Burnup calculations for KIPT accelerator driven subcritical facility using Monte Carlo computer codes-MCB and MCNPX.

Technical Report · Tue Jun 09 04:00:00 EDT 2009

DOI:https://doi.org/10.2172/956924· OSTI ID:956924

Gohar, Y; Zhong, Z; Talamo, A

Argonne National Laboratory (ANL) of USA and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the conceptual design development of an electron accelerator driven subcritical (ADS) facility, using the KIPT electron accelerator. The neutron source of the subcritical assembly is generated from the interaction of 100 KW electron beam with a natural uranium target. The electron beam has a uniform spatial distribution and electron energy in the range of 100 to 200 MeV. The main functions of the subcritical assembly are the production of medical isotopes and the support of the Ukraine nuclear power industry. Neutron physics experiments and material structure analyses are planned using this facility. With the 100 KW electron beam power, the total thermal power of the facility is {approx}375 kW including the fission power of {approx}260 kW. The burnup of the fissile materials and the buildup of fission products reduce continuously the reactivity during the operation, which reduces the neutron flux level and consequently the facility performance. To preserve the neutron flux level during the operation, fuel assemblies should be added after long operating periods to compensate for the lost reactivity. This process requires accurate prediction of the fuel burnup, the decay behavior of the fission produces, and the introduced reactivity from adding fresh fuel assemblies. The recent developments of the Monte Carlo computer codes, the high speed capability of the computer processors, and the parallel computation techniques made it possible to perform three-dimensional detailed burnup simulations. A full detailed three-dimensional geometrical model is used for the burnup simulations with continuous energy nuclear data libraries for the transport calculations and 63-multigroup or one group cross sections libraries for the depletion calculations. Monte Carlo Computer code MCNPX and MCB are utilized for this study. MCNPX transports the electrons and the produced neutrons and photons but the current version of MCNPX doesn't support depletion/burnup calculation of the subcritical system with the generated neutron source from the target. MCB can perform neutron transport and burnup calculation for subcritical system using external neutron source, however it cannot perform electron transport calculations. To solve this problem, a hybrid procedure is developed by coupling these two computer codes. The user tally subroutine of MCNPX is developed and utilized to record the information of the each generated neutron from the photonuclear reactions resulted from the electron beam interactions. MCB reads the recorded information of each generated neutron thorough the user source subroutine. In this way, the neutron source generated by electron reactions could be utilized in MCB calculations, without the need for MCB to transport the electrons. Using the source subroutines, MCB could get the external neutron source, which is prepared by MCNPX, and perform depletion calculation for the driven subcritical facility.

• Research Organization: Argonne National Laboratory (ANL)
• Sponsoring Organization: NE
• DOE Contract Number: AC02-06CH11357
• OSTI ID: 956924
• Report Number(s): ANL-09/07
• Country of Publication: United States
• Language: ENGLISH

Related Subjects

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
29 ENERGY PLANNING, POLICY, AND ECONOMY
43 PARTICLE ACCELERATORS
ACCELERATORS
BURNUP
COMPUTER CODES
COMPUTERS
CROSS SECTIONS
ELECTRON BEAMS
ELECTRON REACTIONS
FISSILE MATERIALS
FISSION
FISSION PRODUCTS
FUEL ASSEMBLIES
NEUTRON FLUX
NEUTRON SOURCES
NEUTRON TRANSPORT
NUCLEAR POWER
PHOTONUCLEAR REACTIONS
SPATIAL DISTRIBUTION
TRANSPORT