pesquisas Energeticas e Nucleares: IPEN-CNEN/SP, Av. Prof. Lineu Prestes 2242, 05508-000 Sao Paulo, SP (Brazil)

2007-09-21

A modified version of a Monte Carlo code called Esquema, developed at the Nuclear Metrology Laboratory in IPEN, Sao Paulo, Brazil, has been applied for simulating a 4{pi}{beta}(PS)-{gamma} coincidence system designed for primary radionuclide standardisation. This system consists of a plastic scintillator in 4{pi} geometry, for alpha or electron detection, coupled to a NaI(Tl) counter for gamma-ray detection. The response curves for monoenergetic electrons and photons have been calculated previously by Penelope code and applied as input data to code Esquema. The latter code simulates all the disintegration processes, from the precursor nucleus to the ground state of the daughter radionuclide. As a result, the curve between the observed disintegration rate as a function of the beta efficiency parameter can be simulated. A least-squares fit between the experimental activity values and the Monte Carlo calculation provided the actual radioactive source activity, without need of conventional extrapolation procedures. Application of this methodology to {sup 60}Co and {sup 133}Ba radioactive sources is presented and showed results in good agreement with a conventional proportional counter 4{pi}{beta}(PC)-{gamma} coincidence system.

Radiation shielding and health physics instrumentation for PET medical cyclotrons

International Nuclear Information System (INIS)

Mukherjee, B.

2002-01-01

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

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)

Industrial gamma irradiation facility with a wet storage source in Syrian Arab Republic

International Nuclear Information System (INIS)

Othman, I.; Moussa, A.; Stepanov, D.G.; Ermakov, V.

1998-01-01

A gamma radiation facility was built in Damascus, Syria. The plant (ROBO) is a Co-60 wet storage, batch/continuous facility with nominal capacity of 1.85x10 16 Bq. The initial activity is 3.7x10 15 Bq. The ratio of maximum absorbed dose to the minimum one within irradiated materials is around 1.3+/-0.03. The irradiator consists of two sections to select required sources for irradiation. Two pools were constructed. The main pool will serve as biological shield for the main sources frame. The second pool will host a fixed circular frame to be used as calibration source or to irradiate small samples to low doses. The conveyor consists of a chain facility moving along trucks. A repair section is provided on the conveyor route in the load-unload area for carrying out inspection, repair, etc. The trucks are holed with a rectangular frames. Loading, unloading and rearrangement of the products is carried out automatically. This mechanism is carried out by seven pneumatic cylinders, lifting devices and roller conveyors. Many safety features were included: push-back platform, followed by pit used as a physical barrier. Interlocks are connected to the platform, pit cover and to ionization chambers. In case of power failure or any overriding of interlocks, the irradiator comes to emergency dropping. Ventilation system, fire system, emergency power and closed water purification system are indicated on control panel. The facility will be utilized for medical products sterilization, research and calibration

Food irradiation: Gamma processing facilities

Energy Technology Data Exchange (ETDEWEB)

Kunstadt, P. [MDS Nordion International, 447 March Road. Kanata, Ontario, K2K148 (Canada)

1997-12-31

The number of products being radiation processed is constantly increasing and today include such diverse items as medical disposable, fruits and vegetables, bulk spices, meats, sea foods and waste effluents. Not only do the products differ but also many products, even those within the same groupings, require different minimum and maximum radiation doses. These variations create many different requirements in the irradiator design. The design of Cobalt-60 radiation processing facilities is well established for a number of commercial applications. Installations in over 40 countries, with some in operation since the early 1960s, are testimony to the fact that irradiator design, manufacture, installation and operation is a well established technology. However, in order to design gamma irradiators for the preservation of foods one must recognize those parameters typical to the food irradiation process as well as those systems and methods already well established in the food industry. This paper discusses the basic design concepts for gamma food irradiators. They are most efficient when designed to handle a limited product density range at an established dose. Safety of Cobalt-60 transport, safe facility operation principles and the effect of various processing parameters on economics, will also be discussed. (Author)

Food irradiation: Gamma processing facilities

Energy Technology Data Exchange (ETDEWEB)

Kunstadt, P [MDS Nordion International, 447 March Road. Kanata, Ontario, K2K148 (Canada)

1998-12-31

The number of products being radiation processed is constantly increasing and today include such diverse items as medical disposable, fruits and vegetables, bulk spices, meats, sea foods and waste effluents. Not only do the products differ but also many products, even those within the same groupings, require different minimum and maximum radiation doses. These variations create many different requirements in the irradiator design. The design of Cobalt-60 radiation processing facilities is well established for a number of commercial applications. Installations in over 40 countries, with some in operation since the early 1960s, are testimony to the fact that irradiator design, manufacture, installation and operation is a well established technology. However, in order to design gamma irradiators for the preservation of foods one must recognize those parameters typical to the food irradiation process as well as those systems and methods already well established in the food industry. This paper discusses the basic design concepts for gamma food irradiators. They are most efficient when designed to handle a limited product density range at an established dose. Safety of Cobalt-60 transport, safe facility operation principles and the effect of various processing parameters on economics, will also be discussed. (Author)

Food irradiation: Gamma processing facilities

International Nuclear Information System (INIS)

Kunstadt, P.

1997-01-01

The number of products being radiation processed is constantly increasing and today include such diverse items as medical disposable, fruits and vegetables, bulk spices, meats, sea foods and waste effluents. Not only do the products differ but also many products, even those within the same groupings, require different minimum and maximum radiation doses. These variations create many different requirements in the irradiator design. The design of Cobalt-60 radiation processing facilities is well established for a number of commercial applications. Installations in over 40 countries, with some in operation since the early 1960s, are testimony to the fact that irradiator design, manufacture, installation and operation is a well established technology. However, in order to design gamma irradiators for the preservation of foods one must recognize those parameters typical to the food irradiation process as well as those systems and methods already well established in the food industry. This paper discusses the basic design concepts for gamma food irradiators. They are most efficient when designed to handle a limited product density range at an established dose. Safety of Cobalt-60 transport, safe facility operation principles and the effect of various processing parameters on economics, will also be discussed. (Author)

Seed irradiation facilities at TRIGA Mark II reactor

International Nuclear Information System (INIS)

Najzer, M.

