Program - Brookhaven National Laboratory

[1] D. Jacquet and M. Morjean, Prog. Part. and Nucl. Phys. 63 (2009) 155 [2] P. Grangé, L. Jun-Qing
and H.A. Weidenmuller, Phys. Rev. C 27 (1983) 2063
Session RC Covariances
Friday March 8, 2013
Room: Empire East at 10:30 AM
RC 1 10:30 AM
Applications of Nuclear Data Covariances to Criticality Safety and Spent Fuel
Characterization
M. L. Williams, G. Ilas, W. J. Marshall, B. T. Rearden
Oak Ridge National Lab
During the last several years much advancement has been made in providing nuclear data covariances
for nuclear technology applications. These advancements include new covariance evaluations released in
ENDF/B-VII.1 [1] for many important nuclides, as well as more approximate “low fidelity” uncertainty
data [2] for nuclides without covariance evaluations in ENDF/B nuclear data files. Using these resources,
a comprehensive library of covariance data [3] has been developed for the SCALE code system. [4] This
covariance library – along with the sensitivity/uncertainty (S/U) computation methods in SCALE – allow
realistic uncertainty analysis of diverse types of applications. [5] Previously, most uncertainty analysis with
SCALE has focused on determining uncertainties in the multiplication factor (keff) for criticality safety
applications. The SCALE S/U methodology was recently used in a comprehensive study to compute
eigenvalue uncertainties for a large number of critical experiments. [6] In this paper we compare results
obtained from uncertainty analysis using the SCALE nuclear data covariances, and the actual distribution
of the computed benchmark multiplication factors. SCALE uncertainty analysis for critical eigenvalues is
based on first order perturbation, which is a very efficient technique for determining the uncertainty in a
single response (keff) for a fixed invariant system. At present this perturbation approach cannot be applied
to burnup calculations where a large number of responses are of interest, and where the composition of
the system (i.e., reactor) varies as a function of time due to non-linear interactions between the neutron
flux and nuclide fields. A new method based on statistical sampling [7] has recently been developed for
SCALE. When combined with a comprehensive covariance library, the statistical sampling method allows
uncertainty analysis to be performed for reactor burnup calculations to obtain uncertainties in spent
fuel responses, which are important for burned fuel transportation, on-site storage, and disposition. [8]
Computed uncertainties are presented for time-dependent responses such as actinide and fission isotopics,
decay heat, and decay activity, which depend on uncertainties in the nuclear data of many fission products
and actinides.
[1] M. B. Chadwick, et al “ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections,
Covariances, Fission Product Yields and Decay Data,” Nuclear Data Sheets 112, 12, 2887-2996 (December
2011). [2] R. C. Little, et al., “Low-fidelity Covariance Project” Nuclear Data Sheets 109, 2828 (2008).
[3] M. L. Williams and B. T. Rearden “Sensitivity/Uncertainty Analysis Capabilities and New Covariance
Data Libraries in SCALE” Nuclear Data Sheets 109, 12, 2796-2800 (December 2008). [4] “SCALE: A
Comprehensive Modeling and Simulation Suite for Nuclear Safety Analysis and Design,” ORNL/TM-
2005/39, Version 6.1, Oak Ridge National Laboratory, Oak Ridge, Tennessee (June 2011). [5] B. T.
Rearden, M. L. Williams, M. A. Jessee, D. E. Mueller, D. A. Wiarda. “Sensitivity and Uncertainty
Analysis Capabilities in SCALE” Nuclear Technology 174, 236-288 (May 2011) [6] W. J. Marshall and
B. T. Rearden, Criticality Safety Validation of SCALE 6.1, ORNL/TM-2011/450, Oak Ridge National
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