5318

4988

كارشناسي ارشد

فيزيك

اصفهان: دانشگاه صنعتي اصفهان، دانشكده فيزيك

1388

[ده]،154ص.:مصور﴿بخش رنگي﴾،جدول،نمودار

ص.ع.به فارسي و انگليسي

احمد شيراني

محمد حسن علامت ساز

13/4/89

اكبر پروازيان، ظفرالله كلانتري

1396/10/03

كتابنامه

فيزيك

ID4988

به فارسي و انگليسي: قابل رويت در نسخه ديجيتالي

Determination of Thermal Neutron Flux for Neutron Sources in Graphite and Water Moderators using MCNPX Code and Determination of The Counting Calibration Coefficient of BF3 Student name Hasan Vosoughian Email address h vosoughian@ph iut ac ir Date of submission February 19 2010 Department of Physics Isfahan University of Technology Isfahan 84156 83111 IranDegree M Sc Language FarsiSupervisor Dr Ahmad Shirani Email shirani@cc iut ac irAbstract Neutron beam can be produced in different nuclear reactions One can not accelerate neutronslike charged particles but beginning with high energy neutron we can obtain a vast range oflower energy by colliding them to the environment atoms like graphite and water Theresulted neutrons can have very low energy so one can produce a source of thermalneutrons less than 1 eV There in Indian Atomic Centre BHABHA by putting 6 neutronsources Am Be in a graphite moderator they produced a thermal source and used it forcalibrating the thermal neutron detectors Neutron flux distribution in the above mentionedCentre have been measured by a neutron activation procedure In the first part of this thesis we calculated the neutron flux distribution in the above mentioned structure by MCNPX codeand the output results were compared with experimental data from BHABHA Atomic Centre Measured results were in good agreement with the experimental data from BHABHA centre which shows that our simulation program is accurate enough to set up a similar calibrationfor the neutron detectors Then using the code and neutron source moderated by water atMALEK ASHTAR UNIVERSITY we calculated counting calibration coefficient of BF3detector For this purpose we firs simulated a neutron source in water environment and then in a certain distance from the source the neutron flux was calculated Using the calculatedflux we determined the counting of the detector In the second part by putting a detector inthe place we measured counting rate experimentally and at last using the measured and thecalculated values we determined the counting calibration coefficient Key words MCNPX Neutron Source BF3 detector Calibration Coefficient GraphiteModerator

احمد شيراني

محمد حسن علامت ساز

اكبر پروازيان، ظفرالله كلانتري