Gamma ray and neutron shielding properties of some alloy materials

Singh, Vishwanath Pratap; Badiger, N.M.

doi:10.1016/j.anucene.2013.10.003

Cited by 180 publications

(49 citation statements)

References 17 publications

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“…Authors [9] developed a fitting for mula, called Geometric Progression (G P) which gave buildup factors of the good agreement with the ANS, 1991. Various researchers investigated gamma ray buildup factors of different materials such as fly ash brick materials [10], alloys [11], optically stimulated luminescence materials [12], bismuth borosilicate glasses [13] and oxide dispersion strengthened steels [14]. These studies show that this G P fitting method is suitable for computation of the buildup factors.…”

Section: Exposure Buildup Factormentioning

confidence: 99%

Shielding efficiency of lead borate and nickel borate glasses for gamma rays and neutrons

2015

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Gamma ray exposure buildup factor (EBF) and neutron shielding efficiency of borate glass sys tems in composition (100 -x -y) Na 2 B 4 O 7 -xPbO-yNiO (where x and y = 0, 2, 4, 6, 8 and 10 weight per centage) were investigated in the present study. Gamma ray EBF values were computed using Geometric Pro gression (G P) method for photon energy 0.015 to 15 MeV up to penetration depths of 40 mfp (mean free path). The macroscopic effective removal cross sections for fast neutron (2 to 12 MeV) were calcu lated. At low and high energy region, the EBF values were found minimum whereas maximum in the inter mediate energy regions with a sharp peak below 100 keV for oxide containing glass samples. The investigation was carried out for potential applications of borate glasses for radiation shielding.

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Section: Exposure Buildup Factormentioning

confidence: 99%

Shielding efficiency of lead borate and nickel borate glasses for gamma rays and neutrons

2015

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“…Studies on effective atomic numbers and electron densities have been reported by several investigators for chemical compounds [1,2], low-Z materials [3,4], alloys and steels [5][6][7][8], glass and minerals [9][10][11][12], biological materials [13], detectors [14,15], tissue substitutes [16][17][18], and composites [19].…”

Section: Introductionmentioning

confidence: 98%

Photon interaction with semiconductor and scintillation detectors

Singh

Badiger

2016

NUCL SCI TECH

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Mass attenuation coefficients, effective atomic numbers, and electron densities for semiconductor and scintillation detectors have been calculated in the photon energy range 1 keV-100 GeV. These interaction parameters have been found to vary with detector composition and the photon energy. The variation in the parameters with energy is shown graphically for all the partial photon interaction processes. The effective atomic numbers of the detector were compared with the ZXCOM program, and the results were found to be comparable. Efficiencies of semiconductor and scintillation detectors are presented in terms of effective atomic numbers. The study should be useful for comparing the detector performance in terms of gamma spectroscopy, radiation sensitivity, radiation measurement, and radiation damage. The results of the present investigation should stimulate research work for gamma spectroscopy and radiation measuring materials.

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“…This research was chosed as a reference to choose the shielding material and to designt the geometry of the treatment room. (Pouryavi, 2015) A research about the properties of some alloy materials for neutron and photon shielding was investigated as in (Singh & Badiger, 2014) . The alloy materials are CS-516, SS-403, SS-410, SS-316, SS-316L, SS-304L, Incoloy-600, Monel-400 and Cupero-Nickel.…”

Section: Methodsmentioning

confidence: 99%

Shield Modelling of Boron Neutron Capture Therapy Facility with Kartini Reactor’s Thermal Column as Neutron Source using Monte Carlo N Particle Extended Simulator

Dwiputra¹,

Harto²,

Sardjono

et al. 2016

ijpna

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Abstract-Studies were carried out to design a shielding for BNCT facility in the end of Kartini reactor's thermal column with predesigned collimator. The design consist of selecting the material and their thickness. The shielding is required to absorb the leaking radiation until the Dose Limit Value of 20 mSv/year for radiation worker is met. The material considered were paraffin, barite concrete, borated polyethylene, stainless steel 304 and lead. The calculation was done using MCNPX tally facility with converted dose limit value of 10.42 µSv/hour. Design number two were chosen as the best from three designs which surrounded a room with length, width and height of, respectively 200 cm, 200 cm and 166.4 cm. The first and main layer are borated polyethyelene and barite concrete of 20 and 30 cm, respectively. The additional layer are borated polyethyelene and barite concrete of 15 cm and 15 cm with less volume than the main layer to decrease the primary straight radiation from the thermal column. Maximum radiation dose rate is 7.0746 µ Sv/ hour in cell 227 with average dose rate of 2.58712 µSv/hour.

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Gamma ray and neutron shielding properties of some alloy materials

Cited by 180 publications

References 17 publications

Shielding efficiency of lead borate and nickel borate glasses for gamma rays and neutrons

Shielding efficiency of lead borate and nickel borate glasses for gamma rays and neutrons

Photon interaction with semiconductor and scintillation detectors

Shield Modelling of Boron Neutron Capture Therapy Facility with Kartini Reactor’s Thermal Column as Neutron Source using Monte Carlo N Particle Extended Simulator

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