1972-01-01

Fast neutrons and gamma-rays with their high and low LET respectively are excellent complementary tools for investigation of the effect of different types of mutations. TRIGA Irradiation Facility and Thermal Column Irradiation Facility were designed and installed for the first time in the TRIGA tank and thermal column respectively. The basic idea of design was the use of depleted uranium as gamma-ray and thermal neutron shield and simultaneously as thermal to fast neutron converter. Low LET radiation, due to direct and thermal neutron capture gamma-rays, is strongly attenuated while fast neutron flux is increased. GIF is made of a cadmium tube inserted in a graphite block. It is located in the central thermal column channel. The basic idea is to convert thermal neutrons to gamma-rays by capture in the cadmium

Contribution to the study of {beta} disintegration and of nuclear structure using experiments on certain {beta}-{gamma} cascades: 198{sub Au}, 86{sub Rb}, 170{sub Tm}; Contribution a l'etude de la desintegration beta et a l'etude de la structure nucleaire a l'aide d'experiences sur certaines cascades beta-gamma: 198{sub Au}, 86{sub Rb}, 170{sub Tm}

Energy Technology Data Exchange (ETDEWEB)

Lachkar, J [Commissariat a l' Energie Atomique, Bruyeres-le-Chatel (France). Centre d' Etudes; Paris-11 Univ., fabulte des Sciences 91 - Orsay (France)

1969-07-01

{beta}{gamma} directional angular correlations and shapes of inner beta spectra leading to the first excited level of the final nucleus enable one to determine the nuclear matrix elements typical of the {beta} transition. In the three observed first forbidden cases: {sup 170}Tm, {sup 86}Rb, {sup 198}Au, these matrix elements do not confirm the independent shell model theory. Other hypotheses are then suggested and discussed. (author) [French] Les experiences de correlation angulaire {beta}{gamma} et l'etude du spectre {beta} conduisant au premier niveau excite du noyau final permettent de determiner les elements de matrices nucleaires caracteristiques de cette transition. Dans les trois cas etudies (transitions une fois interdites): {sup 170}Tm, {sup 86}Rb, {sup 198}Au, ces elements de matrices ne peuvent etre retrouves a l'aide du modele en couches et a particules independantes. D'autres hypotheses sont alors emises et discutees. (auteur)

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

OpenAIRE

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

2014-01-01

The mass attenuation coefficients, μ/ρ, half-value layer, HVL, tenth-value layer, TVL, effective atomic numbers, ZPIeff, and effective electron densities, Ne,eff, of borate glass sample systems of (100-x-y) Na2B4O7 : xPbO : yNiO (where x and y=0, 2, 4, 6, 8, and 10 weight percentage) containing PbO and NiO, with potential gamma ray and neutron shielding applications, have been investigated. The gamma ray interaction parameters, μ/ρ, HVL, TVL, ZPIeff, and Ne,eff, were computed for photon energ...

Radiological protection issues during primary filter housing replacement in a high alpha, beta-gamma shielded facility at Dounreay

International Nuclear Information System (INIS)

Mowat, D.; Watkin, T.

2006-01-01

Dounreay, on the north coast of Scotland, was home to the United Kingdom Fast Breeder Reactor (F.B.R.) development programme. F.B.R. use excess, non-moderated ('fast') neutrons to convert (breed) uranium, in elements positioned at the outer edges of the reactor core, into plutonium which can then be used as fuel.Site construction began in 1955 and three reactors were built and operated; the Dounreay Materials Test Reactor (D.M.T.R.) 1958-1969, the Dounreay Fast Reactor (D.F.R.) 1959-1977 and the Prototype Fast Reactor (P.F.R.) 1974-1994. The D.F.R. was conventionally fuelled by highly enriched uranium whereas the P.F.R. used a ceramic form of plutonium oxide (PuO 2 ) as its fuel. Dounreay was almost entirely self-sufficient in that a fuel cycle (chemical reprocessing) area was constructed complete with recovery plants, laboratories, waste storage and other support services buildings. Liquid plutonium nitrate product was sent to Sellafield, in Cumbria, to convert for future use, and the fuel elements were then fabricated at Springfields. Eventually, recovered P.F.R. plutonium was loaded back into the reactor, closing the fuel cycle. As a matter of interest for this paper, a Post Irradiation Examination (P.I.E.) facility, D2001, was built in the early 1960' s. The Plant was equipped with a suite of ten north and south side cells built to a high level of containment within which irradiated P.F.R. fuel could be remotely disassembled and examined. This work supported the continuing development of F.B.R. design and technology and the Plant has operated very successfully throughout its lifetime. A programme of improvement was implemented to enhance reliability, productivity and to modernize the facility to meet current nuclear and engineering standards. The experience of this work is detailed in this paper. (N.C.)

A Monte Carlo modeling alternative for the API Gamma Ray Calibration Facility.

Science.gov (United States)

Galford, J E

2017-04-01

The gamma ray pit at the API Calibration Facility, located on the University of Houston campus, defines the API unit for natural gamma ray logs used throughout the petroleum logging industry. Future use of the facility is uncertain. An alternative method is proposed to preserve the gamma ray API unit definition as an industry standard by using Monte Carlo modeling to obtain accurate counting rate-to-API unit conversion factors for gross-counting and spectral gamma ray tool designs. Copyright © 2017 Elsevier Ltd. All rights reserved.

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

International Nuclear Information System (INIS)

Hanslik, R.

2006-08-01

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

Preliminary shielding estimates for the proposed Oak Ridge National Laboratory (ORNL) Radioactive Ion Beam Facility (RIBF)

International Nuclear Information System (INIS)

Johnson, J.O.; Gabriel, T.A.; Lillie, R.A.

1996-01-01

The Oak Ridge National Laboratory (ORNL) has proposed designing and implementing a new target-ion source for production and injection of negative radioactive ion beams into the Hollifield tandem accelerator. This new facility, referred to as the Radioactive Ion Beam Facility (RIBF), will primarily be used to advance the scientific communities' capabilities for performing state-of-the-art cross-section measurements. Beams of protons or other light, stable ions from the Oak Ridge Isochronous Cyclotron (ORIC) will be stopped in the RIBF target ion source and the resulting radioactive atoms will be ionized, charge exchanged, accelerated, and injected into the tandem accelerator. The ORIC currently operates with proton energies up to 60 MeV and beam currents up to 100 microamps with a maximum beam power less than 2.0 kW. The proposed RIBF will require upgrading the ORIC to generate proton energies up to 200 MeV and beam currents up to 200 microamps for optimum performance. This report summarizes the results of a preliminary one-dimensional shielding analysis of the proposed upgrade to the ORIC and design of the RIBF. The principal objective of the shielding analysis was to determine the feasibility of such an upgrade with respect to existing shielding from the facility structure, and additional shielding requirements for the 200 MeV ORIC machine and RIBF target room

Development of Point Kernel Shielding Analysis Computer Program Implementing Recent Nuclear Data and Graphic User Interfaces

