• Journal of Radiation Protection and Research
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• v.26 no.2
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• pp.67-71
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• 2001

In order to shield the neutrons affecting the background of Low Level Gamma Ray Spectrometer, a neutron shielder was designed. The method used in this study for neutron shielding was the deceleration of fast neutrons by high density polyethylene(HDPE) and the absorption of those slowing-down neutrons by $B_4C$. The calculation results of neutron Interaction in HDPE using Monte Carlo simulation code MCNP4B showed that the thermal-neutron flux was maximum at 10 cm thickness of HDPE. The results also showed that 95% of the thermal neutrons were absorbed by 2 mm thickness of $B_4C$ absorber Consisted of 30 w% $B_4C$ and 70 w% polymer. The results of the Monte Carlo calculation were in good agreement with the experimental value obtained by a neutron shielding apparatus designed for this purpose.

• Nuclear Engineering and Technology
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• v.48 no.3
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• pp.785-794
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• 2016

The Argonne National Laboratory of the United States and the Kharkov Institute of Physics and Technology of the Ukraine have been collaborating on the design, development and construction of a neutron source facility at Kharkov Institute of Physics and Technology utilizing an electron-accelerator-driven subcritical assembly. The electron beam power is 100 kW using 100-MeV electrons. The facility was designed to perform basic and applied nuclear research, produce medical isotopes, and train nuclear specialists. The biological shield of the accelerator building was designed to reduce the biological dose to less than 5.0e-03 mSv/h during operation. The main source of the biological dose for the accelerator building is the photons and neutrons generated from different interactions of leaked electrons from the electron gun and the accelerator sections with the surrounding components and materials. The Monte Carlo N-particle extended code (MCNPX) was used for the shielding calculations because of its capability to perform electron-, photon-, and neutron-coupled transport simulations. The photon dose was tallied using the MCNPX calculation, starting with the leaked electrons. However, it is difficult to accurately tally the neutron dose directly from the leaked electrons. The neutron yield per electron from the interactions with the surrounding components is very small, ~0.01 neutron for 100-MeV electron and even smaller for lower-energy electrons. This causes difficulties for the Monte Carlo analyses and consumes tremendous computation resources for tallying the neutron dose outside the shield boundary with an acceptable accuracy. To avoid these difficulties, the SOURCE and TALLYX user subroutines of MCNPX were utilized for this study. The generated neutrons were banked, together with all related parameters, for a subsequent MCNPX calculation to obtain the neutron dose. The weight windows variance reduction technique was also utilized for both neutron and photon dose calculations. Two shielding materials, heavy concrete and ordinary concrete, were considered for the shield design. The main goal is to maintain the total dose outside the shield boundary less than 5.0e-03 mSv/h during operation. The shield configuration and parameters of the accelerator building were determined and are presented in this paper.

• Journal of Radiation Protection and Research
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• v.49 no.1
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• pp.50-64
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• 2024

Background: The reference dose coefficients (DCs) of the International Commission on Radiological Protection (ICRP) have been widely used to estimate organ doses of individuals for risk assessments. This approach has been well accepted because individual anatomy data are usually unavailable, although dosimetric uncertainty exists due to the anatomical difference between the reference phantoms and the individuals. We attempted to quantify the individual variation of organ doses for photon external exposures by calculating and comparing organ DCs for 30 individuals against the ICRP reference DCs. Materials and Methods: We acquired computed tomography images from 30 patients in which eight organs (brain, breasts, liver, lungs, skeleton, skin, stomach, and urinary bladder) were segmented using the ImageJ software to create voxel phantoms. The phantoms were implemented into the Monte Carlo N-Particle 6 (MCNP6) code and then irradiated by broad parallel photon beams (10 keV to 10 MeV) at four directions (antero-posterior, postero-anterior, left-lateral, right-lateral) to calculate organ DCs. Results and Discussion: There was significant variation in organ doses due to the difference in anatomy among the individuals, especially in the kilovoltage region (e.g., <100 keV). For example, the red bone marrow doses at 0.01 MeV varied from 3 to 7 orders of the magnitude depending on the irradiation geometry. In contrast, in the megavoltage region (1-10 MeV), the individual variation of the organ doses was found to be negligibly small (differences <10%). It was also interesting to observe that the organ doses of the ICRP reference phantoms showed good agreement with the mean values of the organ doses among the patients in many cases. Conclusion: The results of this study would be informative to improve insights in individual-specific dosimetry. It should be extended to further studies in terms of many different aspects (e.g., other particles such as neutrons, other exposures such as internal exposures, and a larger number of individuals/patients) in the future.

