mcnp4c code

In this research, the use of tungsten (W) and barium (Ba) in composition with lead (Pb) was investigated to decrease weight and to increase the efficiency of the protective aprons in attenuating of the diagnostic X-rays. First, the X-ray tube of the Radiation Research Center (RCC) of the Shiraz University was simulated by MCNP4C code to obtain the bremsstrahlung and characteristic X-ray spectrum. In order to confirm the simulation, the obtained output of this code was compared with the resultant output of the spectrum simulator software, viz., the spectrum processor (SP). The simulated protective apron samples with the dimension of 10cm×10cm and weight of 86.5, 81.5 and 70 g, containing various percentages of Pb, W and Ba, were placed in the path of the simulated spectra, and the attenuation coefficient of radiation was calculated using MCNP4C. The results showed that the composition of Pb, Ba and vinyl, with the percentages of 23, 42 and 35 and also 17, 48 and 35, respectively, improved the attenuation about %46 in comparison with the use of pure Pb.

In this research, a water calorimeter that can be used for dosimetry of electron beams from therapy accelerators at low absorption dose has been simulated using MCNP4C computer code. Then the depth dose curves were calculated and compared with experimental values. In addition, the effect of back scattering layers in depth of maximum dose was studied and it was found that these layers absorb the maximum dose in lower depths.

SVHI-Co-60-T irradiation system has been used as a portable one with Co-60 source for onion irradiation. To optimize and for the change of the system application, sufficient knowledge of the absorbed dose of the products is necessary. In this paper, the absorbed dose in the irradiation chamber is calculated for different points using MCNP4C code and then the results are compared with the measurements by the Red, Amber and Clear dosimeters of the Radiation Application Research School, along the central line of the chamber. The values ​​obtained from dosimeters are in good agreement with the code results. The dose distribution in the irradiation chamber has been calculated on a plane with dimensions of 60cm×30cm. The minimum and maximum dose rates and the dose rate at the center of the same dose area of the isodose curve are 0.3kGy/hr, 2.6kGy/hr and 1.25kGy/hr, respectively. Also, the dose rate measured in the central area of the chamber using Red, Amber and Clear dosimeters is 1.37kGy/hr, 1.303kGy/hr and 1.302kGy/hr, respectively. Differences between the calculated dose using MCNP4C code and the measured dose emplaying three types of dosimeters are 8.7%, 3.8% and 3.8%, respectively. After loading, the spacers have been designed in front of new sources with MCNP4C code to have suitable dose with a uniform ratio to change its application from onion to wheat irradiation and also to compare with the measurements.

This paper is dealing with the reactor parameter determination experiments performed at the light water subcritical reactor (LWSCR) in Esfahan. The parameter determination methods are derived from a theory based on neutron source position. The LWSCR reactor system is modeled by Monte Carlo calculations (MCNP4C), WIMSD5B and CITATION, and the criticality safety and neutron flux distribution problems are presented and analyzed. The analysis of the multiplication factors obtained by MCNP4C are in fairly good agreement with the results obtained by WIMSD5B and CITATION. For the flux distribution calculations, the MCNP4C results differ from the experimental observations by less than 12% and by ~5% for the WIMSD5B and CITATION.

The silver activation counters are commonly used for pulsed-neutron yield measurements especially in plasma focus devices. The counter normally consists of a Geiger-Muller tube along with silver foils and polyethylene (as a moderator), which is calibrated against an Am-Be radioisotope neutron source. The neutrons, after being slowed-down in the polyethylene, activate the silver foils. By measuring the foil activity with a Geiger-Muller counter, the neutron yield is determined. In the present paper, the activation counter’s calibration constant calculation using the MCNP4C code is explained. The calculated calibration constant is in good agreement with the experimental results.