4569

4297

كارشناسي ارشد

فيزيك

اصفهان: دانشگاه صنعتي اصفهان، دانشكده فيزيك

1387

چهارده، 168 ، [II] ص:نمودار ، جدول

ص.ع. به: فارسي وانگليسي

احمد شيراني

ايرج شهابي

88/4/31

اكبر پروازيان

1396/09/12

كتابنامه

فيزيك

ID4297

به فارسي وانگليسي: قابل رويت در نسخه ديجيتال

157Calculation of Reactivity Variations of the MNSR Reactor Due toVariations in the Thickness of the Core Top Beryllium Layer Using theWIMS and MCNP Codes and Comparsion with Experimental Results Leila ranjbar l ranjbar@ph iut ac ir 10 March 2009 Department of Physics Isfahan University of Technology Isfahan 84156 83111 IranDegree M Sc Language FarsiAhmad Shirani Shirani@cc iut ac ir AbstractThe Miniature Neutron Source Reactor MNSR is manufactured by the Chinese Institute of Atomic Energy It is a low power tank in pool type research reactor and similar to the canadian SLOWPOKE reactor Thereactor employs highly enriched uranium as fuel 90 2 light water as moderator and coolant and metalberyllium as reflector Heat generated in the core is removed through natural convection There are 10irradiation sites in the reactor Five of them are inside the annulus beryllium reflector and the other fivesurround the annulus reflector externally The maximum thermal neutron fluxes in the inner and the outersites are 1 0 1012 and 5 0 1011 n cm2 s respectively The nominal thermal power of the reactor at thisneutron flux level is 30 kW The MNSR reactor is used as a source of neutrons for neutron activation analysisat research institutions universities and hospitals The cold excess reactivity was adjusted to less than 4 mkusing three reactivity regulators The excess reactivity of the MNSR decreases with the reactor operating timedue to the fuel burn up and accumulation of fission products in the reactor core During the processes ofirradiation and fuel burn up fissile nuclides are consumed by fission and several hundred fission products areformed some directly and other by radioactive decay In this thesis the Isfahan Miniature Neutron Source Reactor MNSR was simulated using the WIMS codeand the calculations of fuel burn up was performed after 7 years of the reactor operation time reactor revivaltime by adding the top beryllium shim plate the results of which were consistent with the expected resultsfrom other methods Also the MNSR reactor was simulated using the MCNP code that works on the basis ofthe Monte Carlo method and after confidence from the accuracy of the simulated program the results obtainfrom the WIMS code were incorporated in the input of the MCNP program the necessity of adding the 1 5millimeter of top beryllium shim plate after 7 years of the reactor operation time was studied and then byadding the 1 5 millimeter beryllium layer at the top of the reactor core in the input of the MCNP program reactivity variations was studied in new status the results of which were consistent with experimental results

احمد شيراني

ايرج شهابي