International Nuclear Information System (INIS)

Kang, Sang Ho; Lee, Seung Gi; Chung, Chan Young; Lee, Choon Sik; Lee, Jai Ki

2001-01-01

In order to comply with revised national regulationson radiological protection and to implement recent nuclear data and dose conversion factors, KOPEC developed a new point kernel gamma and beta ray shielding analysis computer program. This new code, named VisualShield, adopted mass attenuation coefficient and buildup factors from recent ANSI/ANS standards and flux-to-dose conversion factors from the International Commission on Radiological Protection (ICRP) Publication 74 for estimation of effective/equivalent dose recommended in ICRP 60. VisualShield utilizes graphical user interfaces and 3-D visualization of the geometric configuration for preparing input data sets and analyzing results, which leads users to error free processing with visual effects. Code validation and data analysis were performed by comparing the results of various calculations to the data outputs of previous programs such as MCNP 4B, ISOSHLD-II, QAD-CGGP, etc

$\\beta$3$p$-spectroscopy and proton-$\\gamma$ width determination in the decay of $^{31}$Ar

CERN Multimedia

We propose to perform a detailed study of the $\\beta$-decay of the dripline nucleus $^{31}$Ar. This will allow a detailed study of the $\\beta$-delayed 3$p$-decay as well as provide important information on the resonances of $^{30}$S and $^{29}$P, in particular the ratio between the $p$- and $\\gamma$- partial widths relevant for astrophysics.

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

Science.gov (United States)

Ripin, Azuhar; Mohamed, Faizal; Aman, Asyraf

2018-04-01

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

Alpha-Gamma Hot-Cell Facility at Argonne National Laboratory East

International Nuclear Information System (INIS)

Neimark, L.A.; Jackson, W.D.; Donahue, D.A.

1979-01-01

The Alpha-Gamma Hot-Cell Facility has been in operation at Argonne National Laboratory East (ANL-E) for 15 years. The facility was designed for plutonium research in support of ANL's LMFBR program. The facility consists of a kilocurie, nitrogen-atmosphere alpha-gamma hot cell and supporting laboratories. Modifications to the facility and its equipment have been made over the years as the workload and nature of the work changed. These modifications included inerting the entire hot cell, adding four work stations, modifying in-loading procedures and examination equipment to handle longer test articles, and changing to a new sodium-vapor lighting system. Future upgrading includes the addition of a decontamination and repair facility, use of radio-controlled transfer carts, refurbishment of the zinc bromide windows, and the installation of an Auger microprobe

Primary 4{pi}{beta}-{gamma} coincidence system for standardization of radionuclides by means of plastic scintillators; Sistema primario por coincidencias 4{pi}{beta}-{gamma} para a padronizacao de radionuclideos empregando cintiladores plasticos

Energy Technology Data Exchange (ETDEWEB)

Baccarelli, Aida Maria

2003-07-01

The present work describes a 4{pi}({alpha},{beta})-{gamma} coincidence system for absolute measurement of radionuclide activity using a plastic scintillator in 4{pi} geometry for charged particles detection and a Nal (Tl) crystal for gamma-ray detection. Several shapes and dimensions of the plastic scintillator have been tried in order to obtain the best system configuration. Radionuclides which decay by alpha emission, {beta}{sup -}, {beta}{sup +} and electron capture have been standardized. The results showed excellent agreement with other conventional primary system which makes use of a 4{pi} proportional counter for X-ray and charged particle detection. The system developed in the present work have some advantages when compared with the conventional systems, namely; it does not need metal coating on the films used as radioactive source holders. When compared to liquid scintillators, is showed the advantage of not needing to be kept in dark for more than 24 h to allow phosphorescence decay of ambient light. Therefore it can be set to count immediately after the sources are placed inside of it. (author)

UTN's gamma irradiation facility: design and concept

International Nuclear Information System (INIS)

Mohamad Noor Mohamad Yunus

1986-01-01

UTN is building a multipurpose gamma irradiation facility which compromises of research and pilot scale irradiation cells in The Fifth Malaysia Plan. The paper high-lights the basic futures of the facility in terms of its design and selection including layout sketches. Plant performances and limitations are discussed. Plants safety is briefly highlighted in block diagrams. Lastly, a typical specification brief is tabled in appendix for reference purposes. (author)

Hot-cell shielding system for high power transmission in DUPIC fuel fabrication

International Nuclear Information System (INIS)

Kim, K.; Lee, J.; Park, J.; Yang, M.; Park, H.

2000-01-01

This paper presents a newly designed hot-cell shielding system for use in the development of DUPIC (Direct Use of spent PWR fuel In CANDU reactors) fuel at KAERI (Korea Atomic Energy Research Institute). This hot-cell shielding system that was designed to transmit high power to sintering furnace in-cell from the out-of-cell through a thick cell wall has three subsystems - a steel shield plug with embedded spiral cooling line, stepped copper bus bars, and a shielding lead block. The dose-equivalent rates of the hot-cell shielding system and of the apertures between this system and the hot-cell wall were calculated. Calculated results were compared with the allowable dose limit of the existing hot-cell. Experiments for examining the temperature changes of the shielding system developed during normal furnace operation were also carried out. Finally, gamma-ray radiation survey experiments were conducted by Co-60 source. It is demonstrated that, from both calculated and experimental results, the newly designed hot-cell shielding system meets all the shielding requirements of the existing hot-cell facility, enabling high power transmission to the in-cell sintering furnace. (author)

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

International Nuclear Information System (INIS)

Diop, Cheikh M'Backe.

1980-09-01

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

Perturbed beta-gamma systems and complex geometry

Energy Technology Data Exchange (ETDEWEB)

Zeitlin, Anton M. [Department of Mathematics, Yale University, 442 Dunham Lab, 10 Hillhouse Avenue, New Haven, CT 06511 (United States)], E-mail: anton.zeitlin@yale.edu

2008-05-11

We consider the equations, arising as the conformal invariance conditions of the perturbed curved beta-gamma system. These equations have the physical meaning of Einstein equations with a B-field and a dilaton on a Hermitian manifold, where the B-field 2-form is imaginary and proportional to the canonical form associated with Hermitian metric. We show that they decompose into linear and bilinear equations and lead to the vanishing of the first Chern class of the manifold where the system is defined. We discuss the relation of these equations to the generalized Maurer-Cartan structures related to BRST operator. Finally we describe the relations of the generalized Maurer-Cartan bilinear operation and the Courant/Dorfman brackets.