• Nuclear Engineering and Technology
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• v.54 no.11
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• pp.4231-4235
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• 2022

Considering the importance of deuterium in nuclear science including medical and industrial researches such as (BNCT) and nuclear reactors respectively, it is important to study various possible ways in addition to common methods for measuring its concentration. This study is an effort to measure deuterium concentration using PGNAA. The main idea is to calculate the area under 2.23 MeV gamma-rays photo peak resulting from neutron collision with Hydrogen atoms which are in mix with deuterium in samples. The study carried out by both simulation and experiment. Monte Carlo MCNPX2.6 code has been used for simulation and based on its acceptable results an experimental setup has been arranged. The coordination of results was in the range of R = 0.99 and R = 0.98 in simulation and experiment respectively. The accuracy of the study has been investigated by measuring the concentration of an unknown sample by both PGNAA and Fourier transform infrared spectroscopy (FT-IR) methods in which there were acceptable correlation between these two methods.

• Nuclear Engineering and Technology
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• v.51 no.7
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• pp.1828-1834
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• 2019

The performance of a plastic scintillator for use in an in situ measurement system was analyzed using simulation and experimental methods. The experimental results of four major pure beta-emitting radionuclides, namely $^3H$, $^{14}C$, $^{32}P$, and $^{90}Sr/^{90}Y$, were compared with those obtained using a Monte Carlo N-particle (MCNP) code simulation. The MCNP simulation and experimental results demonstrated good agreement for $^{32}P$ and $^{90}Sr/^{90}Y$, with a relative difference of 1.95% and 0.43% between experimental and simulation efficiencies for $^{32}P$ and $^{90}Sr/^{90}Y$, respectively. However, owing to the short range of beta particles in water, the efficiency for $^{14}C$ was extremely low, and $^3H$ could not be detected. To directly measure the low-energy beta radionuclides considering their short range, a system where the source could flow directly to the scintillator was developed. The optimal thickness of the plastic scintillator was determined based on the suggested diameter. Results showed that the detection efficiency decreases with an increase in the depth of the water. The detection efficiency decreased drastically to approximately 10 cm, and the tendency was gradually constant.

• Nuclear Engineering and Technology
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• v.56 no.5
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• pp.1747-1753
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• 2024

The accurate determination of formation density and the physical properties of rocks is the most critical logging tasks which can be obtained using gamma-ray transport and detection tools. Though the simulation works published so far have considerably improved the knowledge of the parameters that govern the responses of the detectors in these tools, recent studies have found considerable differences between the results of using a conventional model of a homogeneous mixture of formation and fluid and an inhomogeneous fractured medium. It has increased concerns about the importance of the complexity of the model used for the medium in simulation works. In the present study, we have suggested two various models for the flow of the fluid in porous media and fractured rock to be used for logging purposes. For a typical gamma-gamma logging tool containing a ¹³⁷Cs source and two NaI detectors, simulated by using the MCNPX code, a simplified porous (SP) model in which the formation is filled with elongated rectangular cubes loaded with either mineral material or oil was investigated. In this model, the oil directly reaches the top of the medium and the connection between the pores is not guaranteed. In the other model, the medium is a large 3-D matrix of 1 cm³ randomly filled cubes. The designed algorithm to fill the matrix sites is so that this realistic random (RR) model provides the continuum growth of oil flow in various disordered directions and, therefore, fulfills the concerns about modeling the rock textures consist of extremely complex pore structures. For an arbitrary set of oil concentrations and various formation materials, the response of the detectors in the logging tool has been considered as a criterion to assess the effect of modeling for the distribution of pores in the formation on simulation studies. The results show that defining a RR model for describing heterogeneities of a porous medium does not effectively improve the prediction of the responses of logging tools. Taking into account the computational cost of the particle transport in the complex geometries in the Monte Carlo method, the SP model can be satisfactory for gamma-gamma logging purposes.

• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1789-1803
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• 2022

RAST-F is a new full-core analysis code based on the two-step approach that couples a multi-group cross-section generation Monte-Carlo code MCS and a multi-group nodal diffusion solver. To demonstrate the feasibility of using MCS/RAST-F for fast reactor analysis, this paper presents the coupled nodal code verification results for the MET-1000 and CAR-3600 benchmark cores. Three different multi-group cross-section calculation schemes are employed to improve the agreement between the nodal and reference solutions. The reference solution is obtained by the MCS code using continuous-energy nuclear data. Additionally, the MCS/RAST-F nodal solution is verified with results based on cross-section generated by collision probability code TULIP. A good agreement between MCS/RAST-F and reference solution is observed with less than 120 pcm discrepancy in k_eff and less than 1.2% root-mean-square error in power distribution. This study confirms the two-step approach MCS/RAST-F as a reliable tool for the three-dimensional simulation of reactor cores with fast spectrum.

• Progress in Medical Physics
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• v.10 no.2
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• pp.73-81
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• 1999

Scatter correction for single photon emission computed tomography (SPECT) plays an important role to improve image quality and quantitation. The purpose of this study was to investigate three scatter correction methods using Monte Carlo simulation. Point source and Jaszack phantom filled with Tc-99m were simulated by Monte Carlo code, SIMIND. For scatter correction, we applied three methods, Compton window (CW) method, triple window (TW) method, and dual photopeak window (DPW) method. Point sources located at various depths along the center line within a 20-cm phantom were simulated to calculate the window ratios and corresponding scatter fractions by evaluating the polynomial coefficients for DPW method. Energy windows were located in W$_1$=92-125 keV, W$_2$=124-126 keV, W$_3$=136-140 keV, W$_4$=140-141 keV, and W$_{5}$=154-156 keV. The results showed that in Jaszack phantom with cold sphere and hot sphere, the TW gave the closest contrast and percentage recovery to the ideal image, respectively, while CW overestimated and DPW underestimated the contrast of ideal one. All three scatter correction methods showed an improved image contrast. In conclusion, scatter correction is essential for improving image contrast and accurate quantification. The choice of scatter correction method should be made on the basis of accuracies and ease of implementation.

• Nuclear Engineering and Technology
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• v.53 no.11
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• pp.3816-3823
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• 2021

The mechanical features evaluated theoretically using Makishima-Mackenzie's model for glasses xBaO-(50-x) PbO-50P₂O₅ where x = 0, 5, 10, 15, 20, 30, 40, and 50 mol%. Wherefore, the elastic characteristics; Young's, bulk, shear, and longitudinal modulus calculated. The obtained result showed an increase in the calculated values of elastic moduli with the replacement of the PbO by BaO contents. Moreover, the Poisson ratio, micro-hardness, and the softening temperature calculated for the investigated glasses. Besides, gamma and neutron shielding ability evaluated for the barium doped lead phosphate glasses. Monte Caro code (MCNP-5) and the Phy-X/PSD program applied to estimate the mass attenuation coefficient of the studied glasses. The decrease in the PbO ratio has a negative effect on the MAC. The highest MAC decreased from 65.896 cm²/g to 32.711 cm²/g at 0.015 MeV for BPP0 and BPP7, respectively. The calculated values of EBF and EABF showed that replacement of PbO with BaO contents in the studied BPP glasses helps to reduce the number of photons accumulated inside the studied BPP glasses.

• KSII Transactions on Internet and Information Systems (TIIS)
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• v.17 no.3
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• pp.999-1021
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• 2023

Differential spatial modulation uses the antenna index to transmit information, which improves the spectral efficiency, and completely bypasses any channel side information in the recommended setting. A generalized distributed multiple turbo coded-cooperative differential spatial modulation based on distributed multiple turbo code is put forward and its performances in Rayleigh fading channels is analyzed. The generalized distributed multiple turbo coded-cooperative differential spatial modulation scheme is a coded-cooperation communication scheme, in which we proposed a new joint parallel iterative decoding method. Moreover, the code matched interleaver is considered to be the best choice for the generalized multiple turbo coded-cooperative differential spatial modulation schemes, which is the key factor of turbo code. Monte Carlo simulated results show that the proposed cooperative differential spatial modulation scheme is better than the corresponding non-cooperative scheme over Rayleigh fading channels in multiple input and output communication system under the same conditions. In addition, the simulation results show that the code matched interleaver scheme gets a better diversity gain as compared to the random interleaver.