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Journal
2015 | 60 | 2 | 367-371
Article title

Neutronic analysis for core conversion (HEU–LEU) of the low power research reactor using the MCNP4C code

Content
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Languages of publication
EN
Abstracts
EN
Comparative studies for conversion of the fuel from HEU to LEU in the miniature neutron source reactor (MNSR) have been performed using the MCNP4C code. The HEU fuel (UAl4-Al, 90% enriched with Al clad) and LEU (UO2 12.6% enriched with zircaloy-4 alloy clad) cores have been analyzed in this study. The existing HEU core of MNSR was analyzed to validate the neutronic model of reactor, while the LEU core was studied to prove the possibility of fuel conversion of the existing HEU core. The proposed LEU core contained the same number of fuel pins as the HEU core. All other structure materials and dimensions of HEU and LEU cores were the same except the increase in the radius of control rod material from 0.195 to 0.205 cm and keeping the outer diameter of the control rod unchanged in the LEU core. The effective multiplication factor (keff), excess reactivity (ρex), control rod worth (CRW), shutdown margin (SDM), safety reactivity factor (SRF), delayed neutron fraction (βeff) and the neutron fluxes in the irradiation tubes for the existing and the potential LEU fuel were investigated. The results showed that the safety parameters and the neutron fluxes in the irradiation tubes of the LEU fuels were in good agreements with the HEU results. Therefore, the LEU fuel was validated to be a suitable choice for fuel conversion of the MNSR in the future.
Publisher
Journal
Year
Volume
60
Issue
2
Pages
367-371
Physical description
Dates
published
1 - 6 - 2015
received
16 - 9 - 2014
online
22 - 6 - 2015
accepted
6 - 3 - 2015
References
  • 1. CIAE. (1993). Safety Analysis Report (SAR) for the Syrian Miniature Neutron Source Reactor. China.
  • 2. Khamis, I., & Khattab, K. (1999). Lowering the enrichment of the Syrian Miniature Neutron Source Reactor. Ann. Nucl. Energy, 26, 1031–1036.[Crossref]
  • 3. Matos, J., & Lell, R. (2005). Feasibility study on potential LEU fuels for a generic MNSR reactor. In International Meeting on Reduced Enrichment for Research and Test Reactors, November 6–10, 2005. Boston, Massachusetts.
  • 4. Briesmeister, F. J. (2000). MCNP4C manual, Monte Carlo N-Particle Transport Code System. (RSICC code package CCC-700/MCNP4C). Oak Ridge National Laboratory, TN, and DOE, USA.
  • 5. Jonah, S. A., Liaw, J. R., & Matos, J. E. (2007). Monte Carlo simulation of core physics parameters of the Nigeria Research Reactor-1 (NIRR-1). Ann. Nucl. Energy, 34, 953–957.[Crossref][WoS]
  • 6. Khattab, K., & Sulieman, I. (2009). Calculations of the thermal and fast neutron fluxes in the Syrian MNSR irradiation tubes using the MCNP-4C code. Appl. Radiat. Isot., 67, 535–538.[WoS][Crossref]
  • 7. Balogun, G. I. (2003). Automating some analysis and design calculations of miniature neutron source reactors at CERT (1). Ann. Nucl. Energy, 30, 81–92.[Crossref]
  • 8. Tayyab, M., Showket, P., & Masood, I. (2008). Neutronic analysis for core conversion (HEU-LEU) of Pakistan research reactor-2 (PARR-2). Ann. Nucl. Energy, 35, 1440–1446.[WoS]
  • 9. Hainoun, A., Hajhassan, H., & Ghazi, N. (2009). Determination of major kinetic parameters of the Syrian MNSR for different fuel loading using Monte Carlo technique. Ann. Nucl. Energy, 36, 1663–1667.[Crossref][WoS]
Document Type
Publication order reference
YADDA identifier
bwmeta1.element.-psjd-doi-10_1515_nuka-2015-0037
Identifiers
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