Modification and Validation of ATHLET Code for Sodium-cooled Fast Reactor Application

ZHOU Chong; Klaus Huber; CHENG Xu

doi:10.7538/yzk.2013.47.11.2053

Atomic Energy Science and Technology > 2013 > 47(11): 2053-2058. > DOI: 10.7538/yzk.2013.47.11.2053

ZHOU Chong, Klaus Huber, CHENG Xu. Modification and Validation of ATHLET Code for Sodium-cooled Fast Reactor Application[J]. Atomic Energy Science and Technology, 2013, 47(11): 2053-2058. DOI: 10.7538/yzk.2013.47.11.2053

Citation:

PDF (683 KB)

Modification and Validation of ATHLET Code for Sodium-cooled Fast Reactor Application

1.
School of Nuclear Science and Engineering, Shanghai Jiao Tong University, Shanghai 200240, China
2.
Institute of Fusion and Reactor Technology, Karlsruhe Institute of Technology, Karlsruhe 76131, Germany

More Information

Graphical Abstract

Abstract

Abstract

System analysis code is important for the global simulation of the sodium-cooled fast reactor (SFR) system as well as transient and accident safety analysis. In this paper, the best estimate system code ATHLET for light water reactors, developed by Gesellschaft für Anlagen-und Reaktorsicherheit (GRS) in Germany, was modified for SFR application. Thermal-dynamic and transport properties as well as heat transfer correlations for sodium were implemented into the ATHLET code. The modified code was then applied to simulate the Phenix reactor in France, and validation of the code was conducted with the Phenix reactor natural convection test. The calculation results were compared with the test data. The results show that the modified ATHLET code has good applicability in simulating SFR systems.
- sodium-cooled fast reactor,
- ATHLET code,
- natural convection

FullText(HTML)

References (9)

References

[1]	LINEBERRY M J, ALLEN T R. The sodium-cooled fast reactor (SFR)[C]∥Proceedings of the 2002 Americas Nuclear Energy Symposium. Miami, US: [s. n.], 2002.
[2]	IAEA. Status of liquid metal cooled fast reactor technology, IAEA-TECDOC-1083[R]. Austria: IAEA, 1999.
[3]	TENCHINE D. Some thermal hydraulic challenges in sodium cooled fast reactors[J]. Nucl Eng Des, 2010, 240(5): 1195-1217.
[4]	LERCHEL G, AUSTREGESILO H. ATHLET mod 2.1 cycle A, user’s manual, GRS-p-1/Vol. 1, Rev.4[M]. Germany: GRS, 2006.
[5]	FINK J K, LEIBOWITZ L. Thermodynamic and transport properties of sodium liquid and vapor, ANL/RE95/2[R]. USA: Argonne National Laboratory, 1995.
[6]	CHENG X, TAK N. Investigation on turbulent heat transfer to lead-bismuth eutectic flows in circular tubes for nuclear applications[J]. Nucl Eng Des, 2006, 236(4): 385-393.
[7]	MIKITYUK K. Heat transfer to liquid metal: Review of data and correlations for tube bundles[J]. Nucl Eng Des, 2009, 239(4): 680-687.
[8]	PIALLA D, TENCHINE D, GAUTH P, et al. Natural convection test in Phenix reactor and associated CATHARE calculation[C]∥Proceedings of the 14th International Topical Meeting on Nuclear Reactor Thermal Hydraulics. Canada: [s. n.], 2011.
[9]	许义军. 中国实验快堆钠池三维热工水力分析[D]. 北京：中国原子能科学研究院，2003.

Cited By

Get Citation

PDF

XML

Article views (218) PDF downloads (1057)

Turn off MathJax

Article Contents

Abstract

References

Modification and Validation of ATHLET Code for Sodium-cooled Fast Reactor Application

Abstract

References

Catalog

Export File

Citation

Format

Content