Program - Brookhaven National Laboratory

Documentation of Uncertainties in Experimental Neutron Cross Sections for EXFOR
N. Otuka, Nuclear Data Section, International Atomic Energy Agency, A-1400 Wien, Austria. D.L.
Smith, Argonne National Laboratory, 1710 Avenida Del Mundo #1506, Coronado, CA 92118, USA.
Detailed documentation of uncertainties in neutron-induced reaction cross sections has been urged by
nuclear data evaluators who want to incorporate experimental information of cross sections and their uncertainties
properly. Though this request has been discussed between nuclear data centres in the past,
there are few EXFOR entries which satisfy the data evaluators, regardless of the existence of some text
books of error analysis specialized for nuclear data [1,2]. Following a number of recommendations from the
IAEA Technical Meeting on Neutron Cross Section Covariances [3], we have prepared practical guidance
on experimental neutron cross section uncertainties with various examples and also extended the EXFOR
format to accommodate a wide range of uncertainty information, in order to encourage experimentalists
to provide a full description of partial and total uncertainties to data evaluators through the EXFOR
library. The paper will review the flow of data from experiment to users, and present several new options
of the EXFOR formats for uncertainties and covariances with sample EXFOR entries. A full report of the
guidance is also under preparation [4].
[1] D.L. Smith, Probability, Statistics, and Data Uncertainties in Nuclear Science and Technology, American
Nuclear Society, LaGrange Park, IL, USA, 1991. [2] W. Mannhart, “A Small Guide to Generating Covariances
of Experimental Data”, Report INDC(NDS)-0588, IAEA Nuclear Data Section, 2011. [3] A. Trkov,
D.L. Smith, R. Capote Noy (ed.), “Technical Meeting on Neutron Cross Section Covariances”, Report
INDC(NDS)-0582, 2011. [4] D.L. Smith, N. Otuka, “Experimental Nuclear Reaction Data Uncertainties:
Basic Concepts and Documentation”, to be published in Nucl. Data Sheets.
PR 131
Application of GRS Method to Evaluation of Uncertainties of Calculation Parameters of
Perspective Sodium-Cooled Fast Reactor
Anton Peregudov, Olga Andrianova, Kirill Raskach, Anatoly Tsibulya, Institute for Physics and Power
Engineering.
A number of recent studies have been devoted to the estimation of errors of reactor calculation parameters
by the GRS (Generation Random Sampled) method. This method is based on direct sampling input data
resulting in formation of random sets of input parameters which are used for multiple calculations. Once
these calculations are performed, statistical processing of the calculation results is carried out to determine
the mean value and the variance of each calculation parameter of interest. In our study this method is
used for estimation of errors of calculation parameters (Keff, power density, dose rate) of a perspective
sodium-cooled fast reactor. Neutron transport calculations were performed by the nodal diffusion code
TRIGEX and Monte Carlo code. Key Words: GRS-method, uncertainty analysis, covariance matrix, Keff,
power density, dose rate, TRIGEX, MMK.
1.B.T. Rearden, M.L. Williams et al. “Sensitivity and Uncertainty Analysis Capabilities and Data in
SCALE”, Nuclear Technology, Vol. 174, No. 2, 2011. 2.W. Zwermann et al. “Uncertainty Analyses with
Nuclear Covariance Data in Reactor Core Calculations.” Int. Conf. on Nuclear Data for Science and
Technology 2010. Jeju Island, Korea, April 26-30, 2010. 3.A.S. Seregin, T.S. Kislitsina, A.M. Tsiboulia.
TRIGEX.04 Code Package. Preprint IPPE-2846, Institute for Physics and Power Engineering, Obninsk
(2000). (In Russian.) 4.A.A. Blyskavka, G.N. Manturov, M.N. Nikolaev, A.M. Tsiboulia. “Multigroup
Monte Carlo Code MMK-KENO,” Preprint IPPE-3145, Institute for Physics and Power Engineering, Obninsk
(2009). (In Russian.) 5.“Multigroup Constant Set for Calculation of Neutron and Photon Radiation
Fields and Functionals, Including the CONSYST2 Program,” ORNL. RSICC DLC-182 (1995). 6.G.N.
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