Results 1 - 10 of 401
Results 1 - 10 of 401. Search took: 0.029 seconds
|Sort by: date | relevance|
[en] Plane sources with slab shield are frequently encountered in gamma-ray attenuation problems. In the realm of the Point Kernel Integration Method, plane sources with slab shield are often used, as an approximation for more general source-shield geometry in the preliminary reactor shield design and related radiation safety analysis. Recently, the point source buildup factor data approximation in the form of an expanded polynomial set was successfully introduced. In this paper, it is demonstrated that the above function fits into the integrand of the point kernel, for the circular disk source with general cosine emission, in such a fashion that the modified integrand is still as easily integrated as before. The results is expressible in terms of incomplete gamma-function. Furthermore, this result can be used as the basis for the solution of various infinitive slab source-shield geometry problems that lead to the similar integrable kernels
[en] Many methods, especially Monte Carlo simulation technique and Point Kernel method are ideally used for radiation profile studies. However, these methods are either time consuming or fairly accurate when dealing with extended gamma sources particularly for optimization studies. Furthermore, while the buildup factor and the attenuation effects were well investigated in the literature, little work was done about the systematic influence of source extension. In this work we focus of the source geometry using the generalized Laplace's expansion. we express the bare gamma photon flux rate in terms of the standard Cartesian multipole moments. using the properties of these moments we establish a close relationship between the radiation profile and the geometrical features of the source. As applications we propose to use the multipole expansion method to investigate the radiation profile isotropy of the source. A detailed study of the arrangement of the unit pencil sources of the tunisian irradiation facility is performed. Using this method, millions of possible configurations for various load plans were investigated, in few minutes and even multi steps scenarios were considered. As a result, the current configuration of the source was found to be not optimized. Furthermore, using these analytical method it was possible to optimize the activity of each new unit source
[en] In this paper, two models to establish the minimum exposure path in a randomly placed radioactive sources enclosure are developed. The first model establishes the minimum exposure rate path and the second establishes the minimum total exposure on a path normal to the inlet and outlet surfaces. Although the point kernel technique is utilized and the enclosure is assumed rectangular with randomly placed radioactive point sources, the two models are technique independent and can be easily incorporated with deterministic and statistical methods as well as other types of enclosures and source geometries. (Author)
[en] Approximations simply to be evaluated with B = [0.05 <= THETA <= 0.7; 0.01 <= b <= 25] and a maximum relative error not exceeding 3.3 x 10-5 are given for the function F(THETA, b) = /sub 0/∫/sup THETA/ e/sup -b sec β/ dβ often needed in shielding calculations. The methods used for determining the approximations are principally also applicable to approximating other integral functions and are therefore considered in more detail. (author)
[en] The point-kernel method is a widely used practical tool in shielding design. In the point-kernel approach the desired quantity (flux, fluence, or dose equivalent) is obtained by multiplying the part of the quantity resulting from uncollided flux with the parameter called buildup factor which accounts for the scattered radiation inside the shielding medium. It is therefore clear that the accuracy of the final result is highly dependent on the accuracy of the buildup factors used in the calculations. Although buildup factors have been widely used for F067 - ray shielding calculations, their application for neutron transport simulations is very limited. Recently a novel approach, based on support vector regression (SVR) technique, for multi-layer buildup factor calculation has been proposed and tested for F067 - ray shielding calculations, while for neutron shielding calculations some initial analyses have been conducted. Initial tests of the SVR model on mono-layer shields revealed that the relative average deviation (RAD) from the reference data (obtained by Monte Carlo calculation) in the estimation of F067-ray buildup factors is 5.8 percent with maximum deviation of 34 percent observed for water shield 5.21 mfp thick at incident F067-ray energy of 0.5 MeV. RAD for double-layer shields was 9.7 percent with maximum deviation of 68.77 percent observed for a shield comprised of 1.17 mfp of water and 5.75 mfp (mean free path) of iron, at 7.25 MeV F067-ray energy. RADs for triple-layer and quadruple-layer shields were 24.3 percent and 29.7 percent, respectively. The initial tests illustrate clear potential of the novel SVM approach to the F067-ray buildup factor determination. The physical complexity of the neutron radiation transport and the experience gained while developing the F067-ray model indicate that the development of the SVR neutron buildup factor model is possible but would also require intensive research of the learning technique itself. (author).
