TY - GEN

T1 - Effects of a multi-physics coupling approach in monte carlo burnup calculations

AU - Castagna, Christian

AU - Cervi, Eric

AU - Lorenzi, Stefano

AU - Cammi, Antonio

AU - Chiesa, Davide

AU - Nastasi, Massimiliano

AU - Sisti, Monica

AU - Previtali, Ezio

N1 - Publisher Copyright:
Copyright © (2018) by PHYSOR 2018.

PY - 2018/1/1

Y1 - 2018/1/1

N2 - In nuclear reactor analysis, a relevant challenge is to achieve a suitable global description of nuclear systems through the coupling between neutronics and thermal-hydraulics. Indeed, a multi-physics approach improves the reactor safety analysis and the design of different types of nuclear systems; in addition, it allows the investigation of physical effects at different scales of time and space. In this context, a challenging task is the development of multi-physics tools to study the fuel burnup: these tools could improve the fuel management and estimate the amount of long-lived radionuclides in spent nuclear fuel for current and innovative nuclear reactors. This paper presents the study of a burnup analysis with the Serpent Monte Carlo code, that implements an external interface for the coupling with OpenFOAM, in order to import material temperatures and density field calculated by a thermal-hydraulics solver. In particular, we carried out a burnup analysis for the entire fuel cycle of a simplified fuel cell, composed by an UO2 pin surrounded by water. We evaluated the effects of the multi-physics coupling by comparing the results from simulations that adopt uniform distributions of material temperatures and densities, to those obtained with the multi-physics coupled approach. Particularly, we will show the differences in nuclide densities and the results from the transport calculation (neutron fluxes, reaction rates and criticality).

AB - In nuclear reactor analysis, a relevant challenge is to achieve a suitable global description of nuclear systems through the coupling between neutronics and thermal-hydraulics. Indeed, a multi-physics approach improves the reactor safety analysis and the design of different types of nuclear systems; in addition, it allows the investigation of physical effects at different scales of time and space. In this context, a challenging task is the development of multi-physics tools to study the fuel burnup: these tools could improve the fuel management and estimate the amount of long-lived radionuclides in spent nuclear fuel for current and innovative nuclear reactors. This paper presents the study of a burnup analysis with the Serpent Monte Carlo code, that implements an external interface for the coupling with OpenFOAM, in order to import material temperatures and density field calculated by a thermal-hydraulics solver. In particular, we carried out a burnup analysis for the entire fuel cycle of a simplified fuel cell, composed by an UO2 pin surrounded by water. We evaluated the effects of the multi-physics coupling by comparing the results from simulations that adopt uniform distributions of material temperatures and densities, to those obtained with the multi-physics coupled approach. Particularly, we will show the differences in nuclide densities and the results from the transport calculation (neutron fluxes, reaction rates and criticality).

KW - Burnup analysis

KW - Monte Carlo

KW - Multi-Physics

KW - OpenFOAM

KW - Serpent

UR - http://www.scopus.com/inward/record.url?scp=85105998107&partnerID=8YFLogxK

M3 - Conference contribution

AN - SCOPUS:85105998107

T3 - International Conference on Physics of Reactors, PHYSOR 2018: Reactor Physics Paving the Way Towards More Efficient Systems

SP - 2115

EP - 2125

BT - International Conference on Physics of Reactors, PHYSOR 2018

PB - Sociedad Nuclear Mexicana, A.C.

T2 - 2018 International Conference on Physics of Reactors: Reactor Physics Paving the Way Towards More Efficient Systems, PHYSOR 2018

Y2 - 22 April 2018 through 26 April 2018

ER -