Skip to content
Licensed Unlicensed Requires Authentication Published by De Gruyter April 15, 2014

Neutron flux investigation on certain alternative fluids in a hybrid system by using MCNPX Monte Carlo transport code

Untersuchung des Neutronenflusses an alternativen Flüssigkeiten mit Hilfe des MCNPX-Monte-Carlo-Transportcodes in einem hybriden System
  • M. Günay
From the journal Kerntechnik

Abstract

In this study, the molten salt-heavy metal mixtures 93 – 85 % Li20Sn80 + 5 % SFG-PuO2 and 2 – 10 % UO2, 93 – 85 % Li20Sn80 + 5 % SFG-PuO2 and 2 – 10 % NpO2, 93 – 85 % Li20Sn80 + 5 % SFG-PuO2 and 2 – 10 % UCO were used as fluids. The fluids were used in the liquid first wall, blanket and shield zones of the designed hybrid reactor system. Four centimeter thick 9Cr2WVTa ferritic steel was used as the structural material. In this study, the effect of mixture components on the neutron flux was investigated in a designed fusion–fission hybrid reactor system. The neutron flux was investigated according to the mixture components, radial flux distribution and energy spectrum in the designed system. Three-dimensional analyses were performed using the most recent MCNPX-2.7.0 Monte Carlo radiation transport code and the ENDF/B-VII.0 nuclear data library.

Kurzfassung

In dieser Studie wurden Flüssigsalz-Schwermetall-Mischungen aus 93 – 85 % Li20Sn80 + 5 % SFG-PuO2 und 2 – 10 % UO2, 93 – 85 % Li20Sn80 + 5 % SFG-PuO2 und 2 – 10 % NpO2, 93 – 85 % Li20Sn80 + 5 % SFG-PuO2 und 2 – 10 % UCO als Flüssigkeiten verwendet. Die Flüssigkeiten wurden an der ersten Flüssigwand, am Blanket- und Shield-Bereich des konzipierten hybriden Reaktorsystems eingesetzt. Als strukturelles Baumaterial wurde vier Zentimeter dicker 9Cr2WVTa ferritischer Stahl verwendet. In dieser Studie wurde untersucht, welche Einflüsse die Mischungsbestandteile auf den Neutronenfluss im konzipierten hybriden Reaktorsystem haben. Ausgehend von den Mischungsbestandteilen, der Verteilung der radialen Strömung und dem Energiespektrum wurde der Neutronenfluss im System berechnet. Dreidimensionale Analysen wurden mit Hilfe der neuesten Version des Monte Carlo Strahlungstransportcodes MCNPX-2.7.0 und der nuklearen Datenbibliothek ENDF/B-VII.0 gemacht.

References

1 Şahin, S.; Übeyli, M.: Radiation Damage Studies on The First Wall of a HYLIFE-II Type Fusion Breeder. Energy Conversion and Management46 (2005) 318510.1016/j.enconman.2005.03.007Search in Google Scholar

2 Şahin, S.; et al.: Effects of spectral shifting in an inertial confinement fusion systemKerntechnik70 (2005) 23310.3139/124.100251Search in Google Scholar

3 Şarer, B.Günay, M.; et al.: Three-Dimensional Neutronic Calculations For The Fusion Breeder Apex ReactorFusion Science and Technology52 (2007) 10710.13182/FST07-A1490Search in Google Scholar

4 Günay, M.; et al.: Three-Dimensional Neutronic Calculations for a Fusion Breeder Apex Reactor Using Some Libraries. Annals of Nuclear Energy38 (12) (2011) 275710.1016/j.anucene.2011.08.007Search in Google Scholar

5 Günay, M.; et al.: Three-dimensional Monte Carlo Calculation of Gas Production in Structural Material of APEX Reactor for Some Evaluated Data Files. Annals of Nuclear Energy55 (2013) 29210.1016/j.anucene.2013.01.001Search in Google Scholar

6 Günay, M.; et al.: The Effect on Radiation Damage of Structural Material in a Hybrid System By Using a Monte Carlo Radiation Transport Code. Annals of Nuclear Energy63 (2013) 15710.1016/j.anucene.2013.07.038Search in Google Scholar

