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BENCHMARKING OF THE MCNP CODE FOR DOSE ESTIMATION ACCURACY FROM INTERNAL PHOTON SOURCES IN ORNL ANALYTICAL PHANTOMS AT VARIOUS AGES
and anthropomorphic phantoms designed in six age groups, by
MCNP4C Monte Carlo code. In these phantoms we modified the thyroid and neck models. The
results of calculations of Specific Absorbed Fractions (SAFs) from internal photon sources, in...
A plastic scintillator-based 2D thermal neutron mapping system for use in BNCT studies
In this study, a scintillator-based measurement instrument is proposed which is capable of measuring a two-dimensional map of thermal neutrons within a phantom based on the detection of 2.22MeV gamma rays generated via ...
Improving the safety of a body composition analyser based on the PGNAA method
and the optimum radius of the spherical Pb
shield have been investigated, using the MCNP-4C code, and compared with the
unfiltered case, the bare source. Finally, experimental results demonstrate that
an optimised gamma-ray shield...
A study of the effect of the lung shape on the lung absorbed dose in six standard photon and neutron exposure geometries
According to the published results of radiation dosimetry studies, there are significant discrepancies in the organ absorbed doses of existing adult male phantoms. As stated, differences in the organ absorbed doses may be ...
Low-energy neutron flux measurement using a resonance absorption filter surrounding a lithium glass scintillator
filter surrounding a miniature-type lithium glass scintillator. the count with and without the filter surrounding the detector gives the number of resonance-energy neutrons. Some preliminary results and a comparison with the MCNP code are shown....
Modeling an HPGe detector response to gamma-rays using MCNP5 code
been developed by using the MCNP5 code. To validate the simulated model, the simulations from mono-energetic sources have been compared to the corresponding measured data. Any deviation from the measurement could be attributed to the unmodeled details...
محاسبه دز مؤثر در سانحه هسته ای فوکوشیما با استفاده از کد محاسباتی MCNP و مقایسه با نتایج تحلیلی
سانحه هسته ای فوکوشیمای ژاپن سبب انتشار مقادیر قابل ملاحظه ای ایزوتوپه ای پرتوزا در محیط اطراف شده است.
حضور این ایزوتوپ ها در محیط اطراف مشکلات اساسی را برای انسان به وجود می آورد. از این رو با بهره گیری از فانتوم ...
Neutron spectrum measurements from 1-16 MeV in beryllium assemblies with a central D-T neutron source
compared with MCNP Monte Carlo calculations using the ENDF/B-VI data set. For all three
shells the experimental results lie above those calculated for neutron energies between 8 and
11 MeV, whilst between 1 and 4 MeV they lie below. It is concluded...
شبیه سازی مدل کامل سر و گردن در فانتوم بزرگسال براساس مدل MIRD-15 با استفاده از روش مونت کارلو و کد شبیه سازی MCNP4C و محاسبه دز رسیده از چشمه خارجی 252Cf بر اعضای این مدل با استفاده از کد MCNPX
با توجه به حضور پرتوها در زندگی بشر سترسی به روش های برتر دزسنجی اهمیت فراوانی پیدا می کند. تابش های یوننده اثر بیولوژیکی مخربی روی سلول های بدن دارند. برای بررسی این اثرات بایستی دز رسیده به بافت های بدن محاسبه شود. مدل ...
دز سنجی پرتوهای مصنوعی در محیط زیست با استفاده از کد MCNP
Nuclear disaster spread radioisotopes greatly to the environment. The isotopes fall on the surface earth, radiate and pollute the environment. So dosimetric data for estimating the accident is necessary. For this reason, ...