Radiation shielding for neutron guides

International Nuclear Information System (INIS)

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

2005-01-01

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

Radiation shielding for neutron guides

International Nuclear Information System (INIS)

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

2006-01-01

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

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)

Utilization of radiation facilities at TNRC for shielding researches and related topics

Energy Technology Data Exchange (ETDEWEB)

Akki, T S [Physics Department, Nuclear Physics and Radiation Shielding Division Tajura Nuclear Research Center, Tripoli (Libyan Arab Jamahiriya)

1997-12-31

This paper presents the running shielding research activities at Tajura Nuclear research center. The main area of researches are concentrated on the investigation of different types of concrete made from local materials such as conventional concrete, Magnetite-Limonite concrete, and heat resistant concrete. The measuring techniques used were neutron-gamma spectrometry, and activation foils. The measurements were performed using collimated beam of reactor neutrons emitted from one of the horizontal channels, as well as from californium-252 neutron source. The transmitted neutron spectra through concrete barriers of different thicknesses were measured by a scintillation spectrometer with NE-213 liquid organic scintillator. A non-destructive testing of some reactor materials were also carried out using neutron and gamma ray computerized tomography technique (CT). Some experiments were also carried out related to measurements of neutron depth dose distributions inside tissue equivalent materials. 10 figs.

Application of Monte Carlo method in study of the padronization for radionuclides with complex disintegration scheme in 4{pi}{beta}-{gamma} coincidence System; Aplicacao do metodo de Monte Carlo no estudo da padronizacao de radionuclideos com esquema de desintegracao complexos em sistema de coincidencias 4{pi}{beta}-{gamma}

Energy Technology Data Exchange (ETDEWEB)

Takeda, Mauro Noriaki

2006-07-01

The present work described a new methodology for modelling the behaviour of the activity in a 4{pi}{beta}-{gamma} coincidence system. The detection efficiency for electrons in the proportional counter and gamma radiation in the NaI(Tl) detector was calculated using the Monte Carlo program MCNP4C. Another Monte Carlo code was developed which follows the path in the disintegration scheme from the initial state of the precursor radionuclide, until the ground state of the daughter nucleus. Every step of the disintegration scheme is sorted by random numbers taking into account the probabilities of all {beta}{sup -} branches, electronic capture branches, transitions probabilities and internal conversion coefficients. Once the final state was reached beta, electronic capture events and gamma transitions are accounted for the three spectra: beta, gamma and coincidence variation in the beta efficiency was performed simulating energy cut off or use of absorbers (Collodion). The selected radionuclides for simulation were: {sup 134}Cs, {sup 72}Ga which disintegrate by {beta}{sup -} transition, {sup 133}Ba which disintegrates by electronic capture and {sup 35}S which is a beta pure emitter. For the latter, the Efficiency Tracing technique was simulated. The extrapolation curves obtained by Monte Carlo were filled by the Least Square Method with the experimental points and the results were compared to the Linear Extrapolation method. (author)

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

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

Science.gov (United States)

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

2018-01-01

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

Radiation shielding analysis

International Nuclear Information System (INIS)

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

1982-01-01

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

Study of the Nuclear Structure of 39P Using Beta-Delayed Gamma Spectroscopy

Science.gov (United States)

Abromeit, Brittany; NSCL Experiment E14063 Team Team

2016-03-01

Investigation of nuclei with neutron and proton imbalance is at the forefront of nuclear physics research today. This is driven by the fact that the structure in these regimes may vary with that seen near the valley of stability. With eight neutrons more than the stable isotope of phosphorous, 39P is a neutron-rich exotic nucleus that has very limited information on it: previous studies of 39P produce only three known energy levels and gamma rays. The fragmentation of a 48Ca primary beam on a 564mg/cm2 thick Be target at the National Superconducting Cyclotron Laboratory (NSCL) was used to produce exotic 39Si. Using the NSCL Beta Counting System (BCS), consisting of a thick planner germanium double-sided strip detector (GeDSSD) and 16 High-purity germanium detectors in an array, SeGA, the beta-gamma coincidences from the decay of 39Si to 39P were analyzed. The resulting level scheme of 39P, including over 12 new gamma rays and energy states, confirmation of the previously measured half-life, and first-time logft values will be presented. This work was supported by the NSF under Grant No. 1401574.

Development of Neutron Shielding Material for Cask and Accelerator

International Nuclear Information System (INIS)

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

2008-01-01

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

Study on bulk shielding for a spallation neutron source facility in the high-intensity proton accelerator project

CERN Document Server

Maekawa, F; Takada, H; Teshigawara, M; Watanabe, N

2002-01-01

Under the JAERI-KEK High-Intensity Proton Accelerator Project, a spallation neutron source driven by a 3 GeV-1 MW proton beam is planed to be constructed in a main part of the Materials and Life Science Facility. This report describes results of a study on bulk shielding performance of a biological shield for the spallation neutron source by means of a Monte Carlo calculation method, that is important in terms of radiation safety and cost reduction. A shielding configuration was determined as a reference case by considering preliminary studies and interaction with other components, then shielding thickness that was required to achieve a target dose rate of 1 mu Sv/h was derived. Effects of calculation conditions such as shielding materials and dimensions on the shielding performance was investigated by changing those parameters. By taking all the results and design margins into account, a shielding configuration that was identified as the most appropriate was finally determined as follows. An iron shield regi...

Evaluation of the performance of the shields in the EPMAs used for radioactive samples

International Nuclear Information System (INIS)

Matsui, Hiroki; Suzuki, Miho; Obata, Hiroki; Kanazawa, Hiroyuki

2014-06-01

The Reactor Fuel Examination Facility (RFEF) in Japan Atomic Energy Agency (JAEA) has been used for Post Irradiation Examinations (PIEs) to verify the reliability and safety of the nuclear fuels irradiated in commercial reactors. EPMA (Electron Probe Micro Analyzer) has been utilized for the qualitative analysis of the fission product in the fuel pellet and the detailed observation of the oxide layers formed at the inner and outer surfaces of fuel cladding. Commercial EPMAs were remodeled so that the EPMAs can be applied for radioactive samples. Several shields were set in the EPMA to avoid the gamma-rays which radiate from a radioactive sample to the proportional counter in the EPMA. It is important to calculate this shielding performance adequately to maintain the precision of analysis. This report describes the results of re-evaluation of the performance of the shields in the EPMAs in the RFEF by using the Particle and Heavy Ion Transport Code System (PHITS) and the examination results of gamma-ray effect to the X-ray spectrum data by using a radioactive sample. (author)

The Portuguese gamma irradiation facility

International Nuclear Information System (INIS)

Mendes, C.M.; Almeida, J.C.; Botelho, M.L.; Cavaco, M.C.; Almeida-Vara, E.; Andrade, M.E.

