Depletion uncertainties propagation for replica calculations in the lead cooled fast reactor

• Autorzy: Stanisz P. , Oettingen M. , Kępisty G.
• Języki publikacji: EN

Abstrakty

EN

Most Monte Carlo codes are used to determine certain values with their uncertainty accompanying through stochastic process. Those estimations are crucial information to determine the logistics of frontend and the back-end of nuclear chain supply management. Monte Carlo method simulate physics interactions, where correct results can be obtained if users is running a sufficient number of neutron histories adequately to sample all significant regions of the problem. The code by using internal random walks of neutrons is able to estimate a nuclear parameter k-eff (multiplication factor) and fission source distribution responsible for the ratio of new neutrons generation in the following step. Each neutron generation converges to the fix distribution, which can be characterized by Shannon entropy. Tallies of k-eff and spatial reaction rates starts accumulated information after adjusted cut-off step. However, convergence can stop at some level causing neutron distribution tilt and introducing influence to the reaction rate. Locally slightly different power distribution can occurs resulting in slightly different density evolution of the isotopes. In this paper we apply technics of multi “independent replicas” calculations. The ide based on many simulations of the same system using different random sequences to obtain slightly various solutions which will allows us to build any probability density function. Statistical analysis of the results would allow assessing the uncertainties in the calculated isotopes densities. In this work we examine multi recycle scheme in the fast neutron spectrum based on The Lead-cooled Fast Reactor (LFR) defined and studied at the level of technical design in order to demonstrate its propagation of isotopes evolutions together with uncertainties and highlight systematic errors, due to the number of simulated particles. All simulated aspect has to be considered while performing Monte Carlo burnup simulations.

Słowa kluczowe

EN

uncertainties propagation; lead fast reactor; burnup uncertainties; closed nuclear fuel cycle; Monte Carlo replica calculations

• Wydawca: Sieć Badawcza Łukasiewicz - Poznański Instytut Technologiczny
• Czasopismo: Logistyka
• Rocznik: 2015
• Tom: nr 4
• Strony: 9794--9804, CD3
• Opis fizyczny: Bibliogr. 10 poz., rys., wykr., tab.

Twórcy

autor

Stanisz P.

• AGH University of Science and Technology, Faculty of Energy and Fuels, Department of Nuclear Energy, Krakow, Poland

autor

Oettingen M.

• AGH University of Science and Technology, Faculty of Energy and Fuels, Department of Nuclear Energy, Krakow, Poland

autor

Kępisty G.

• AGH University

Bibliografia

• [1] T. Ueki and F. B. Brown, Stationarity Diagnostics Using Shannon Entropy in Monte Carlo Criticality Calculation I: F Test, Trans. Am. Nuc, 87, 156 (2002), and Los Alamos National
• [2] Laboratory report LA-UR-02-3783 (2002).
• [3] J. Cetnar, W. Gudowski and J. Wallenius. MCB A continuous energy Monte Carlo Burnup simulation code. In Actinide and Fission Product Partitioning and Transmutation, EUR8898 EN, OCDE/NEA (1999) 523.
• [4] G. Kępisty, J. Cetnar Instabilities of Monte-Carlo burnup calculations for nuclear reactors—Demonstration and dependence from time step model Nuclear Engineering and Design 2015, Nuclear Engineering and Design
• [5] Cinotti, L. et al. (2007). The potential of the LFR and the ELSY project. Nice, France: ICAPP’07.
• [6] J. Cetnar, P. Stanisz, G. Domańska Adiabatic fuel cycle assessment of LFR core with MOX using MCB system, Study for the 7th FP LEADER project, AGH WEiP KEJ/2013/4
• [7] J. Cetnar, P. Stanisz, G. Domańska Transition to the Adiabatic-LFR: preliminary definition of the start-up core and MA-burning capabilities evaluation, Study for the 7th FP LEADER project, AGH WEiP KEJ/2013/5
• [8] J. Cetnar Development and applications of MCB Monte CarloContiunous Energy Burnup Code WFiIS AGH Kraków (2006) ISBN 83-921064-4-XSmith
• [9] NEA/NSC/DOC(2006)18. Processing of the JEFF-3.1 Cross Section Library into Continuous Energy Monte Carlo Radiation Transport and Criticality Data Library. http://www.nea.fr/abs/html/nea-1768.html, 2006
• [10] J. Cetnar General solution of Bateman equations for nuclear transmutations Annals of Nuclear Energy Volume: 33, Issue: 7, May, 2006, pp. 640-645