Photon Point Source Buildup Factors for Air, Water, and Iron

Chilton, A. B.; Eisenhauer, C.M.; Simmons, G.L.

doi:10.13182/nse80-a18714

Cited by 80 publications

(27 citation statements)

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“…The build-up is defined as the ratio of the total value of a specified radiation quantity at any point to the contribution to that value from radiation reaching the point without having undergone a collision. Build-up factors have been computed by various codes such as PALLAS (Takeuchi and Tanaka, 1984), ADJMON-I (Simmons, 1973;Chilton et al, 1980), ASFIT (Gopinath and Samthanam, 1971) and EGS4 (Nelson et al, 1985). These codes use an accurate algorithm for the Klein-Nishina cross-section which eliminates other sources of error.…”

Section: Introductionmentioning

confidence: 99%

Gamma-ray exposure build-up factors of some brick materials in the state of Punjab

Singh

Badiger

2013

Radioprotection

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Section: Introductionmentioning

confidence: 99%

Gamma-ray exposure build-up factors of some brick materials in the state of Punjab

Singh

Badiger

2013

Radioprotection

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“…For air, water, concrete, and iron, the buildup factors were calculated using the methods of moments by Eisenhauer and Simmons 10) and by Chilton et al 11) Concrete had the composition of ''NBS'' concrete. 10) Both Eisenhauer and Simmons 10) and Chilton et al 11) used interaction coefficients, attenuation coefficients, and energy deposition coefficients consistent with those used by Hubbell. 12) Their calculations accounted for Compton scattering and annihilation but did not include fluorescence or bremsstrahlung photons. The buildup values for lead and uranium were determined using the discrete ordinates-integral transport method with the PALLAS code, which formally included secondary photons produced by annihilation, fluorescence, and bremsstrahlung.…”

Section: )mentioning

confidence: 99%

Energy Absorption Buildup Factors and Energy Conservation

Overcamp

2009

Journal of Nuclear Science and Technology

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The ANSI/ANS energy absorption buildup factors (ANSI/ ANS-6.4.3-1991) were used to determine if they conserve absorbed energy from a point source of monoenergetic photons in an infinite, absorbing medium. Air, water, concrete, iron, lead, and uranium were considered for photon energies ranging from 0.015 to 15 MeV. The primary calculations were based on fitting natural cubic splines to the energy absorption buildup factors and analytical integration of the energy absorbed using the cubic spline curve fit. This method was supplemented by linear spline interpolation. For air, water, concrete, and iron, the ANSI/ ANS energy absorption buildup factors were determined using the method of moments with the energy attenuation and deposition coefficients of Hubbell (NSRDS-NBS 29). The results of this study essentially confirmed energy conservation for these four materials using the energy attenuation and deposition coefficients of Hubbell. For lead and uranium, the ANSI/ANS energy absorption buildup factors were determined using the discrete ordinates-integral transport method with PHOTX energy attenuation and deposition coefficients that are essentially the same as those in the ANSI/ANS report. The calculations in this study could not confirm energy conservation for these two materials.

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“…The factors of accumulation of the absorbed dose and kerma in water are calculated by various methods in [8,15] with an accuracy of 3-5%. We used the data presented in [8].…”

Section: R 2 S(o) P(-g(e ) )Lte(e )Ebd(er) (3)mentioning

confidence: 99%

Three-dimensional fourier transformation calculation of the radiation dose received from a60Co source

Васильев

1999

Biomed Eng

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The main difficulty of the determination of the absorbed dose distribution in a patient's body consists of calculating the scattered radiation contribution. As shown in [4,9,12], the error of calculation of the scattered radiation dose by simple one-dimensional methods (calculation of effective radiological thickness, Bataut's equation, etc.) in some cases can be rather large (up to 20% and more in case of elaborate heterogenic structures).As shown in [5][6][7], the calculation of the dose absorbed by a three-dimensional object can be reduced using several simplifying assumptions to solving the equation of convolution in three-dimensionai space: D(r) = IIf d~(r')p(r')K(r-r')dv ,V where dp(r ~) is the nonscattered photon flux in the irradiated object; 9(r') is the electron density distribution in the object; K(r -r) is the convolution core determining the contribution of photons scattered in the elementary volume (voxel) dv with the coordinates r' into the dose absorbed in the elementary volume with the coordinates r. The convolution equation can be solved using an efficient method based on Fourier transformation. According to the convolution theorem [13], convolution of two functions is equal to the product of their Fourier transforms: f(x) is the product qb(r)p(r); the convolution core g(x), the scattering core K(r -r).If operating with discrete data, the calculation algorithm can be reduced to the three-dimensional fast Fourier transformation (FFT) of two three-dimensional arrays (the function to be convoluted and the scattering core), multiplication of arrays, and inverse FFT of the resulting array.The method of the absorbed dose calculation described above was proposed in the 1980s [5][6][7]. However, in spite of approximately 5000-fold acceleration of calculation provided by this method as compared with direct integration, it did not find wide application because its implementation requires rather powerful computers. According to our estimates [2], modern PCs allow implementation of this method. The goal of this work was to develop special software for implementing this method, construct a convenient physical model of the scattering core K(r -r'), and determine the program operation rate and the accuracy of the results obtained. A 6°Co source of gamma-radiation was used in the test. Dose Calculation ModelThe dose distribution was calculated for two types of irradiated objects: a cubical tissue-equivalent phantom of side 32 cm irradiated from a distance of 75 cm (irradiation field, from 4 x 4 cm to 20 x 20 cm) and a tissue-equivalent cylinder 12 cm in diameter (irradiation field, 20 x 20 cm; the distance from the radiation source to the cylinder axis, 75 cm). The results of calculation were compared to the data contained in standard dose field atlases and to the results of calculation by the Monte Carlo technique.

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Photon Point Source Buildup Factors for Air, Water, and Iron

Cited by 80 publications

References 0 publications

Gamma-ray exposure build-up factors of some brick materials in the state of Punjab

Gamma-ray exposure build-up factors of some brick materials in the state of Punjab

Energy Absorption Buildup Factors and Energy Conservation

Three-dimensional fourier transformation calculation of the radiation dose received from a60Co source

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