1990-01-01

A Gamma Radiation Facility was built up in the National Laboratory of Industrial Technology and Engineering (LNETI), Lisbon, Portugal. This plant (UTR GAMA-Pi) is a Cobalt-60 dry storage continuous facility with a nominal capacity of 1.5 x 10 16 Bq. The initial activity is 1.1 x 10 16 Bq and the throughput capacity 10 3 ton/year for product with a bulk density of 0.2 g/cm 3 treated with a minimum absorbed dose of 25 kGy. Complementary control devices were installed: ventilation system, closed water refrigeration circuit, internal TV system, detection and extinction fire system and emergency power group. It must be emphasized that the best attention was given to the conception and efficiency of the interlock safety systems. This facility will be utilized mainly for radiosterilization of medical articles and decontamination of wine cork stoppers. (author)

Three-dimensional coupled Monte Carlo-discrete ordinates computational scheme for shielding calculations of large and complex nuclear facilities

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)

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

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

International Nuclear Information System (INIS)

Dhang, Prosenjit; Verma, Rishi; Shyam, Anurag

2015-01-01

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

Designing shields for KeV photons with genetic algorithms

International Nuclear Information System (INIS)

Asbury, Stephen; Holloway, James P.

2011-01-01

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

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)

On the importance of low-energy beta-beams for supernova neutrino physics

International Nuclear Information System (INIS)

Jachowicz, N.; McLaughlin, G.C.

2005-01-01

Beta beams, which are neutrino beams produced by the beta decay of nuclei that have been accelerated to high gamma factor, were original proposed for high energy applications, such as the measurement of the third neutrino mixing angle θ 13 . Volpe suggested that a beta beam run at lower gamma factor, would be useful for neutrino measurements in the tens of MeV range. We suggest to exploit the flexibility these beta beam facilities offer, combined with the fact that beta-beam neutrino energies overlap with supernova-neutrino energies, to construct 'synthetic' spectra that approximate an incoming supernova-neutrino energy-distribution. Using these constructed spectra we are able to reproduce total and differential folded supernova-neutrino cross-sections very accurately. We illustrate this technique using Deuterium, 16 O, and 208 Pb. This technique provides an easy and straightforward way to apply the results of a beta-beam neutrino-nucleus measurement to the corresponding supernova-neutrino detector, virtually eliminating potential uncertainties due to nuclear-structure calculations. (author)

Detector Background Reduction by Passive and Active Shielding

International Nuclear Information System (INIS)

Bikit, I.; Bikit, K.; Forkapic, S.; Mrda, D.; Nikolov, J.; Slivka, J.; Todorovic, N.

2013-01-01

The operational problems of the gamma ray spectrometer shielded passively with 12 cm of lead and actively by five 0.5 m × 0.5 m × 0.05 m plastic veto shields are described. The active shielding effect from both environmental gamma ray, cosmic muons and neutrons was investigated. For anticoincidence gating wide range of scintillator pulses, corresponding to the energy range of 150 keV-75 MeV, were used. With the optimal set up the integral background, for the energy region of 50 - 3000 keV, of 0.31 c/s was achieved. The detector mass related background was 0.345 c/(kg s). The 511 keV annihilation line was reduced by the factor of 7 by the anticoincidence gate. It is shown that the plastic shields increase the neutron capture gamma line intensities due to neutron termalization.(author)

Copper(II) 12-metallacrown-4 complexes of alpha-, beta- and gamma-aminohydroxamic acids: a comparative thermodynamic study in aqueous solution.

Science.gov (United States)

Tegoni, Matteo; Remelli, Maurizio; Bacco, Dimitri; Marchiò, Luciano; Dallavalle, Francesco

2008-05-28

A complete thermodynamic study of the protonation and Cu(II) complex formation equilibria of a series of alpha- and beta-aminohydroxamic acids in aqueous solution was performed. The thermodynamic parameters obtained for the protonation of glycine-, (S)-alpha-alanine-, (R,S)-valine-, (S)-leucine-, beta-alanine- and (R)-aspartic-beta-hydroxamic acids were compared with those previously reported for gamma-amino- and (S)-glutamic-gamma-hydroxamic acids. The enthalpy/entropy parameters calculated for the protonation microequilibria of these three types of ligands are in very good agreement with the literature values for simple amines and hydroxamic acids. The pentanuclear complexes [Cu5L4H(-4)]2+ contain the ligands acting as (NH2,N-)-(O,O-) bridging bis-chelating and correspond to 12-metallacrown-4 (12-MC-4) which are formed by self-assembly between pH 4 and 6 with alpha-aminohydroxamates (HL), while those with beta- and gamma-derivatives exist in a wider pH range (4-11). The stability order of these metallomacrocycles is beta- > alpha- > gamma-aminohydroxamates. The formation of 12-MC-4 with alpha-aminohydroxamates is entropy-driven, and that with beta-derivatives is enthalpy-driven, while with gamma-GABAhydroxamate both effects occur. These results are interpreted on the basis of specific enthalpies or entropy contributions related to chelate ring dimensions, charge neutralization and solvation-desolvation effects. The enthalpy/entropy parameters of 12-MC-4 with alpha-aminohydroxamic acids considered are also dependent on the optical purity of the ligands. Actually, that with (R,S)-valinehydroxamic acid presents an higher entropy and a lower enthalpy value than those of enantiopure ligands, although the corresponding stabilities are almost equivalent. Moreover, DFT calculations are in agreement with a more exothermic enthalpy found for metallacrowns with enantiomerically pure ligands.

Spatial interpolation of gamma dose in radioactive waste storage facility

Science.gov (United States)

Harun, Nazran; Fathi Sujan, Muhammad; Zaidi Ibrahim, Mohd

2018-01-01

External radiation measurement for a radioactive waste storage facility in Malaysian Nuclear Agency is a part of Class G License requirement under Atomic Licensing Energy Board (AELB). The objectives of this paper are to obtain the distribution of radiation dose, create dose database and generate dose map in the storage facility. The radiation dose measurement is important to fulfil the radiation protection requirement to ensure the safety of the workers. There are 118 sampling points that had been recorded in the storage facility. The highest and lowest reading for external radiation recorded is 651 microSv/hr and 0.648 microSv/hour respectively. The calculated annual dose shows the highest and lowest reading is 1302 mSv/year and 1.3 mSv/year while the highest and lowest effective dose reading is 260.4 mSv/year and 0.26 mSv/year. The result shows that the ALARA concept along time, distance and shield principles shall be adopted to ensure the dose for the workers is kept below the dose limit regulated by AELB which is 20 mSv/year for radiation workers. This study is important for the improvement of planning and the development of shielding design for the facility.

