Point Kernel Technique Research Articles

3DGShield: A new 3D gamma ray shielding code for arbitrary source and shield geometry based on point kernel technique

3DGShield, A new 3D gamma ray shielding code, based on the method of point kernel technique, has been developed for arbitrary regular & irregular shaped source and shield geometries. The 3D geometry modeling and its mesh generation are performed using Gmsh, a free open source finite element mesh generation software. The output of the Gmsh software is in Gmsh MSH (.msh) file format which will be used as an input of geometry description of our code. The salient features of 3DGShield are (i) ability to handle any type of complex source geometries including regular source shapes like point, line, disc, box, cylinder, sphere, hexagonal cylinder, cone etc. and irregular source shapes like pieces of broken glass, piece of brick etc., (ii) handling discrete as well as distributed photon energy sources, (iii) modeling any kind of regular and irregular complex multilayered shield geometries, (iv) having provision to export the output results into ParaView, an open source data analysis and visualization software, for quick visualization and analyzing of the results. Buildup factor is the correction factor which takes account the contribution of scattered photon. Various appropriate multilayer build-up factor formulas have been incorporated in our code. Out of these, the most popular methods are Blizard's method, Goldstein method of an effective atomic number and Broders's method. Verification and validation of the code have been done by a number of benchmark problems available in literature. The results are presented in this paper which demonstrate the accuracy and efficiency of the code in complex geometries. This code is written in Python-2.7 language and compatible both in Windows and Linux operating system.

U-Shielder − an open source software for the estimation of gamma shielding using depleted uranium

U-Shielder is an open source software for the estimation of gamma-ray shielding using depleted uranium. It can be used to estimate thickness of shield for gamma-ray energies, ranging from 500 keV to 10 MeV. The computational methodology is based on the point kernel technique. It uses exponential attenuation model with Taylor's form of build-up factor. Numerical method has been used to calculate shielding thickness. The software was developed in TURBO-C++ and tested in both DOS and Windows operating systems. The software is available at Nuclear Energy Agency (NEA) website. The code proved useful in the designing of radioactive material transport container.

Calculation of dose rate conversion factors for 238U, 232Th and 40K in concrete structures of various dimensions, with application to Nis, Serbia

The absorbed gamma dose rate in indoor air due to natural radionuclides in concrete as a building material was determined in this work. The dose rate conversion factors for (238)U, (232)Th and (40)K, for standard rooms as well as rooms with different sets of dimensions, were evaluated by the point kernel technique, using Harima (geometric progression) build-up factors. The values of the conversion factors, in units (nGy h(-1) (Bq kg(-1))(-1)) calculated for the standard room are: 0.76, 0.91 and 0.070, respectively for (238)U, (232)Th and (40)K. The fitting formula was obtained for dose rate conversion factors, enabling them to be conveniently calculated for a room with arbitrary dimensions. For concrete block samples collected in the area of Niš, Serbia, the measurement of the radionuclide activity concentrations was also carried out. The evaluated absorbed dose rate conversion factors were then applied in the assessment of corresponding indoor gamma dose rates, finding that all the concrete samples fulfilled the usage requirement.

IGSHIELD: A new interactive point kernel gamma ray shielding code

IGSHIELD is an interactive gamma ray shielding code (stand alone) developed in Visual Basic Version 6.0 for windows operating system. The computational methodology is based on the point kernel technique employed in QADCGPIC (and hence QADCGGP) code. In IGSHIELD, all the features of QADCGPIC code written in FORTRAN and its user interface GUI2QAD-3D written in visual basic have been incorporated and several additional improvements have been made. The new improvements made in the code are (i) handling of multiple source shapes with accompanying energy distributions, (ii) incorporation of hexagonal cylinder as a shield body in the combinatorial geometry routine, (iii) provision to declare hexagonal cylinder and truncated right circular cone as source bodies, (iv) provision to declare point, line and plane as source bodies and (v) arbitrary orientation of sources such as cylinder, hexagonal cylinder and truncated cone. The newly introduced source bodies will be of immense use for calculations involving hexagonal shaped cylindrical sub-assemblies of fast reactors and cone shaped radioactive plume release from stack. The source array option will be of use for calculations involving stacked radioactive drums in vast storage areas. The code is validated by comparing the results with (i) simple source-shield configurations that can be solved using analytical equations and MCNP code, (ii) a point source in an infinite shield medium configuration that is solved using simple analytical equation and (iii) a standard problem in which an irradiated fuel is surrounded by various shielding materials and the problem is solved using MCNP. Several sample problems encountered in nuclear industry and other useful information required for shield designers and analysts are the other highlights of the code. The input for any kind of situations can be prepared easily using the forms that flash with appropriate prompts. Since entry of data is the important part of the input, its validity is checked immediately and the user is cautioned against incorrect data. The input module accepts only the correct data. If the data happens to be an object, the code instantaneously display the 3D view of all the objects built so far. Thus, the Interactive Gamma Ray Shielding Code will cater the needs of a shield designer as well as serve as an excellent educational tool.

Comparison of Line Beam Response Function for Gamma-ray Skyshine Analysis Based on Single Scattering Method with the Monte Carlo Calculations

A simple calculation method for estimating gamma-ray skyshine dose rates has been developed on the basis of the line beam response function (LBRF). The new data of LBRFs were generated by the single scattering with point kernel technique (single-scattering method). These resulting LBRFs were compared with the EGS4 and MCNP Monte Carlo calculations. The values of the new LBRFs for the large emitted angle become smaller than the LBRFs obtained with the Monte Carlo calculations with an increase of the distance from the source. This discrepancy increases with an increase of the photon energy.

