Program - Brookhaven National Laboratory

NC 3 11:20 AM
Propagation of Nuclear Data Uncertainties for ELECTRA Burn-Up Calculations
Henrik Sjöstrand, Erwin Alhassan, Junfeng Duan, Cecilia Gustavsson, Michael Osterlund, Stephan Pomp
Division of Applied Nuclear Physics, Department of Physics and Astronomy, Uppsala University, Box
516, 751 20, Uppsala, Sweden
Arjan Koning, Dimitri Rochman
Nuclear Research and Consultancy Group (NRG),P.O. Box 25,3 Westerduinweg, 1755 ZG Petten, The
Netherlands
The European Lead-Cooled Training Reactor (ELECTRA) [1] has been proposed as a training reactor for
fast systems within the Swedish nuclear program. It is a low-power fast reactor cooled by liquid pure lead
and includes many attractive safety features such as negative coolant and fuel temperature coefficients.
It is important that these features are retained during the full lifetime of the core. In this work we
investigate the uncertainty in the fuel inventory of ELECTRA during the reactor life and how this affects
the safety of the reactor. For this we use the Total Monte Carlo approach [2] for burn-up calculations on
the ELECTRA design. I.e., a reaction model code system, in our case a TALYS based [3], is used. The
nuclear models input parameters are randomized within their uncertainties and a large set of nuclear data
libraries are generated. The libraries are used as inputs to reactor codes, in our case SERPENT [4], to
perform uncertainty analysis of nuclear reactor inventory during burn-up. The uncertainty in the inventory
during the life of the reactor, combined with the uncertainty in the cross-section, is subsequently used to
assess the uncertainties in the safety parameters.
[1] Wallenius, J., Suvdantsetseg, E., and Fokau, A., 2012. ELECTRA: European Lead-Cooled Training
Reactor. Fission Reactors. Nuclear Technology 177, p. 303-313. [2] Koning, A.J., and Rochman, D.,
2008. Towards sustainable nuclear energy: Putting nuclear physics to work. Annals of Nuclear Energy
35, p. 2024-230. [3] Rochman, D., and Koning, A.J., 2011. How to randomly evaluate nuclear data: A
new data adjustment method applied to Pu-239. Nuclear Science and Engineering 168(1), p. 68-80 [4] J.
Leppanen, 2010. Psg2 / serpent - a Continuous-energy Monte Carlo Reactor Physics Burnup Calculation
Code, Technical report, VTT Technical Research Centre of Finland, Finland, http://montecarlo.vtt.fi.
NC 4 11:40 AM
Total Monte-Carlo Method Applied to a Pressurized Water Reactor Fuel Assembly -
Quantification of Uncertainties due to 235,238 U, 239,240,241 Pu and Fission Products Nuclear
Data Uncertainties
D.F. da Cruz, D. Rochman and A.J. Koning
Nuclear Research and Consultancy Group NRG
The Total Monte-Carlo (or TMC) method has been applied to a single pressurized water reactor fuel
assembly to quantify the uncertainties in reactivity and discharged fuel inventory as a result of uncertainties
in nuclear data. Different from the perturbation-theory-based methods applied for decades, TMC relies
on the higher computational power available nowadays to propagate uncertainties from nuclear data to
reactor physics parameters. TMC involves a large number of calculations for the same model, performed
with different nuclear data in each of them. This paper discusses the uncertainty analysis on reactivity and
inventory for a Westinghouse 3-loop fuel element as a result of uncertainties in nuclear data of the actinides
235,238 U and 239,240,241 Pu, and of a set of the 119 most important fission products. A fuel assembly fuelled
with UO2 fuel with 4.8% enrichment has been selected. TMC has been applied using the deterministic
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