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

International Nuclear Science and Technology Conference 2016

Sangkaew, S.^1 ; Angwongtrakool, T.^1 ; Srimok, B.^1

Bureau of Nuclear Safety Regulation, Office of Atoms for Peace, 16 Vibhawadee-Rangsit Road, LadYao, Chatuchak, Bangkok

10900, Thailand^1

关键词: Bangkok , Thailand;    Computer codes;    Excess reactivity;    Irradiation position;    Modelling softwares;    Nuclear technology;    Power distributions;    Radiation transport;   

Others  :  https://iopscience.iop.org/article/10.1088/1742-6596/860/1/012033/pdf
DOI  :  10.1088/1742-6596/860/1/012033

来源: IOP

【 摘 要 】

This paper presents the fuel burnup analysis of the Thai research reactor (TRR-1/M1), TRIGA Mark-III, operated by Thailand Institute of Nuclear Technology (TINT) in Bangkok, Thailand. The modelling software used in this analysis is MCNPX (MCNP eXtended) version 2.6.0, a Fortran90 Monte Carlo radiation transport computer code. The analysis results will cover the core excess reactivity, neutron fluxes at the irradiation positions and neutron detector tubes, power distribution, fuel burnup, and fission products based on fuel cycle of first reactor core arrangement.