The production of fission products during reactor operation has a very important effect on  reactor reactivity .Results of neutron cross section evaluations are presented for the main product nuclides considered as being the most important  for reactor calculation and burn-up consideration . Data from the main international libraries considered as containing the most up-to-date nuclear data   and the latest experimental measurements are considered in the evaluation processes, we describe the evaluated cross sections of the fission product nuclides by making inter comparison of the data and point out the discrepancies among libraries.

  In this study , extraction of  irradiated high concentration hexavallent uranium from fission products ( high gamma radiation ) was carried out using multistages countercurrent continuous technique (mixer setller) , employing this technique requires a designing flow sheet that recover all the amounts of uranium and to minimize its losses in the nuclear waste streams. Due to the several parameters required to reach this design, SEPHIS program which is one of the famous code in this field were used to select the optimum conditions through many theoretical runs  , finally the experimental results give a good assurance in SEPHIS results and its optimum conditions. 

    In this research, we have achieved the description of radionuclides that exist in the samples of Diyala river sediments as well as to measure the specific activities using gamma-ray spectroscopy. The eleven samples were collected among the length of Diyala River starting from Al- Rustumiya and finishing at the point where Diyala River meets Tigris which is in Baghdad.      Gamma-ray spectrometry system consists of high-purity germanium detector (HpGe) with 50% efficiency and resolution (2.2 keV) for the energy (1332 keV) was used for standard source 60Co.     Card spectrum analyzer connected to the PC type Pentium 4 was used to view the spectrum. And  rates of the speci

The method of measurement dosimetry in neutron – gamma field by using CaSo4 : Dy (PTFE) disc which has a diameter of 1.3mm and thickness of 0.2mm and using hydrogenated material as a converters of neutron to recoil protons (n-p) reaction, the discs were irradiated by neutron source (241Am-Be) with flux of 4.5?105 n/cm2s for different time to obtain different dose. The TL signals, which we have been obtained by using the converters, are increases to 71%. So we can resolve the neutron and gamma in mixed field.

Multipole mixing ratios for gamma transition populated in from reaction have been studied by least square fitting method also transition strength ] for pure gamma transitions have been calculated taking into account the mean life time for these levels .

  The calculated neutron yields from (α, n) reactions are very important in analyzing radiation shielding of spent fuel storage, transport and safe handling. The cross sections of 63Cu (α, n) 66Ga and 65Cu (α, n) 68Ga reactions are calculated for different α-energies using different sets of programs using Matlab language. The values deduced energy is from threshold to Eα= 30 MeV and to Eα= 40 MeV for 63Cu (α, n) 66Ga and 65Cu (α, n) 68Ga respectively. The weight average cross section was then  used to calculate the neutron yields y0 (n/106α) for each reaction .The empirical formula was then suggested to calculate total neutron yield to each isotope.
 

The Local manufacturing scanning gamma system designed in Tuwaitha site for nondestructive assay method of radioactive waste drums, where it consist of two main parts with their belongings for controlling the of detector and drum movements up-down and rotation respectively. The volume of the used drum is 220 L with 85 cm height. The drum filled with Portland cement. Six cylindrical holes were made within cement drum and distributed in radial arrangement.The152Eu source inserted in these holes individually, to measure the average angular count rate of gamma radiation. The full energy efficiency value for geometry of drum and detector is computed for thirteen photo peaks. The average efficiency represented by the curve of these peaks indic

In this work, the total linear attenuation coefficients µ(cm
-1
) were calculated and studied
for particulate reinforced polymer-based composites. Unsaturated polyester (UP) resin was
used as a matrix filled with different concentrations of Al, Fe, and Pb metal powders as
reinforcements. The effect of the metal powders addition at different weight percentages in
the range of (10,20,30,40,50)wt % and gamma energy on attenuation coefficients was studied.
The results show, as the metallic particulates content increase, the attenuation coefficients will
increase too, while it, were exhibited a decrease in their values when the gamma energy
increase.The total linear attenuation coefficients of gamma ray fo

In this study, dependence of gamma-ray absorption coefficient on the size of Pb particle size ranging from 200µm up to 2.5mm, using different weights of each particle size. The results show that gamma-ray attenuation coefficient is inversely proportional with the size of Pb particle size due to the reduction of the spaces between the lead particles.

The present calculation covers the building shield during irradiation process and under water storage of three milion curries Cobalt-60 radiation source the calculation results in design requirement of 8m depth of water in the source stoeage pool