Neutronic analysis and validation of boiling water reactor core designed by MCNPX code. (February 2015)

• Record Type: Journal Article
• Authors: Galahom, Ahmed Abdelghafar; Bashter, I.I.; Aziz, Moustafa
• Is Part Of: Annals of nuclear energy. Volume 76(2015:Feb.)
• Page Start: 461
• Page End: 468
• Publication Date: 2015-02
• Subjects: Boiling water reactor core -- Neutronic characteristics -- MCNPX -- Normalized power -- Thermal neutron flux -- Pu concentration; Nuclear energy -- Periodicals; Nuclear engineering -- Periodicals; 621.4805
• DOI: 10.1016/j.anucene.2014.10.025
• Languages: English
• ISSNs: 0306-4549

Abstract

Highlights: MCNPX code is used to design a model for BWR core. The fuel enrichment is distributed in such a way to flat the power. Validation of the BWR core model designed by MCNPX code. Calculate Pu and its isotopes concentration at different burnup. Abstract: This paper presents a design of boiling water reactor BWR model using MCNPX to develop benchmarks for checking the fuel management computer code packages. MCNPX code based on Monte Carlo method, is used to design a three dimensional model for BWR fuel assembly in typical operating temperature and pressure conditions. A test case was compared with a benchmark problem and good agreement was found. This design is used to study the thermal neutron flux and the pin by pin power distribution through the BWR core assemblies. The fuel used in BWR core is UO2 with three different types of enrichment (0.711%, 1.76% and 2.19%). This enrichment is distributed in such a way as to flatten the power. The effect of different enrichment values on the radial normalized power distribution is analyzed. The spent fuel in the reactor can be recycled, and plutonium and its isotopes can be extracted.