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Monte Carlo Simulation of BN-600 LMFR Hybrid Core

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材料科学与工程（中英文B版） 2011年第6期1卷 838-842页

作者： mohga I. hassan moustafa aziz National Center for Nuclear Safety and Radiation Control Nasr City Egypt

MCNPX code is used to design a three dimensional model to simulate and study the whole hybrid core of the BN-600 prototype LMFR. In the model, the core is composed of eight radial zones and eight axial zones. The spectrum of neutron flux is divided into 23 energy groups. Results from this model were verified using published results of benchmark problems that utilized diffusion or transport theories in calculation and found acceptable. The maximum difference was about 4% for multiplication factor, 1% for control rod worth, and 2.6% for radial power distribution. The model is used to calculate power distributions at different zones of the reactor core. Partial insertion of control rods is simulated and analyzed. The effect of burnup on reactivity is also determined in typical operating conditions of the reactor core. Burnup reactivity loss is about 2.7%.

关键词：蒙特卡罗模拟混合动力三维模型计算理论功率分布中子通量操作条件控制棒

Antifouling evaluation of extracts from Red Sea soft corals against primary biofilm and biofouling

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Asian Pacific Journal of Tropical Biomedicine 2017年第11期7卷 991-997页

作者： Yosry Abdel aziz Soliman Ahmed Mohammed Brahim Ahmed Hussein moustafa Mohamed Abdel Fattah Hamed Marine Chemistry Lab. Marine Environment DivisionNational Institute of Oceanography and FisheriesSuezEgypt Department of Chemistry Faculty of ScienceZagazig UniversityZagazigEgypt

Objectives: To evaluate antifouling property of extracts from Red Sea soft corals against primary biofilm and ***: Seven species of soft corals Sarcophyton glaucum(a), Sinularia compressa,Sinularia cruciata(a), Heteroxenia fuscescens(a), Sarcophyton glaucum(b), Heteroxenia fuscescens(b) and Sinularia cruciata(b) were chosen to test their extracts as antibacterial and antifouling agents in Eastern Harbour of Alexandria, Mediterranean *** compounds of soft corals were extracted by using methanol and concentrated under vacuum. The residues of extracts were mixed in formulation of inert paint which consisted of rosin, chlorinated rubber and ferrous oxide against micro and macro fouling organisms. The formulated paints were then applied on PVC panels twice by brush,hanged in a steel frame and immersed in Eastern Harbour of Alexandria Mediterranean Sea followed by visual inspection and photographic ***: After 185 days of immersion in seawater, the antifouling results agreed with the antibacterial results where extracts of Sinularia compressa and Heteroxenia fuscescens(b) gave the best activity against marine fouling tubeworms and barnacles. The inhibition activity was correlated with the major functional groups(hydroxyl, amino, carbonyl,aliphatic(fatty acids), C]C of alkene or aromatic rings and CeC l of aryl halides) of the ***: The strong antifouling activity makes them promising candidates for new antifouling additives. After the screening and application of natural organic compounds from soft corals, marine organisms show activity against micro and macro fouling organisms.

关键词： Antibacterial Antifouling Paints Seawater Marine Organisms

Calculations of Radiations Attenuation and Dose Assessments of Tunnels Passing under Water Canals

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Journal of Environmental Science and Engineering(A) 2015年第11期4卷 607-613页

作者： Mohamed Farahat moustafa aziz Department of Siting and Environment Egyptian Nuclear and Radiological Regulatory Authority Cairo 11787 Egypt Department of Safety Engineering Egyptian Nuclear and Radiological Regulatory Authority Cairo 11787 Egypt

Most of the developed countries have used their tunnels as protective structures （public nuclear shelters） in case of nuclear emergencies to protect their citizens from the dangerous effects of nuclear weapons. The research aims to explain how to use tunnels to protect some people from neutrons and gamma rays and account the required safe depth. The computer code （MCNP5） is used at this model for such tunnel to account attenuation of both neutrons and gamma rays passing through the canal water, sand, soil and reinforced concrete wall layers. The last one （thickness 105 cm） constructed the tunnel construction. Also, the computer code is used to account the dose inside the tunnel, and account （neutron） dose, （neutron, gamma） dose, prompt （gamma） dose, total （gamma） dose and total （neutron ＋ gamma） dose estimated by μsv/h, at different depths from the water surface level （depths 200 cm, 500 cm, 1,000 cm, 1,600 cm, 2,200 cm, 2,600 cm, 2,800 cm, 3,400 cm, 3,700 cm, 4,000 cm and 4,600 cm）. Then, account these doses for three bombs （its intensity 20 KT, 100 KT and 1,000 KT）.

关键词： Calculations of radiations radiations attenuation dose assessments tunnels water canals.

Validation of the Monte Carlo Model Designed to Simulate the Neutronic Characteristics of Advanced Boiling Water Reactor Assembly

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Journal of Physical Science and Application 2014年第5期4卷 310-316页

作者： Ahmed Abdelghafar Galahom Ibrahim Ismail Bashter moustafa aziz Higher Technological Institute l Oth of Ramadan City Sharkia Egypt Zagazig University Faculty of Science Physics Department Zagazig City Sharkia Egypt Nuclear Regulatory Commission and Radiological Nasr City Cairo Egypt

In the last few years, interest in burnup calculations using Monte Carlo methods has increased. Previous burnup codes have used diffusion theory for the neutronic portion of the codes. Diffusion theory works well for most reactors. However, diffusion theory does not produce accurate results in burnup problems that include strong absorbers or large voids. MCNPX code based on Mont Carlo Method, is used to design a three dimensional model for a BWR fuel assembly in a typical operating temperature and pressure conditions. A test case was compared with a benchmark problem and good agreement was found. The model is used to calculate the distribution of pin by pin power and flux inside the assembly. The effect of axial variation of water （coolant） density, and of control rods motion on the neutron flux and power distribution is analyzed. The effect of addition of Gd2O3 to natural uranium （0.711%） on both the thermal neutron flux and normalized power are analyzed. The concentration of U^235, U^238, Pu^239, and its isotopes is also calculated at burn-up 50 GWD/T.

