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Published: 2011-06-30
Abstract Views: 53
DOI: https://doi.org/10.53747/nst.v1i2.298
Issue
Vol. 1 No. 2 (2011)
Section
Articles
How to Cite
LUONG , B. V., LE , V. V., HUYNH , T. N., & NGUYEN , K. C. (2011). SAFETY AND TRANSIENT ANALYSES FOR FULL CORE CONVERSION OF THE DALAT NUCLEAR RESEARCH REACTOR. Nuclear Science and Technology, 1(2), 1-15. https://doi.org/10.53747/nst.v1i2.298

SAFETY AND TRANSIENT ANALYSES FOR FULL CORE CONVERSION OF THE DALAT NUCLEAR RESEARCH REACTOR

LUONG BA VIEN1, LE VINH VINH1, HUYNH TON NGHIEM1, NGUYEN KIEN CUONG1
1 Reactor Center, Nuclear Research Institute, Vietnam Atomic Energy Institute 01 Nguyen Tu Luc St, Dalat, Vietnam

Main Article Content

Abstract

Preparing for full core conversion of Dalat Nuclear Research Reactor, safety and transient analyses were carried out to confirm the ability to operate safely of proposed Low Enriched Uranium (LEU) working core. The initial LEU core consisting of 92 LEU fuel assemblies and 12 Beryllium rods was analyzed under initiating events of uncontrolled mwithdrawal of a control rod, cooling pump failure, earthquake and fuel cladding fail. Working LEU core’s response was evaluated under these initial events based on RELAP/MOD 3.2 computer code and other supported codes like ORIGEN, MCNP. Obtained results showed that safety of the reactor is maintained for all transients/accidents analyzed.

Article Details

Keywords

HEU, LEU, RELAP, RIA, LOFA, ORIGEN, MCNP

References

[1] Safety Analysis Report for the Dalat Nuclear Research Reactor, NRI, 2009.
[2] Technical Design for Reconstruction and Enlargement of the Dalat Nuclear Research Reactor - Volume 3. State Design Institute, USSR State Committee for the Utilization of Atomic Energy, Moscow, 1979 (in Russian).
[3] Additional Physics and Thermal-Hydraulic Data for Reactor IVV-9, State Design Institute, USSR State Committee for the Utilization of Atomic Energy, Moscow, 1980 (in Russian).
[4] Report on Neutronics and Thermal-hydraulics Analysis for Full LEU Core Configurations, NRI, 2009.
[5] Huynh Ton Nghiem and Le Vinh Vinh, Preliminary Results of Pre-feasibility Study for Full LEU Core Conversion of the DNRR, 6th National Conference on Nuclear Physics and Technology, Dalat, 2005 (in Vietnamese).
[6] K.A. Konoplev, R.G. Pikulin, A.S. Zakharov, L.V. Tedoradze, G.V. Paneva, D.V. Tchmshkyan, LEU WWR-M2 Fuel Qualification, Proceedings of the 2002 International Meeting on Reduced Enrichment for Research and Test Reactors, Bariloche, Argentina, 3-8 November 2002, ANL/TD/TM03-04, p.117.
[7] G.A. Kirsanov, K.A. Konoplev, R.G. Pikulik, Yu.P. Sajkov, D.V. Tchmshkyan, L.V. Tedoradze, and A.S. Zakharov, LEU WWR-M2 Fuel Assemblies Burnable Test, Proceedings of the 2000 International Meeting on Reduced Enrichment for Research and Test Reactors, Las Vegas, Nevada, USA, 1-6 October 2000.
[8] V. Chernyshov, V. Aden, E. Kartashov, V. Lukichev, and A. Enin, Russian RERTR Program Works status, Proceedings of the 2002 International Meeting on Reduced Enrichment for Research and Test Reactors, Bariloche, Argentina, 3-8 November 2002, ANL/TD/TM03-04, p.109.
[9] VVR-M2 and VVR-M5 Fuel Assemblies-Operation Manual, 0001.04.00.000 PЭ, 2006.
[10] Report on the 6th Core Loading for the Dalat Nuclear Research Reactor, NRI, Dalat, 2007
(in Vietnamese).
[11] Report on the 7th Core Loading for the Dalat Nuclear Research Reactor, NRI, Dalat, 2009
(in Vietnamese).
[12] RELAP5/MOD3 Code Manual, SCIENTECH, Inc. Rockville, Maryland, 1999.
[13] W. L. Woodruff, et. al., A comparison of the PARET/ANL and RELAP5/MOD3 Codes for
the Analysis of IAEA Benchmark Transients and the SPERT Experiments, RERTR Program, ANL.
[14] V. V. Le and T. N. Huynh, Application of RELAP5/MOD3.2 for the DNRR, Proceedings
of JAEA Conf. 2006-001.
[15] V. V. Le, T. N. Huynh, B. V. Luong, V. L. Pham, J. R. Liaw, J. Matos, Comparative Analyses for Loading LEU Instead of HEU Fuel Assemblies in the DNRR, RERTR Int’l Meeting, Boston, 2005.
[16] M. M. Shah, Improved General Correlation for Critical Heat Flux during Upflow in Uniformly Heated Vertical Tubes, International Journal of Heat and Fluid Flow, Vol. 8, No. 4, pp. 326-335 (1987)