The interaction of fast neutrons with shielding and fusion blanket materials

In the present work the neutron emission spectra from a graphite cube, and from natural uranium, lithium fluoride, graphite, lead and steel slabs bombarded with 14.1 MeV neutrons were measured to test nuclear data and calculational methods for D - T fusion reactor neutronics. The neutron spectra measured were performed by an organic scintillator using a pulse shape discrimination technique based on a charge comparison method to reject the gamma rays counts. A computer programme was used to analyse the experimental data by the differentiation unfolding method. The 14.1 MeV neutron source was obtained from T(d,n)4He reaction by the bombardment of T - Ti target with a deuteron beam of energy 130 KeV. The total neutron yield was monitored by the associated particle method using a silicon surface barrier detector. The numerical calculations were performed using the one-dimensional discrete-ordinate neutron transport code ANISN with the ZZ-FEWG 1/ 31-1F cross section library. A computer programme based on Gaussian smoothing function was used to smooth the calculated data and to match the experimental data. There was general agreement between measured and calculated spectra for the range of materials studied. The ANISN calculations carried out with P3 - S8 calculations together with representation of the slab assemblies by a hollow sphere with no reflection at the internal boundary were adequate to model the experimental data and hence it appears that the cross section set is satisfactory and for the materials tested needs no modification in the range 14.1 MeV to 2 MeV. Also it would be possible to carry out a study on fusion reactor blankets, using cylindrical geometry and including a series of concentric cylindrical shells to represent the torus wall, possible neutron converter and breeder regions, and reflector and shielding regions

Comparison between the calculated and measured dose distributions for four beams of 6 MeV linac in a human-equivalent phantom

Radiation dose distributions in various parts of the body are of importance in radiotherapy. Also, the percent depth dose at different body depths is an important parameter in radiation therapy applications. Monte Carlo simulation techniques are the most accurate methods for such purposes. Monte Carlo computer calculations of photon spectra and the dose ratios at surfaces and in some internal organs of a human equivalent phantom were performed. In the present paper, dose distributions in different organs during bladder radiotherapy by 6 MeV X-rays were measured using thermoluminescence dosimetry placed at different points in the human-phantom. The phantom was irradiated in exactly the same manner as in actual bladder radiotherapy. Four treatment fields were considered to maximize the dose at the center of the target and minimize it at non-target healthy organs. All experimental setup information was fed to the MCNP-4b code to calculate dose distributions at selected points inside the proposed phantom. Percent depth dose distribution was performed. Also, the absorbed dose as ratios relative to the original beam in the surrounding organs was calculated by MCNP-4b and measured by thermoluminescence dosimetry. Both measured and calculated data were compared. Results indicate good agreement between calculated and measured data inside the phantom. Comparison between MCNP-4b calculations and measurements of depth dose distribution indicated good agreement between both

Effectiveness of X and Gamma Rays for Scanning Cargo Containers

This paper describes the work performed to assess the effectiveness of scanners based on X-ray and gamma-ray transmission techniques for scanning cargo containers before entering different ports. Scanning by X-rays was done using X-ray machine operating at  135 KV and current of 4 mA. However, Scanning by gamma-rays was performed using collimated slit beam of 1 cm width and 5 cm height emitted from 0.5 Ci Cobalt-60 source. The transmitted gamma-rays through the container and the hidden object was measured by a gamma counting system applies NaI(Tl) detector. Objects of different physical and chemical composition were used in this study. The obtained images by X-ray transmission technique show incapability to distinguish between objects of nearly the same density. However the obtained images resulting from scanning by gamma-rays shows more capability to distinguish between all the examined objects

Analysis of Fuel Burnup and Transmutations at High Burnup of Sodium Fast Breeder Reactor

In this paper, the Monte Carlo N-Particle extended  computer code (MCNP) were used to design a model of the European Sodium-cooled Fast Reactor. The multiplication factor, conversion factor, delayed neutrons fraction, doppler constant, control rod worth, sodium void worth, masses for major heavy nuclei, radial and axial power distribution at high burnup are studied. The results show that the reactor breeds fissile isotopes with a conversion ratio of 0.994 at fuel burnup 70 (GWd/T), and minor actinides are buildup inside the reactor core. The study aims to check the efficiency of the model on the calculation of the neutronic parameters of the core at high burnup