[en] In this work, the progress in the development of a new point kernel code is illustrated. The code is focused on treating the stratified shields using the iterative scheme. The results of the simulation of the code on simple stratified shields cases are discussed here. The shields consist of thin-layers combinations of lead-water and iron-water. This work is restricted for the treatment of 1 MeV photons only. The results show good consistency and stability in the performance of the developed code, at least within the limits of this study. Even though the code implements very simple methodology for approaching the stratified shields, the shown results are satisfactory. The deviations off the MCNP5 results are mainly due to the buildup factor behavior through the iterative scheme process. This suggests an additional focus on the iterative scheme itself is needed. Many factors in the current scheme can be reconsidered, mainly; the use of Kalos' formula and the imposed characteristics of the equivalent layers. Other sources of error can be things such as the type of tally/detector used in the MCNP5 simulation and the type of the basic single-layer buildup factor used to produce the implemented fitting functions. The code is yet to be tested with other cases of material compositions and photon energies. The algorithm and basic theory also needs to be upgraded in order to produce less deviation
[en] In this work a Monte Carlo transport code named TART is engaged for total neutron buildup factors determination for NBS04 concrete type. The first step to model the total neutron buildup factors is to conduct a validation process of the TART code by calculating transmission factors for both neutrons and secondary photons. The transmission factors are then used for total ambient ambient dose equivalent estimation. The direct component is calculated manually by using the exponential law, with the cross sections obtained from MCNP6. Finally, total ambient dose equivalent neutron buildup factor is calculated as the ratio of the total ambient dose equivalent and the direct component. The results are compared with the MCNP6 results. Obtained buildup factors are then embedded in a point kernel code named QAD-GCCP. A test case is modeled in TART, MCNP6 and QAD-CGGP and the results of the three codes are compared.(author).
[en] In an effort to reduce the dose to operating technicians performing fixed-time procedures on encapsulated source material, a program has been developed to optimize the layout of workstations within a facility by use of a genetic algorithm. Taking into account the sources present at each station and the time required to complete each procedure, the program utilizes a point kernel dose calculation tool for dose estimates. The genetic algorithm driver employs the dose calculation code as a cost function to determine the optimal spatial arrangement of workstations to minimize the total worker dose.
[en] The Clinac 2100C of Varian has the option of a dynamic wedge. The control of the dynamic wedge is done by means of a segmentation table, which specifies a number of asymmetric fields with corresponding numbers of monitor units. Advantages of this dynamic wedge are the possibility to have more wedge angles and an increased flexibility of operation; a disadvantage is that the measurement of the data needed for a planning system may be time consuming. For the calculation of the dose distribution, we have developed an algorithm which eradicates this disadvantage because it only uses measured data from square fields of the open beam: central axis depth-doses, profiles and Peak Scatter Factors. It is based on a previously developed method which separates the dose into three factors: depth-dose, boundary distribution and envelope profile, and computes the dose of irregular field as well. Depth-dose and boundary distribution are computed by convolution of a field intensity with 'scatter' and 'boundary' pencil beam kernels respectively. For a rectangular field, the field intensity is computed by a weighted sum of intensity matrices, derived from the segmentation table. The matrix elements are equal to 1 within the asymmetric fields and equal to an effective transmission factor under the jaws. The weighting factors in the sum take into account the number of monitor units and the collimator scatter back to the monitor. For an irregular field, the blocked parts of the field are modified, taking into account the transmission through the blocks. This algorithm has been implemented in CADPLAN (Varian-Dosetek) and has been compared with measurements. The results are in good agreement with the international requirements on dose accuracy
[en] Trapezoidal rule and Gauss-Legendre quadratures are representative of the numeric techniques used in integrating over radiation source regions in point-kernel shielding programs. The orders of quadrature selected for such integrations are important since a sparse quadrature may calculate inaccurate results while unnecessarily large orders of quadrature waste computer time. Rules are given for choosing trapezoidal and Gauss quadrature orders for linear, radial, and azimuthal intervals of integrate, based on problem geometry and source attenuation. These rules show that for like accuracy, a trapezoidal rule quadrature of order N may be replaced by a Gauss quadrature with order between the square root of N and N/2. Replacing trapezoidal-scale quadratures by lesser order Gauss quadratures can save large amounts of computer time. Gauss quadratures, on the other hand, ideally should be set up individually for detector points in different locations