Field measurement and interpretation of beta doses and dose rates

International Nuclear Information System (INIS)

Selby, J.M.; Swinth, K.L.; Hooker, C.D.; Kenoyer, J.L.

1983-01-01

A wide variety of portable survey instruments employing GM, ionization chamber and scintillation detectors exist for the measurement of gamma exposure rates. Often these same survey instruments are used for monitoring beta fields. This is done by making measurements with and without a removable shield which is intended to shield out the non-penetrating component (beta) of the radiation field. The difference does not correspond to an absorbed dose rate for the beta field due to a variety of factors. Among these factors are the dependence on beta energy, source-detector geometries, mixed fields and variable ambient conditions. Attempting to use such measurements directly can lead to errors as high as a factor of 100. In many instances correction factors have been derived, that if properly applied, can reduce these errors substantially. However, this requires some knowledge of the beta spectra, calibration techniques and source geometry. This paper discusses some aspects of the proper use of instruments for beta measurements including the application of appropriate correction factors. Ionization type instruments are commonly used to measure beta dose rates. Through design and calibration these instruments will give an accurate reading only for uniform irradiation of the detection volume. Often in the field it is not feasible to meet these conditions. Large area uniform distributions of activity are not generally encountered and it is not possible to use large source-to-detector distances due to beta particle absorption in air. An example of correction factors required for various point sources is presented when a cutie pie ionization chamber is employed. The instrument reading is multiplied by the appropriate correction factor to obtain the dose rate at the window. When a different detector is used or for other geometries, a different set of correction factors must be used

Improvements at the biological shielding of BNCT research facility in the IEA-R1 reactor

International Nuclear Information System (INIS)

Souza, Gregorio Soares de

2011-01-01

The technique of neutron capture in boron is a promising technique in cancer treatment, it uses the high LET particles from the reaction 10 B (n, α) 7 Li to destroy cancer cells.The development of this technique began in the mid-'50s and even today it is the object of study and research in various centers around the world, Brazil has built a facility that aims to conduct research in BNCT, this facility is located next to irradiation channel number three at the research nuclear reactor IEA-R1 and has a biological shielding designed to meet the radiation protection standards. This biological shielding was developed to allow them to conduct experiments with the reactor at maximum power, so it is not necessary to turn on and off the reactor to irradiate samples. However, when the channel is opened for experiments the background radiation in the experiments salon increases and this background variation makes it impossible to perform measurements in a neutron diffraction research that utilizes the irradiation channel number six. This study aims to further improve the shielding in order to minimize the variation of background making it possible to perform the research facility in BNCT without interfering with the action of the research group of the irradiation channel number six. To reach this purpose, the code MCNP5, dosimeters and activation detectors were used to plan improvements in the biological shielding. It was calculated with the help of the code an improvement that can reduce the average heat flow in 71.2% ± 13 and verified experimentally a mean reduce of 70 ± 9% in dose due to thermal neutrons. (author)

FPDCYS and FPSPEC: computer programs for calculating fission-product beta and gamma multigroup spectra from ENDF/B-IV data

International Nuclear Information System (INIS)

Stamatelatos, M.G.; England, T.R.

1977-05-01

FPDCYS and FPSPEC are two FORTRAN computer programs used at the Los Alamos Scientific Laboratory (LASL), in conjunction with the CINDER-10 program, for calculating cumulative fission-product beta and/or gamma multigroup spectra in arbitrary energy structures, and for arbitrary neutron irradiation periods and cooling times. FPDCYS processes ENDF/B-IV fission-product decay energy data to generate multigroup beta and gamma spectra from individual ENDF/B-IV fission-product nuclides. FPSPEC further uses these spectra and the corresponding nuclide activities calculated by the CINDER-10 code to produce cumulative beta and gamma spectra in the same energy grids in which FPDCYS generates individual isotope decay spectra. The code system consisting of CINDER-10, FPDCYS, and FPSPEC has been used for comparisons with experimental spectra and continues to be used at LASL for generating spectra in special user-oriented group structures. 3 figures

Measurements of Neutron and Gamma Attenuation in Massive Laminated Shields of Concrete and a Study of the Accuracy of some Methods of Calculation

Energy Technology Data Exchange (ETDEWEB)

Aalto, E; Nilsson, R

1964-09-15

Extensive neutron and gamma attenuation measurements have been performed in magnetite and ordinary concrete up to a depth of 2 metres in order to study the accuracy attainable by some shield calculation methods. The effect of thin, heavy layers (Pb) has also been studied. Experimental facilities and instrumentation, especially the foil detection methods used for thermal and epithermal neutrons, are described in some detail. Great weight is laid upon a thorough error analysis. The fluxes measured are compared to those calculated by an earlier version of the British 18-group removal method (RASH B{sub 3}), by an improved removal method (NRN) developed at AB Atomenergi, and by numerical integration of the Boltzmann equation (NIOBE). The results show that shielding calculations with the newer methods give fluxes that are generally within a factor of 2-3 from the true values. A greater accuracy seems to be difficult to obtain in practice in spite of possible improvements in the mathematical solution of the transport problem. The greatest errors originate in the translation between the true and calculation geometries in the uncertainty of material properties in the case of concrete, and in approximations and inaccuracies of radiation sources.

Measurements of Neutron and Gamma Attenuation in Massive Laminated Shields of Concrete and a Study of the Accuracy of some Methods of Calculation

International Nuclear Information System (INIS)

Aalto, E.; Nilsson, R.

1964-09-01

Extensive neutron and gamma attenuation measurements have been performed in magnetite and ordinary concrete up to a depth of 2 metres in order to study the accuracy attainable by some shield calculation methods. The effect of thin, heavy layers (Pb) has also been studied. Experimental facilities and instrumentation, especially the foil detection methods used for thermal and epithermal neutrons, are described in some detail. Great weight is laid upon a thorough error analysis. The fluxes measured are compared to those calculated by an earlier version of the British 18-group removal method (RASH B 3 ), by an improved removal method (NRN) developed at AB Atomenergi, and by numerical integration of the Boltzmann equation (NIOBE). The results show that shielding calculations with the newer methods give fluxes that are generally within a factor of 2-3 from the true values. A greater accuracy seems to be difficult to obtain in practice in spite of possible improvements in the mathematical solution of the transport problem. The greatest errors originate in the translation between the true and calculation geometries in the uncertainty of material properties in the case of concrete, and in approximations and inaccuracies of radiation sources

Neutron and gamma characterization within the FFTF reactor cavity

International Nuclear Information System (INIS)

Bunch, W.L.; Carter, L.L.; Moore, F.S.; Werner, E.J.; Wilcox, A.D.; Wood, M.R.

