Article Sidebar

PDF
Published: 2016-03-30
Abstract Views: 291
Views PDF: 32
DOI: https://doi.org/10.53747/jnst.v6i1.142
Issue
Vol. 6 No. 1 (2016)
Section
Articles
How to Cite
Nguyenn, K. C., & Vo Doan, H. D. (2016). Conceptual design of critical assembly using Low enriched uranium fuel and moderated light wate. Nuclear Science and Technology, 6(1), 14-31. https://doi.org/10.53747/jnst.v6i1.142

Conceptual design of critical assembly using Low enriched uranium fuel and moderated light wate

Cuong Kien Nguyen1, Dang Hai Doan Vo1
1 Vietnam Atomic Energy Institute, Dalat Nuclear Research Institute, Reactor Center

Main Article Content

Abstract

Critical assembly is a very important facility to serve for fundamental reactor physics research, application of neutron source, training and education. In nuclear engineering, critical assembly is a facility for carrying out measurement of reactor physics parameters, creating benchmark problem, validation of neutron physics calculation tool in computer codes and nuclear data. Basing on concept using commercial Nuclear Power Plant (NPP) fuels such as PWR (AP-1000) and VVER-1000 fuel rods with limited 2 meter in length and fully controlled by water level, the conceptual design of the critical assembly has been carried out in neutronic, thermal hydraulics and safety analysis. Ten benchmark critical core configurations of critical assembly are established and investigated to show safety during normal opeartion and accident conditions. Design calculation results show that NPP fuels are fully adequate for critical assembly operating under nominal power 100W and having average neutron flux about 3×109 neutron/cm2.s.

Article Details

Keywords

Critical assembly, NPP fuels, neutronics, thermal hydraulics, safety analysis

References

[1] Westinghouse Electric Corporation. “The Westinghouse Pressurized Water Reactor Nuclear Power Plant”. Pittburgh Pennsylvania 15230. USA. 1984.
[2] Dan Gabriel Cacuci (Ed.). “Handbook of Nuclear Engineering”. Sringer Science+Business Media. New York USA. 2010.
[3] IAEA-TECDOC-384, “Technology and use of low power research reactors”, Internaltional meeting, IAEA, Beijing, China, 30 April to 3 May 1985.
[4]. IAEA-TECDOC-1234, “The applications of research reactors”, Vienna, Austria, August 2001.
[5] X-5 Monte Carlo Team. “MCNP — A General Monte Carlo N-Particle Transport Code. Version 5”. Los Alamos national laboratory. April 2005.
[6] Keisuke OKUMURA, Teruhico KUGO, Kunio KANEKO and Keichiro TSUCHIHASHI. “SRAC2006: A Comprehensive Neutronics Calculation Systems”. JAEA. Japan. Feb. 2007.
[7] J. R. Deen, W. L. Woodruff, C. I. Costescu, and L. S. Leopando, “WIMS-ANL User Manual.
[8] A. P. Olson, “A Users Guide for the REBUS-PC Code, Version 1.4” ANL/RERTR/TM02-32, December 21, 2001. Rev. 5,” ANL/RERTR/TM-99-07, Argonne National Laboratory, February 2003.
[9] A.G. Croff, “A User Manual for the ORIGEN2 Computer Code”, Oak Ridge NL, 1980.
[10] Nuclear Safety Analysis Division, “RELAP5/MOD3.3 Code Manual”, Idaho National Laboratory, USA, 2001.
[11] Yasunobu NAGAYA, Keisuke OKUMURA, Takamasa MORI and Masayuki NAKAGAWA, “MVP/GMVP II: General Purpose Monte Carlo Codes for Neutron and Photon Transport Calculations based on Continuous Energy and Multigroup Methods”, JAEA, Japan, Sep. 2004.
[12] Keisuke OKUMURA, Yasunobu NAGAYA and Takamasa MORI. “MVP-BURN: Burn-up Calculation Code Using A Continuous-energy Monte Carlo Code MVP”. JAEA. Japan. January. 2005.
[13] Kien Cuong NGUYEN, et.al, “The development depletion code couped with Monte Carlo computer code”, Nuclear Science and Technology Conference XI, Danang, Vietnam.