TY - GEN

T1 - Cross-section uncertainty propagation and adjustment algorithms for BWR core simulation

AU - Jessee, Matthew A.

AU - Turinsky, Paul J.

AU - Abdel-Khalik, Hany S.

PY - 2009

Y1 - 2009

N2 - Computational capability has been developed to adjust multi-group neutron cross-sections to improve the fidelity of boiling water reactor (BWR) core modeling and simulation. The method involves propagating multi-group neutron cross-section uncertainties through various BWR computational models, i.e., multi-group generation codes, resonance self-shielding treatment, lattice physics calculations, and core calculations, to evaluate uncertainties in key core attributes such as core k-effective, nodal power distributions, thermal margins, and in-core detector readings. Uncertainty-based inverse theory methods are then employed to adjust multi-group cross-sections to minimize the disagreement between BWR modeling predictions and measured plant data. For this paper, measured plant data are virtually simulated in the form of perturbed 3-D nodal power distributions with discrepancies with predictions of the same order of magnitude as expected from plant data. The major focus of this work is to efficiently propagate multi-group neutron cross-section uncertainty through BWR lattice physics calculations. For energy groups in the resolved resonance energy range, multi-group cross-section uncertainties are computed by multi-group generation codes using an infinitely-dilute approximation of the neutron flux. In order to accurately account for spatial and energy resonance self-shielding effects in unit cell calculations, the multi-group cross-section covariance matrix is reformulated to include the uncertainty in resonance correction factors, or self-shielding factors, which are used to calculate the self-shielded multi-group cross-sections used in the lattice physics calculation.

AB - Computational capability has been developed to adjust multi-group neutron cross-sections to improve the fidelity of boiling water reactor (BWR) core modeling and simulation. The method involves propagating multi-group neutron cross-section uncertainties through various BWR computational models, i.e., multi-group generation codes, resonance self-shielding treatment, lattice physics calculations, and core calculations, to evaluate uncertainties in key core attributes such as core k-effective, nodal power distributions, thermal margins, and in-core detector readings. Uncertainty-based inverse theory methods are then employed to adjust multi-group cross-sections to minimize the disagreement between BWR modeling predictions and measured plant data. For this paper, measured plant data are virtually simulated in the form of perturbed 3-D nodal power distributions with discrepancies with predictions of the same order of magnitude as expected from plant data. The major focus of this work is to efficiently propagate multi-group neutron cross-section uncertainty through BWR lattice physics calculations. For energy groups in the resolved resonance energy range, multi-group cross-section uncertainties are computed by multi-group generation codes using an infinitely-dilute approximation of the neutron flux. In order to accurately account for spatial and energy resonance self-shielding effects in unit cell calculations, the multi-group cross-section covariance matrix is reformulated to include the uncertainty in resonance correction factors, or self-shielding factors, which are used to calculate the self-shielded multi-group cross-sections used in the lattice physics calculation.

KW - BWR simulation

KW - Data adjustment

KW - Uncertainty analysis

UR - http://www.scopus.com/inward/record.url?scp=74549199474&partnerID=8YFLogxK

M3 - Conference contribution

AN - SCOPUS:74549199474

SN - 9781615673513

T3 - American Nuclear Society - 4th Topical Meeting on Advances in Nuclear Fuel Management 2009, ANFM IV

SP - 1357

EP - 1376

BT - American Nuclear Society - 4th Topical Meeting on Advances in Nuclear Fuel Management 2009, ANFM IV

T2 - American Nuclear Society - 4th Topical Meeting on Advances in Nuclear Fuel Management 2009, ANFM IV

Y2 - 12 April 2009 through 15 April 2009

ER -