An In-Depth Examination of the Natural Radiation and Radioactive Dangers Associated with Regularly Used Medicinal Herbs

The specific activity of U-238 and Th-232, as well as K-40 radionuclides, in twenty-nine investigated medicinal herbs used in Egypt has been measured using a high-purity germanium (HP Ge) detector. The measured values ranged from the BDL to 20.71 ± 1.52 with a mean of 7.25 ± 0.54 (Bq kg−1) for uranium-238, from the BDL to 29.35 ± 1.33 with a mean of 7.78 ± 0.633 (Bq kg−1) for thorium-232, and from 172 ± 5.85 to 1181.2 ± 25.5 with a mean of 471.4 ± 11.33 (Bq kg−1) for potassium-40. Individual herbs with the highest activity levels were found to be 20.71 ± 1.52 (Bq kg−1) for uranium-238 (H4, Thyme herb), 29.35 ± 1.33 (Bq kg−1) for thorium-232 (H20, Cinnamon), and 1181.2 ± 25.5 (Bq kg−1) for potassium-40 (H24, Worm-wood). (AACED) Ingestion-related effective doses over the course of a year of uranium-238 and thorium-232, as well as potassium-40 estimated from measured activity concentrations, are 0.002304 ± 0.00009 (minimum), 0.50869 ± 0.0002 (maximum), and 0.0373 ± 0.0004 (average)(mSv/yr). Radium equivalent activity (Raeq), annual gonadal dose equivalent (AGDE), absorbed gamma dose rate (Doutdoor, Dindoor), gamma representative level index (I), annual effective dose (AEDtotal), external and internal hazard index (Hex, Hin), and excess lifetime cancer risk were determined in medicinal plants (ELCR). The radiological hazards assessment revealed that the investigated plant species have natural radioactivity levels that are well within the internationally recommended limit. This is the first time that the natural radioactivity of therapeutic plants has been measured in Egypt. In addition, no artificial radionuclide (for example, 137Cs) was discovered in any of the samples. Therefore, the current findings are intended to serve as the foundation for establishing a standard safety and guideline for using these therapeutic plants in Egypt. © 2022 by the authors. Licensee MDPI, Basel, Switzerland.PNURSP2022R173This work was funded by Princess Nourah bint, Abdulrahman University, Research Supporting Project number (PNURSP2022R173) Princess Nourah bint, Abdulrahman University, Riyadh, Saudi Arabia

Prediction of mechanical and radiation parameters of glasses with high Bi2O3 concentration

This study aims to perform multidirectional characterizations on nuclear shielding efficiencies on some bismuth-based glasses. Accordingly, the γattenuation coefficients for xBi2O3-(75-x)B2O3–25Li2O (x = 0, 10, 20, 30, 40, 50, 60 the 70 mol%) were widely evaluated using simulations and theoretical methods. Linear attenuation coefficient (LAC) of the glasses was obtained by the Monte Carlo general-purpose simulation code FLUKA and compared with the XCOM database up to 15 MeV. Moreover, LAC values have been utilized to evaluate related parameters like mass attenuation coefficient (MAC), total molecular cross-section (σt), total atomic cross-section (σa), half-value layer (HVL), total electronic cross-section (σe), mean free path (MFP), effective atomic number (Zeff), and effective electron density (Neff). The results noted that the XCOM and FLUKA data of the shielding parameters are in great agreement. Relatively higher density (5.818 g/cm3), greater LAC, MAC, Zeff, and lower HVL, MFP values are achieved for 70Bi2O3-5B2O3–25Li2O glass. Accordingly, this glass sample is a better gamma shield. © 2021 The AuthorsTaif University Researchers Supporting Project number (TURSP-2020/45) Taif University, Taif, Saudi Arabia

The interaction of fast neutrons with shielding and fusion blanket materials

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