title: Performance of the MTR core with MOX fuel using the MCNP4C2 code.
abstract: The MCNP4C2 code was used to simulate the MTR-22 MW research reactor and perform the neutronic analysis for a new fuel namely: a MOX (U3O8&PuO2) fuel dispersed in an Al matrix for One Neutronic Trap (ONT) and Three Neutronic Traps (TNTs) in its core. Its new characteristics were compared to its original characteristics based on the U3O8-Al fuel. Experimental data for the neutronic parameters including criticality relative to the MTR-22 MW reactor for the original U3O8-Al fuel at nominal power were used to validate the calculated values and were found acceptable. The achieved results seem to confirm that the use of MOX fuel in the MTR-22 MW will not degrade the safe operational conditions of the reactor. In addition, the use of MOX fuel in the MTR-22 MW core leads to reduce the uranium fuel enrichment with (235)U and the amount of loaded (235)U in the core by about 34.84% and 15.21% for the ONT and TNTs cases, respectively.
authors.affiliation: Nuclear Engineering Dept., Atomic Energy Commission, P.O.Box 6091, Damascus, Syria. Electronic address: pscientific10@aec.org.sy.
authors.forename: Ismail
authors.initials: I
authors.lastname: Shaaban
authors.affiliation: Nuclear Engineering Dept., Atomic Energy Commission, P.O.Box 6091, Damascus, Syria.
authors.forename: Mohamad
authors.initials: M
authors.lastname: Albarhoum
keywords: Criticality safety; MCNP4C2 code; MOX fuel; MTR-22 MW reactor; Neutronic parameters