Unprotected Loss of Flow Accident in Small Long Life Gas Cooled Fast Reactor

Abstract:

In post Fukushima nuclear accidents inherent safety capability is necessary against some standard accidents such as unprotected loss of flow (ULOF), unprotected rod run-out transient over power (UTOP), unprotected loss of heat sink (ULOHS). Gas cooled fast reactors is one of the important candidate of 4th generation nuclear power plant and in this paper the safety analysis related to unprotected loss of flow in small long life gas cooled fast reactors has been performed. Accident analysis of unprotected loss of flow include coupled neutronic and thermal hydraulic analysis which include adiabatic model in nodal approach of time dependent multigroup diffusion equations. The thermal hydraulic model include transient model in the core, steam generator, and related systems. Natural circulation based heat removal system is important to ensure inherent safety capability during unprotected accidents. Therefore the system similar to RVACS (reactor vessel auxiliary cooling system) is investigated. As the results some simulations for small 60 MWt gas cooled fast reactors has been performed and the results show that the reactor can anticipate complete pumping failure inherently by reducing power through reactivity feedback and remove the rest of heat through natural circulations.

• Periodical: Applied Mechanics and Materials (Volume 751)
• Pages: 263-267
• DOI: https://doi.org/10.4028/www.scientific.net/AMM.751.263
• Online since: April 2015
• Authors: Su'ud Zaki*
• Keywords: Gas Cooled Fast Reactor, Natural Circulation, Neutronic Model, Thermal Hydraulic Model, Unprotected Loss of Flow

References

[1] T.Y.C. Wei and K.D. Weaver: Initial Requirements for Gas-Cooled Fast Reactor (GFR) System Design, Performance, and Safety Analysis Models. (Gen IV Nuclear Energy System, INEEL/EXT-04-02242, 2004).

DOI: 10.2172/911212

[2] Su'ud Z, Design Study of Small Gas Cooled Fast Nuclear Power Plant for Synergetic Energy System with Renewable Energy by Employing Pump Storage, AMR Vol 983(2014) pp.233-237.

DOI: 10.4028/www.scientific.net/amr.983.233

[3] Su'ud, Z., Sekimoto, H., Annals of Nuclear Energy Vol. 54 (2013), pp.58-66.

[4] Y. Baoan, Y. Yongwei and D. Zhengping, FRLOF-2, A computer program for fast reactors safety analysis, Int. Conf. on Fast Reactor and Related Fuel Cycles, FR '91, Kyoto, (1991), p.3. 1. 1-3. 1. 8.

[5] S.K. Cheng and N.E. Todreas, Hydrodynamic models and correlationx for bare and wire-wrapped hexagonal rod bundles, Nucl. Eng. Des. 92 (1986) 227.

DOI: 10.1016/0029-5493(86)90249-9

[6] R.D. Coffield and R.A. Markley, Natural convection analysis and verification for LMFBR cores, Nucl. Eng. Des. 62 (1980) 181-198.

DOI: 10.1016/0029-5493(80)90028-x

[7] J.G. Guppy, Super System Code, NUREG/CR-3169, (1993).

[8] S. Kakac, R.K. Shah and W. Aung, Handbook of Single Phase Convective Heat Transfer, Wiley, (1987), p.7. 1-7. 62.

[9] N. Nakagawa and K. Tsuchihashi, SLAROM: a code for cell homogenization calculation of fast reactor, (Japan Atomic Energy Research Institute, Ibaraki), JAERI M-1294, (1984).

[10] K.O. Ott and R.J. Neuhold, Nuclear Reactor Dynamics, ANS, (1985).

[11] H. Sekimoto and S. Zaki, Design study of lead and lead-bismuth cooled small long-life nuclear power reactors using metallic and nitride fuel, Nucl. Technol. 109 (1995) 307-313.

DOI: 10.13182/nt109-307

[12] A.E. Waltar and A.B. Reynolds, Fast Breeder Reactors, Pergamon, NewYork, (1981).

[13] Zaki S. dan Sekimoto, H.: Accident Analysis of Lead or Lead-Bistmuth Cooled Small Safe Long-Life Fast Reactor using Metallic or Nitride Fuel, Nucl. Eng. And Design, 162, Elsevier Science, (1996), pp.205-222.

DOI: 10.1016/0029-5493(95)01131-5

[14] Zaki S. : Comparative Study on Safety Performance of Nitride Fueled Lead-Bismuth Cooled Fast Reactor with Various Power Level", Progress of Nuclear Energy, Vol. 32, No. ¾, (1998) p.571~577.

DOI: 10.1016/s0149-1970(97)00045-0