Skip to content
Licensed Unlicensed Requires Authentication Published by De Gruyter December 8, 2016

Effect of ultra high temperature ceramics as fuel cladding materials on the nuclear reactor performance by SERPENT Monte Carlo code

Untersuchung des Einflusses von Ultra-Hochtemperatur-Keramiken als Material für Brennstoff-Hüllrohre auf die Reaktorleistung mit Hilfe des SERPENT-Monte-Carlo-Codes
  • T. Korkut , A. Kara and H. Korkut
From the journal Kerntechnik


Ultra High Temperature Ceramics (UHTCs) have low density and high melting point. So they are useful materials in the nuclear industry especially reactor core design. Three UHTCs (silicon carbide, vanadium carbide, and zirconium carbide) were evaluated as the nuclear fuel cladding materials. The SERPENT Monte Carlo code was used to model CANDU, PWR, and VVER type reactor core and to calculate burnup parameters. Some changes were observed at the same burnup and neutronic parameters (keff, neutron flux, absorption rate, and fission rate, depletion of U-238, U-238, Xe-135, Sm-149) with the use of these UHTCs. Results were compared to conventional cladding material zircalloy.


Ultra-Hochtemperatur-Keramiken (UHTCs) haben eine niedrige Dichte und einen hohen Schmelzpunkt. Sie sind deshalb bei der Auslegung von Reaktorkernen von Vorteil. Drei UHTCs (Siliciumcarbid, Vanadiumcarbid und Zirkoniumcarbid) wurden als Materialien für Brennstoff-Hüllrohre bewertet. Der SERPENT-Monte-Carlo-Code wurde verwendet, um die Reaktorkerne von CANDU-, PWR- und VVER-Anlagen zu modellieren und die Abbrandparameter zu berechnen. Bei Verwendung der UHTC-Materialien wurden bei gleichem Abbrand und gleichen Neutronenparametern (keff, Neutronfluss, Absorptionsrate and Spaltrate, Abreicherung von U-238, U-238, Xe-135, Sm-149) einige Unterschiede festgestellt. Die Ergebnisse wurden verglichen mit dem konventionallen Hüllrohrmaterial Zirkaloy.

* Corresponding author: E-mail:


1 Leppänen, J.: Development of a new Monte Carlo reactor physics code. VTT Publications 640. D.Sc. Thesis, Helsinki University of Technology, 2007Search in Google Scholar

2 Chersola, D.; Lomonacoa, G.; Marottaa, R.; Mazzinic, G.: Comparison between SERPENT and MONTEBURNS codes applied to burnup calculation of a GFR-like configuration. Nuclear Engineering and Design273 (2014) 54255410.1016/j.nucengdes.2014.03.035Search in Google Scholar

3 Ochoa, R.; Vázquez, M.; Álvarez-Velarde, F.; Martín-Fuertes, F.; García-Herranz, N.; Cuervo, D.: A comparative study of Monte Carlo-coupled depletion codes applied to a Sodium Fast Reactor design loaded with minor actinides. Annals of Nuclear Energy57 (2013) 324010.1016/j.anucene.2013.01.039Search in Google Scholar

4 Rachamin, R.; Wemple, C.; Fridman, E.: Neutronic analysis of SFR core with HELIOS-2, Serpent, and DYN3D codes. Annals of Nuclear Energy55 (2013) 19420410.1016/j.anucene.2012.11.030Search in Google Scholar

5 Leppänen, J.: Performance of Woodcock delta-tracking in lattice physics applications using the Serpent Monte Carlo reactor physics burnup calculation code. Annals of Nuclear Energy37 (2010) 71572210.1016/j.anucene.2010.01.011Search in Google Scholar

6 Aufiero, M.; Cammi, A.; Fiorina, C.; Leppänen, J.; Luzzi, L.; Ricotti, M. E.: An extended version of the SERPENT-2 code to investigate fuel burn-up and core material evolution of the Molten Salt Fast Reactor, Journal of Nuclear Materials441 (2013) 47348610.1016/j.jnucmat.2013.06.026Search in Google Scholar

7 Ghosh, D.; Subhash, G.; Orlovskaya, N.: Measurement of scratch-induced residual stress within SiC grains in ZrB2–SiC composite using micro-Raman spectroscopy. Acta Materialia56 (2008) 5345535410.1016/j.actamat.2008.07.031Search in Google Scholar

8 Glasstone, S.; Sesonske, A.: Nuclear Reactor Engineering. D. Van Nostrand Company, Inc., 08685a30, 1967Search in Google Scholar

9 Agostinelli, S. et al.: Geant4-a simulation toolkit. Nucl. Instrum. Methods A506 (2003) 25030310.1016/S0168-9002(03)01368-8Search in Google Scholar

10 Allison, J. et al.: Geant4 developments and applications. IEEE Trans. Nucl. Sci.53 (2006) 27027810.1109/TNS. 2006.869826Search in Google Scholar

11 Dorval, E.; Leppänen, J.: Monte Carlo current-based diffusion coefficients: Application to few-group constants generation in Serpent. Ann. Nucl. Energy78 (2015) 10411610.1016/j.anucene.2014.12.011Search in Google Scholar

12 Rintala, V.; Suikkanen, H.; Leppänen, J.; Kyrki-Rajamaki, R.: Modeling of realistic pebble bed reactor geometries using the Serpent Monte Carlo code. Ann. Nucl. Energy77 (2015) 22323010.1016/j.anucene.2014.11.018Search in Google Scholar

13 Alferov, V. P.; Radaev, A. I.; Shchurovskaya, M. V.; Tikhomirov, G. V.; Hanan, N. A.; van Heerden, F. A.: Comparative validation of Monte Carlo codes for the conversion of a research reactor. Ann. Nucl. Energy77 (2015) 27328010.1016/j.anucene.2014.11.032Search in Google Scholar

14 Korkmaz, M. E.; Agar, O.; Buyuker, E.: Burnup analysis of the VVER-1000 reactor using thorium-based fuel. Kerntechnik79 (6) (2014) 47848310.3139/124.110449Search in Google Scholar

15 Korkmaz, M. E.; Agar, O.: The Investigation of Burnup Characteristics using the SERPENT Monte Carlo Code for a Sodium Cooled Fast Reactor. Nuclear Enineering and Technology46 (2014) 40741210.5516/NET.07.2013.050Search in Google Scholar

16 Pusa, M.; Leppänen, J.: Computing the Matrix Exponential in Burnup Calculations. Nuclear Science and Engineeing164 (2010) 140150, in Google Scholar

Received: 2015-10-05
Published Online: 2016-12-08
Published in Print: 2016-12-16

© 2016, Carl Hanser Verlag, München

Downloaded on 1.3.2024 from
Scroll to top button