关键词： MCNPX Code boiling water reactor thermal neutron flux normalized power multiplication factor.

Radiation Shielding Analysis for Pressurized Heavy Water Reactors (CANDU) Using MCNPX Code

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材料科学与工程（中英文B版） 2022年第2期12卷 50-57页

作者： Afrah El-Khawlani moustafa aziz Ali Ellithi Faculty of Science Sana’a UniversitySana’a 72738Yemen Nuclear and Radiological Regulatory Authority Nasr City 11762Cairo College of Science Cairo UniversityCairo 12613Egypt

MCNPX(Monte Carlo N-Particle Transport Code)computer code is used to design a model to CANDU(Canada Deuterium Uranium)reactor core and its shielding *** is assumed that reactor core is fueled with natural *** core radiation sources are calculated which consider prompt neutrons,neutron induced gamma and prompt gamma *** total neutron flux and dose rate are calculated along the shield and at outer shield *** results indicated that the major dose rate component at outer shield points is due to neutron induced gamma dose rate(μSv/h).

关键词： CANDU reactor MCNPX code reactor shielding natural uranium radiation source

Analysis of Neutronic Characteristics of Uranium Zirconium Hydride Fuel in Advanced Boiling Water Reactor

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材料科学与工程（中英文A版） 2013年第6期3卷 437-442页

作者： Ahmed Abdelghafar Galahom Ibrahim Ismail Bashter moustafa aziz Higher Technological Institute l Oth of Ramadan city Sharkia Egypt Physics Department Faculty of Science Zagazig University Zagazig City Sharkia Egypt Nuclear Regulatory Commission and Radiological Regulatory Authority Nasr City Egypt

This paper presents a design of an advanced boiling water reactor (ABWR) equilibrium bundle using UZrH1.6 as a fuel. Monte Carlo N-Particle Transport (MCNPX) code extended has been used for design advanced boiling water reactor model. MCNPX code based on Mont Carlo Method is used to design a three dimensional model for a BWR fuel assembly in a typical operating temperature and pressure conditions. A test case was compared with a benchmark problem, and good agreement was found. This model is used to calculate the distribution of pin by pin power and flux inside the assembly, calculate the axial thermal neutron flux and the axial normalized power across the fuel rod at different insertion of control rod, and to study the effect of increasing the fuel enrichment on the axial thermal neutron flux and the axial normalized power.

关键词：先进沸水堆燃料组件中子通量氢化锆蒙特卡罗法特性三维模型 ABWR

Analysis of Neutronic Parameters for Supercell of CANDU Reactor Using MCNPX Code

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材料科学与工程（中英文B版） 2013年第8期3卷 550-553页

作者： Afrah EL-Khawlani moustafa aziz Ibrahim Ali Ellithi College of Science Cairo University Cairo 12613 Egypt Physics Department Faculty of Science High Education Sana 'a Yemen Nuclear and Radiological Regulatory A uthority Nasr City Cairo

MCNPX code has been used for modeling and simulation of a supercell of CANDU Fuel, the supercell consists of two fuel bundle and adjuster rod. The fuel bundle are burnt in normal operation conditions of CANDU reactors. Natural uranium fuel is used in the model. The multiplication factor of the bundle is calculated during fuel burnup. The concentration of both uranium and plutonium isotopes are analysed in the bundle. The worth of the adjuster rod is calculated. Comparison of multiplication factor and worth of the adjuster rod with the previous published references showed good agreement.

关键词：代码分析超级单体 CANDU堆中子学 CANDU反应堆燃耗计算铀燃料运行条件

Analysis of Reactivity Induced Accident for Control Rods Ejection with Loss of Cooling

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材料科学与工程（中英文B版） 2013年第2期3卷 128-137页

作者： Hend Mohammed El Sayed Saad Hesham Mohammed Mohammed Mansour moustafa aziz Abd ElWahab Nuclear and Radiological Regulatory Authority Nasr City Cairo 11762 Egypt Physics Department Faculty of Science Cairo University Giza 12613 Egypt

Understanding of the time-dependent behavior of the neutron population in nuclear reactor in response to either a planned or unplanned change in the reactor conditions, is a great importance to the safe and reliable operation of the reactor. In the present work, the point kinetics equations are solved numerically using stiffness confinement method (SCM). The solution is applied to the kinetics equations in the presence of different types of reactivities and is compared with different analytical solutions. This method is also used to analyze reactivity induced accidents in two reactors. The first reactor is fueled by uranium and the second is fueled by plutonium. This analysis presents the effect of negative temperature feedback with the addition positive reactivity of control rods to overcome the occurrence of control rod ejection accident and damaging of the reactor. Both power and temperature pulse following the reactivity- initiated accidents are calculated. The results are compared with previous works and satisfactory agreement is found.

关键词：诱导反应控制棒动力学方程冷却弹射弹棒事故温度反馈反应器