7 Christofilos, N. C.: Design for A High Power-Density Astron Reactor. Journal of Fusion Energy8 (1989) 9710.1007/BF01050784Search in Google Scholar

8 Moir, R. W.: Liquid First Walls For Magnetic Fusion Energy Configurations. Nuclear Fusion37 (1997) 55710.1088/0029-5515/37/4/I13Search in Google Scholar

9 Abdou, M. A.; et al.: Chapter 1: Overview. APEX Interim Report, 1999Search in Google Scholar

10 Abdou, M. A. et al.: On The Exploration of Innovative Concepts for Fusion Chamber Technology. Fusion Engineering and Design54 (2001) 18110.1016/S0920-3796(00)00433-6Search in Google Scholar

11 Abdou, M. A.: Preface. Fusion Engineering and Design72 (2004) 110.1016/j.fusengdes.2004.10.001Search in Google Scholar

12 Abdou, M. A.; et al.: Overview of Fusion Blanket R&D in The US Over The Last Decade. Nuclear Engineering and Technology37 (5) (2005) 401Search in Google Scholar

13 Ying, A.; et al.: Chapter 5: Thick Liquid Blanket Concept, APEX Interim Report, 1999Search in Google Scholar

14 Youssef, M. Z.; Abdou, M. A.: Heat Deposition, Damage and Tritium Breeding Characteristics in Thick Liquid Wall Blanket Concepts. Fusion Engineering and Design49 (2000) 71910.1016/S0920-3796(00)00179-4Search in Google Scholar

15 Youssef, M. Z.; et al.: The Breeding Potential of “Flinabe” and Comparison to “Flibe” in “CLiFF” High Power Density Concept. Fusion Engineering and Design61 (2002) 49710.1016/S0920-3796(02)00245-4Search in Google Scholar

16 IAEA: International Atomic Energy Agency IAEA-TECDOC-1349, 2003Search in Google Scholar

17 Piera, M.; et al.: Hybrid Reactors: Nuclear Breeding or Energy Production?. Energy Conversion and Management51 (2010) 175810.1016/j.enconman.2010.01.025Search in Google Scholar

18 Şarer, B.; et al.: Calculations of Neutron-Induced Production Cross-Sections of 180,182,183,184,186W up to 20 MeV. Annals of Nuclear Energy36 (2009) 41710.1016/j.anucene.2008.11.025Search in Google Scholar

19 Günay, M.: Investigation of radiation damage in structural material of APEX reactor by using Monte Carlo method. Annals of Nuclear Energy53 (2013) 5910.1016/j.anucene.2012.06.038Search in Google Scholar

20 Chadwick, M. B.; et al.: ENDF/B-VII.0: Next Generation Evaluated Nuclear Data Library for Nuclear Science and Technology. Nuclear Data Sheets, 107 (2006) 293110.1016/j.nds.2006.11.001Search in Google Scholar

21 Pelowitz, D. B.: MCNPX User's Manual, Version 2.7.0, LA-CP-11-00438, 2011Search in Google Scholar

22 Günay, M.: Investigation of Neutronic Effects in Structural Material of a Hybrid Reactor by using The MCNPX Monte Carlo Transport Code. Kerntechnik 78:3 (2013) 198Search in Google Scholar

23 Günay, M.; Kasap, H.: Neutronic Investigation of The Application of Certain Plutonium-Mixed Fluids in a Fusion-Fission Hybrid Reactor. Annals of Nuclear Energy63 (2014) 43210.1016/j.anucene.2013.08.024Search in Google Scholar

24 Şarer, B.; et al.: Comparisons of The Calculations Using Different Codes Implemented in MCNPX Monte Carlo Transport Code for Accelerator Driven System Target. Fusion Science Technology61 (2012) 30210.13182/FST12-A13437Search in Google Scholar

Received: 2013-12-24
Published Online: 2014-04-15
Published in Print: 2014-04-28

© 2014, Carl Hanser Verlag, München

Downloaded on 28.3.2024 from https://www.degruyter.com/document/doi/10.3139/124.110408/html
Scroll to top button