TY - GEN
T1 - Improved diffusion coefficents for SPN axial solvers in the MPACT 2D/1D method applied to the AP1000® PWR start-up core models
AU - Stimpson, Shane
AU - Franceschini, Fausto
AU - Collins, Benjamin
AU - Godfrey, Andrew
AU - Kim, Kang Seog
AU - Graham, Aaron
AU - Downar, Thomas
N1 - Publisher Copyright:
© Copyright (2015) by the American Nuclear Society.
PY - 2015
Y1 - 2015
N2 - As part of the Virtual Environment for Reactor Applications Core Simulator (VERA-CS), the 2D/1D capability in the MPACT code is being developed collaboratively by Oak Ridge National Laboratory and the University of Michigan. MPACT was used to model the AP1000® reactor start-up cores. One of the major shortcomings observed in initial results was the ability to accurately resolve the pin-wise fission rate distributions for cases with partial-length burnable poison pins. The primary source of the errors was determined to be the diffusion coefficients that are used in the axial transport solvers of the 2D/1D scheme. The work here demonstrates the deficiency of the previous method used to determine the diffusion coefficients by employing the out-scatter approximation to calculate the transport cross sections. New results are obtained by using the in-scatter and what is being termed the "Neutron Leakage Conservation" approximations to more accurately determine the transport cross sections being used to construct the diffusion coefficients. Additionally, the methods were applied to both 3D assembly and quarter core problems with comparisons of the Nodal Expansion Method and Simplified PN axial transport solvers with these improved diffusion coefficients. Significant improvements are observed, particularly in the single-assembly cases with partial-length Wet Annular Burnable Absorber pins. In the quarter-core cases, the improvements are less apparent because the power distribution is flatter, though the results demonstrate that it is still very worthwhile to incorporate these approaches for determining the transport cross sections.
AB - As part of the Virtual Environment for Reactor Applications Core Simulator (VERA-CS), the 2D/1D capability in the MPACT code is being developed collaboratively by Oak Ridge National Laboratory and the University of Michigan. MPACT was used to model the AP1000® reactor start-up cores. One of the major shortcomings observed in initial results was the ability to accurately resolve the pin-wise fission rate distributions for cases with partial-length burnable poison pins. The primary source of the errors was determined to be the diffusion coefficients that are used in the axial transport solvers of the 2D/1D scheme. The work here demonstrates the deficiency of the previous method used to determine the diffusion coefficients by employing the out-scatter approximation to calculate the transport cross sections. New results are obtained by using the in-scatter and what is being termed the "Neutron Leakage Conservation" approximations to more accurately determine the transport cross sections being used to construct the diffusion coefficients. Additionally, the methods were applied to both 3D assembly and quarter core problems with comparisons of the Nodal Expansion Method and Simplified PN axial transport solvers with these improved diffusion coefficients. Significant improvements are observed, particularly in the single-assembly cases with partial-length Wet Annular Burnable Absorber pins. In the quarter-core cases, the improvements are less apparent because the power distribution is flatter, though the results demonstrate that it is still very worthwhile to incorporate these approaches for determining the transport cross sections.
KW - 2D/1D
KW - AP1000
KW - Axial transport
KW - Diffusion coefficient
KW - MPACT
UR - http://www.scopus.com/inward/record.url?scp=84949507232&partnerID=8YFLogxK
M3 - Conference contribution
AN - SCOPUS:84949507232
T3 - Mathematics and Computations, Supercomputing in Nuclear Applications and Monte Carlo International Conference, M and C+SNA+MC 2015
SP - 1702
EP - 1716
BT - Mathematics and Computations, Supercomputing in Nuclear Applications and Monte Carlo International Conference, M and C+SNA+MC 2015
PB - American Nuclear Society
T2 - Mathematics and Computations, Supercomputing in Nuclear Applications and Monte Carlo International Conference, M and C+SNA+MC 2015
Y2 - 19 April 2015 through 23 April 2015
ER -