نتایج جستجو برای: mcnp simulation code
تعداد نتایج: 704198 فیلتر نتایج به سال:
in this study, a general theoretical model to foresee and calculate the current-voltagecharacteristics in a plateau zone and the associated efficiency of the neutron detector (sensitivity) is presented. this study is to complete the previous studies in this field. the model also considers electric field distortion resulting from the charge collection effect. the characteristics curve is obtaine...
The Neutron induced activity and decay heat are very important for the nuclear design of fusion reactor. In this paper, activation analysis of DEMO reactor with helium-cooled solid blanket (HCSB-DEMO) was performed. Ceramics Li4SiO4 and Beryllium options were considered as tritium breeding material and neutron multiplier, respectively. Chinese low-activation ferritic steel (CLF-1) developed in ...
In this work, the Isfahan Miniature Neutron Source Reactor (MNSR) is first simulated using the WIMSD code, and its fuel burn-up after 7 years of operation ( when the reactor was revived by adding a 1.5 mm thick beryllium shim plate to the top of its core) and also after 14 years of operation (total operation time of the reactor) is calculated. The reactor is then simulated using the MCNP code,...
Detonation initiation by shock is an important issue in the explosive safety assessment and design of the explosive train and explosive devices. Experimental studies in this area are very difficult, expensive, and require advanced equipment. Therefore, simulation is a useful and suitable way for studying this phenomenon. The purpose of this article is to develop a one-dimensional computer code ...
Background: Lead-based radiation shields are widely used in radiology departments to protect both workers and patients from any unnecessary exposure to ionizing radiation. Recently there has been a great deal of concern expressed about the toxicity of lead. Human lead toxicity is well documented. In that light, production of environmentally-friendly lead-free radiation shields with less weight ...
The MCNP code is a general Monte Carlo N-Particle Transport program that is widely used in health physics, medical physics and nuclear engineering for problems involving neutron, photon and electron transport. However, due to the stochastic nature of the algorithms employed to solve the Boltzmann transport equation, MCNP generally exhibits a slow rate of convergence. In fact, engineers and scie...
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