1980-08-01

Neutron and gamma ray measurements were made within the reactor cavity of the Fast Flux Test Facility (FFTF) to establish the operating characteristics of the Ex-Vessel Flux Monitoring (EVFM) system as a function of reactor power level. A significant effort was made to obtain absolute flux values in order that the measurements could be compared directly with shield design calculations. Good agreement was achieved for neutrons and for both the prompt and delayed components of the gamma ray field. 8 figures, 3 tables

Measurement of Q{sub {beta}} values of neutron-rich Tc to Pd isotopes in the mass range A=110 to A=117; Messung von Q{sub {beta}}-Werten neutronenreicher Tc- bis Pd-Isotope im Massenbereich A=110 bis A=117

Energy Technology Data Exchange (ETDEWEB)

Kloeckl, Ingo

2008-06-15

The present work describes the measuring of Q{sub {beta}} values of {beta}-instable isotopes of Tc, Ru, Rh and Pd. The mass range A=110 to 117 comprises neutron-rich, short-living isotopes. Due to their small (fission) abundances, few data are known, especially regarding level schemes or gamma radiation. The proton-induced fission and a fast online mass separation was used to produce these nuclides in the IGISOL facility located in Jyvaeskylaein Finland. The {beta},{gamma},X coincidence apparatus used during the experiments allows measuring Q{sub {beta}} values as well as {gamma},X coincidences. The latter represent the basic input data for a calculation of Q{sub {beta}} values out of {beta},{gamma} coincidences. It is so possible to examine nuclides with incomplete level schemes; similarly, these level schemes can be extended using beta,gamma coincidence data. Twelve Q{sub {beta}} values of neutron-rich Tc to Pd isotopes could be determined, yielding nuclear masses, mass defects and neutron separation energies. Eight of them were determined for the first time; another one could be confirmed. For three more, the error of earlier values could be decreased by a factor of nearly ten. The resulting data are of interest for the review of nuclear mass models, they represent also input in astrophysical network calculations. (orig.)

ORNL fusion reactor shielding integral experiments

International Nuclear Information System (INIS)

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

1980-01-01

Integral experiments that measure the neutron and gamma-ray energy spectra resulting from the attenuation of approx. 14 MeV T(D,n) 4 He reaction neutrons in laminated slabs of stainless steel type 304, borated polyethylene, and a tungsten alloy (Hevimet) and from neutrons streaming through a 30-cm-diameter iron duct (L/D = 3) imbedded in a concrete shield have been performed. The facility, the NE-213 liquid scintillator detector system, and the experimental techniques used to obtain the measured data are described. The two-dimensional discrete ordinates radiation transport codes, calculational models, and nuclear data used in the analysis of the experiments are reviewed

Beta and gamma oscillatory activities associated with olfactory memory tasks: different rhythms for different functional networks?

Science.gov (United States)

Martin, Claire; Ravel, Nadine

2014-01-01

Olfactory processing in behaving animals, even at early stages, is inextricable from top down influences associated with odor perception. The anatomy of the olfactory network (olfactory bulb, piriform, and entorhinal cortices) and its unique direct access to the limbic system makes it particularly attractive to study how sensory processing could be modulated by learning and memory. Moreover, olfactory structures have been early reported to exhibit oscillatory population activities easy to capture through local field potential recordings. An attractive hypothesis is that neuronal oscillations would serve to "bind" distant structures to reach a unified and coherent perception. In relation to this hypothesis, we will assess the functional relevance of different types of oscillatory activity observed in the olfactory system of behaving animals. This review will focus primarily on two types of oscillatory activities: beta (15-40 Hz) and gamma (60-100 Hz). While gamma oscillations are dominant in the olfactory system in the absence of odorant, both beta and gamma rhythms have been reported to be modulated depending on the nature of the olfactory task. Studies from the authors of the present review and other groups brought evidence for a link between these oscillations and behavioral changes induced by olfactory learning. However, differences in studies led to divergent interpretations concerning the respective role of these oscillations in olfactory processing. Based on a critical reexamination of those data, we propose hypotheses on the functional involvement of beta and gamma oscillations for odor perception and memory.

Beta and gamma oscillatory activities associated with olfactory memory tasks: Different rhythms for different functional networks?

Directory of Open Access Journals (Sweden)

Claire eMartin

2014-06-01

Full Text Available Olfactory processing in behaving animals, even at early stages, is inextricable from top down influences associated with odor perception. The anatomy of the olfactory network (olfactory bulb, piriform and entorhinal cortices and its unique direct access to the limbic system makes it particularly attractive to study how sensory processing could be modulated by learning and memory. Moreover, olfactory structures have been early reported to exhibit oscillatory population activities easy to capture through local field potential recordings. An attractive hypothesis is that neuronal oscillations would serve to ‘bind’ distant structures to reach a unified and coherent perception. In relation to this hypothesis, we will assess the functional relevance of different types of oscillatory activity observed in the olfactory system of behaving animals. This review will focus primarily on two types of oscillatory activities: beta (15-40 Hz and gamma (60-100 Hz. While gamma oscillations are dominant in the olfactory system in the absence of odorant, both beta and gamma rhythms have been reported to be modulated depending on the nature of the olfactory task. Studies from the authors of the present review and other groups brought evidence for a link between these oscillations and behavioral changes induced by olfactory learning. However, differences in studies led to divergent interpretations concerning the respective role of these oscillations in olfactory processing. Based on a critical reexamination of those data, we propose hypotheses on the functional involvement of beta and gamma oscillations for odor perception and memory.

Several problems in accelerator shielding study

International Nuclear Information System (INIS)

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

1980-01-01

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

A triple-crystal phoswich detector with digital pulse shape discrimination for alpha/beta/gamma spectroscopy

International Nuclear Information System (INIS)

White, Travis L.; Miller, William H.