Comparison of Line Beam Response Function for Gamma-ray Skyshine Analysis Based on Single Scattering Method with the Monte Carlo Calculations.

A simple calculation method for estimating gamma-ray skyshine dose rates has been developed on the basis of the line beam response function (LBRF). The new data of LBRFs were generated by the single scattering with point kernel technique (single-scattering method). These resulting LBRFs were compared with the EGS4 and MCNP Monte Carlo calculations. The values of the new LBRFs for the large emitted angle become smaller than the LBRFs obtained with the Monte Carlo calculations with an increase of the distance from the source. This discrepancy increases with an increase of the photon energy.

Use of computer code for dose distribution studies in A 60CO industrial irradiator

This paper presents a benchmark comparison between calculated and experimental absorbed dose values tor a typical product, in a 60Co industrial irradiator, located at ININ, México. The irradiator is a two levels, two layers system with overlapping product configuration with activity around 300kCi.Experimental values were obtanied from routine dosimetry, using red acrylic pellets. Typical product was Petri dishes packages, apparent density 0.13 g/cm3; that product was chosen because uniform size, large quantity and low density. Minimum dose was fixed in 15 kGy. Calculated values were obtained from QAD-CGGP code. This code uses a point kernel technique, build-up factors fitting was done by geometrical progression and combinatorial geometry is used for system description. Main modifications for the code were related with source sumilation, using punctual sources instead of pencils and an energy and anisotropic emission spectrums were included.Results were, for maximum dose, calculated value (18.2 kGy) was 8% higher than experimental average value (16.8 kGy); for minimum dose, calculated value (13.8 kGy) was 3% higher than experimental average value (14.3 kGy).

A dosimetric model for intrathecal treatment with 131I AND 67GA

Purpose : Calculations were performed of absorbed dose distributions of the β-emitter 131I and the Auger emitter 67Ga for intrathecal administration. Methods and Materials : The proposed dosimetric model accounts for the macroscopic distribution of the activity, by means of a Medical Internal Radiation Dose Committee approach, and for the microscopic distribution of activity, by means of a point kernel technique. This point kernel approach was used in combination with a distance histogram technique, to study in more detail the absorbed dose distribution in the cerebro-spinal fluid, in the surface of the central nervous system, and in tumor sites. We simulated decreased uptake, as well as highly selective uptake in free-floating tumor cells and in meningeal lesions (1–16 cells thick). Results : In case of limited access to lesions adherent to the pia mater, the β-emitter 131I provides crossfire from the CSF, resulting in a higher absorbed dose (Gy/MBq) in these lesions as compared with the Auger emitter 67Ga. In case of increasing radionuclide uptake, the increment of the absorbed dose in the adherent lesions and the free floating cells from 67Ga is considerable because of the local deposition of energy by this radionuclide. Conclusions : The model might be useful to select the optimal emission characteristics of radionuclides applicable for intrathecal therapy, which is demonstrated in a comparison of the Auger emitter 67Ga and the β-emitter 131I.

Effect of head phantom size on 10B and 1H[n,gamma]2H dose distributions for a broad field accelerator epithermal neutron source for BNCT.

The effect of head phantom size on the 10B and 1H[n,gamma]2H dose distributions for a broad epithermal neutron radiation field generated by an accelerator-based epithermal neutron source for boron neutron capture therapy (BNCT) have been studied. Also two techniques for calculating the absorbed gamma dose from a measured gamma-ray source distribution are compared: a Monte Carlo technique, which is well accepted in the BNCT community, and a Point Kernel technique. The count-rate distribution in the central plane of three rectangular parallelopiped head water phantoms irradiated with an epithermal neutron field was measured with a boron trifluoride (BF3) detector. This epithermal neutron field was produced at the Ohio State University Van de Graaff Accelerator Facility. The 10B absorbed dose and the gamma-ray source have the same distribution in the head phantom as the BF3 count-rate distribution. The absorbed gamma dose from the measured source distribution was calculated using MCNP, a Monte Carlo code, and QAD-CGGP, a Point Kernel code. The most pronounced effect of phantom size on 10B absorbed dose was on the dose rate at the depth of maximum dose, dmax. An increase in dose rate at dmax was observed with a decrease in phantom size, the dose rate in the smallest phantom being larger by a factor of 1.4 than the dose rate in the largest phantom. Also, dmax for the phantoms shifted deeper with a decrease in phantom dimensions. The shift between the largest and the smallest phantoms was 6 mm. Finally, the smaller phantoms had lower entrance 10B dose as a percent of the dose at dmax, or better skin sparing. Our calculations for the gamma dose show that a Point Kernel technique can be used to calculate the dose distribution as accurately as a Monte Carlo technique, in much shorter computation times.

Optimization of radiation defectoscopy parameters by Monte Carlo and Point Kernel technique

Optimization of radiation defectoscopy parameters by Monte Carlo and Point Kernel technique

Photon shielding for 241Am surface and point sources.

To determine radiation protection requirements for work with actinide elements, a method for rapidly estimating effective dose-equivalent rates from low-energy photons has been developed. This paper describes results obtained from a personal computer program that incorporates the point kernel technique to predict radiation fields from shielded and unshielded sources containing 241Am. Information generated has been used to determine procedures and to design facilities for handling actinides at Argonne National Laboratory's site at the Idaho National Engineering Laboratory. Area or point sources can be treated; effects of Compton scattering in air and in solid shields are considered. Users can select an appropriate response function; their choice has a strong influence on predicted dose rates from unshielded sources.