Fuel burnup analysis for Thai research reactor by using MCNPX computer code

Determination ofresearch reactor fuel burnup

Comparison of (Th-233U) O2 and (Th-235U) O2 fuel burn up into a thermal research reactor using MCNPX 2.6 code

Background: Decrease of economically accessible uranium resources motivates consideration of breeding of fertile elements such as thorium. Material and Method: Thorium oxide fuel burn up calculation of a simulated research reactor cooled heavy water has been proposed in the present work using MCNPX 2.6 code. Two 233U and 235U isotopes have been used as fissile element of thorium oxide fuel. 135...

Flow Blockage Accident Analysis of Tehran Research Reactor Fuel Assembly

Tehran Research Reactor (T.R.R.) is a pool-type, 5 MW thermal research reactor. One probable event is that if some external objects or debris fall down into the reactor core and cause obstruction of the coolant flow through one of the fuel assemblies, decreasing the surface flow area, ceases the coolant flow, and also raises the fuel and sheaths tempe...

Designed free-lead composite shields for diagnostic medicine radiation fields by using MCNPX code

 Introduction: Nowadays, applications of free-lead radiation shields are increased specially for weighting factor and toxicity of lead elements. Although touching the lead may be results in the nervousand blood system of body, it is not exactly proven yet. Protective measures are the fundamental parts of radiation protection. On the other hand, high weighting factor of lead sh...

Proton Therapy of eye using MCNPX code

Introduction: Proton radiotherapy is the one of advanced teletherapy methods. The protons deposit their maximum energy in a position called Bragg peak. Therefore, for treatment of cancer, the tumor should be placed at the Bragg peak or SOBP. The scattered photons and neutrons is a challenge in proton radiotherapy. The aim of this study is calculation of absorbed dose from scatt...