1999-01-01

Researchers at the University of Missouri - Columbia have developed a three-crystal phoswich detector coupled to a digital pulse shape discrimination system for use in alpha/beta/gamma spectroscopy. Phoswich detectors use a sandwich of scintillators viewed by a single photomultiplier tube to simultaneously detect multiple types of radiation. Separation of radiation types is based upon pulse shape difference among the phosphors, which has historically been performed with analog circuitry. The system uses a GaGe CompuScope 1012, 12 bit, 10 MHz computer-based oscilloscope that digitally captures the pulses from a phoswich detector and subsequently performs pulse shape discrimination with cross-correlation analysis. The detector, based partially on previous phoswich designs by Usuda et al., uses a 10 mg/cm 2 thick layer of ZnS(Ag) for alpha detection, followed by a 0.254 cm CaF 2 (Eu) crystal for beta detection, all backed by a 2.54 cm NaI(Tl) crystal for gamma detection. Individual energy spectra and count rate information for all three radiation types are displayed and updated periodically. The system shows excellent charged particle discrimination with an accuracy of greater than 99%. Future development will include a large area beta probe with gamma-ray discrimination, systems for low-energy photon detection (e.g. Bremsstrahlung or keV-range photon emissions), and other health physics instrumentation

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

DEFF Research Database (Denmark)

Jensen, Per Hedemann

1985-01-01

A computer model DEPSHIELD for the calculation of shielding factors for gamma radiation at indoor residences in multistorey and single-family houses has been developed. The model is based on the exponential point kernel that links the radiation flux density at a given detector point to a point...... it possible to determine the dose reduction effect from a decontamination of the different surfaces. The model has been used in a study of the consequences of land contamination of Danish territory after hypothetical core-melt accidents at the Barseback nuclear power plant in Sweden. The model has also been...

Contribution to the study of {beta} disintegration and of nuclear structure using experiments on certain {beta}-{gamma} cascades: 198{sub Au}, 86{sub Rb}, 170{sub Tm}; Contribution a l'etude de la desintegration beta et a l'etude de la structure nucleaire a l'aide d'experiences sur certaines cascades beta-gamma: 198{sub Au}, 86{sub Rb}, 170{sub Tm}

Energy Technology Data Exchange (ETDEWEB)

Lachkar, J. [Commissariat a l' Energie Atomique, Bruyeres-le-Chatel (France). Centre d' Etudes; Paris-11 Univ., fabulte des Sciences 91 - Orsay (France)

1969-07-01

{beta}{gamma} directional angular correlations and shapes of inner beta spectra leading to the first excited level of the final nucleus enable one to determine the nuclear matrix elements typical of the {beta} transition. In the three observed first forbidden cases: {sup 170}Tm, {sup 86}Rb, {sup 198}Au, these matrix elements do not confirm the independent shell model theory. Other hypotheses are then suggested and discussed. (author) [French] Les experiences de correlation angulaire {beta}{gamma} et l'etude du spectre {beta} conduisant au premier niveau excite du noyau final permettent de determiner les elements de matrices nucleaires caracteristiques de cette transition. Dans les trois cas etudies (transitions une fois interdites): {sup 170}Tm, {sup 86}Rb, {sup 198}Au, ces elements de matrices ne peuvent etre retrouves a l'aide du modele en couches et a particules independantes. D'autres hypotheses sont alors emises et discutees. (auteur)

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

Energy Technology Data Exchange (ETDEWEB)

Hattori, Seiichi; Ueki, Kohtaro.

1988-08-01

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

An Optimized Design of Single-Channel Beta-Gamma Coincidence Phoswich Detector by Geant4 Monte Carlo Simulations

Directory of Open Access Journals (Sweden)

Weihua Zhang

2011-01-01

Full Text Available An optimized single-channel phoswich well detector design has been proposed and assessed in order to improve beta-gamma coincidence measurement sensitivity of xenon radioisotopes. This newly designed phoswich well detector consists of a plastic beta counting cell (BC404 embedded in a CsI(Tl crystal coupled to a photomultiplier tube. The BC404 is configured in a cylindrical pipe shape to minimise light collection deterioration. The CsI(Tl crystal consists of a rectangular part and a semicylindrical scintillation part as a light reflector to increase light gathering. Compared with a PhosWatch detector, the final optimized detector geometry showed 15% improvement in the energy resolution of a 131mXe 129.4 keV conversion electron peak. The predicted beta-gamma coincidence efficiencies of xenon radioisotopes have also been improved accordingly.

Beta and gamma decay heat measurements between 0.1s--50,000s for neutron fission of 235U, 238U and 239Pu

International Nuclear Information System (INIS)

Schier, W.A.; Couchell, G.P.

1993-01-01

A helium-jet/tape-transport system is employed in the study of beta-particle and gamma-ray energy spectra of aggregate fission products as a function of time after fission. During the initial nine months of this project we have investigated the following areas: Design, assembly and characterization of a beta-particle spectrometer; Measurement of 235 U(n th ff) beta spectra for delay times 0.2 s to 12,000 s; Assembly and characterization of a 5 x 5 Nal(Tl) gamma-ray spectrometer; Measurement of 235 U(n th ff) gamma-ray spectra for delay times 0.2s to 1 5,500s; Assembly and characterization of HPGe gamma-ray spectrometer with a Nal(Tl) Compton-and-background-suppression annulus; Measurement of 235 U(n th ,ff) high-resolution gamma-ray spectra for delay times 0.6 s to over 100,000 s; Comparison of individual gamma-line intensities with ENDF/B-VI; Adaptation to our computer of unfolding program FERDO for beta and gamma aggregate fission-product energy spectra and development of a spectrum-stripping program for analysis of HPGe gamma-ray spectra; Study of the helium-jet fission-fragment elemental transfer efficiency. This work has resulted in the publication of twelve BAPS abstracts of presentations at scientific meetings. There are currently four Ph.D. and two M.S. candidates working on dissertations associated with the project

JSD1000: multi-group cross section sets for shielding materials

International Nuclear Information System (INIS)

Yamano, Naoki

1984-03-01

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

Analysis of a shield design for a DT neutron generator test facility.

Science.gov (United States)

Chichester, D L; Pierce, G D

2007-10-01

Independent numerical simulations have been performed using the MCNP5 and SCALE5 radiation transport codes to evaluate the effectiveness of a concrete facility designed to shield personnel from neutron radiation emitted from DT neutron generators. The analysis considered radiation source terms of 14.1 MeV monoenergetic neutrons located at three discrete locations within the two test vaults in the facility, calculating neutron and photon dose rates at 44 locations around the facility using both codes. In addition, dose rate contours were established throughout the facility using the MCNP5 mesh tally feature. Neutron dose rates calculated outside of the facility are predicted to be below 0.01 mrem/h at all locations when all neutron generator source terms are operating within the facility. Similarly, the neutron dose rate in one empty test vault when the adjacent test vault is being utilized is also less then 0.01 mrem/h. For most calculation locations outside the facility the photon dose rates were less then the neutron dose rates by a factor of 10 or more.

Double-layer neutron shield design as neutron shielding application

Science.gov (United States)

Sariyer, Demet; Küçer, Rahmi

2018-02-01

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

Measurements and FLUKA simulations of bismuth and aluminium activation at the CERN Shielding Benchmark Facility (CSBF)