Abstract Post shutdown decay heat removal in a fast reactor is one of the most important safety functions which must be accomplished with a very high reliability. To achieve high reliability, the fast breeder reactor design has emphasized on passive or near passive decay heat removal systems utilizing the natural convection in the heat removal path. A typical passive decay heat removal system used in recent designs of fast breeder reactors consists of a sodium to sodium heat exchanger and sodium to air heat exchanger which together remove heat directly from the hot pool to the final heat sink, which is air. Since these are safety systems, it is necessary to confirm the design with detailed numerical analysis. The numerical studies include pool hydraulics, natural convection phenomena in closed loops, flow through narrow gaps between SA, multi-scale modeling, etc. Toward understanding the evolution of thermal hydraulic parameters during natural convection decay heat removal, a three-dimensional CFD model for the primary system coupled with an appropriate one-dimensional model for the secondary system is proposed. The model has been validated against the results of natural convection test conducted in PHENIX reactor. Adopting the model for the Indian PFBR, six different decay heat removal cases have been studied which bring out the effect of safety grade decay heat removal system (SGDHRS) capacity, secondary sodium inventory and inter-wrapper flow heat transfer on the subassembly outlet temperatures that are important for safety evaluation of the reactor. From the results, it is concluded that the delay in initiation of SGDHRS, replacement of intermediate sodium in SGDHRS with NaK and a decrease in the AHX air inlet temperature do not change the temperatures of the primary circuit significantly. The secondary sodium inventory plays an important role in reducing the temperatures in the primary coolant. The beneficial effect of inter-wrapper flow heat transfer on primary temperatures is limited to about 20 K in the fissile zone and 50 K in the blanket zone. These results are very important and give direction for future designs of fast breeder reactors.
[1]
William F. Hughes,et al.
Basic equations of engineering science
,
1964
.
[2]
Hideki Kamide,et al.
Transient experiments on fast reactor core thermal-hydraulics and its numerical analysis: Inter-subassembly heat transfer and inter-wrapper flow under natural circulation conditions
,
2000
.
[3]
H. Kamide,et al.
An experimental study of inter-subassembly heat transfer during natural circulation decay heat removal in fast breeder reactors
,
1998
.
[4]
Frank S. Castellana,et al.
SUBCHANNEL FLOW AND ENTHALPY DISTRIBUTIONS AT THE EXIT OF A TYPICAL NUCLEAR FUEL CORE GEOMETRY.
,
1972
.
[5]
M. A. Gotovskii,et al.
Heat transfer to liquid metals in longitudinally wetted bundles of rods
,
1969
.
[6]
D. Katz,et al.
Fluid Dynamics and Heat Transfer
,
1959
.
[7]
E. H. Novendstern.
Turbulent flow pressure drop model for fuel rod assemblies utilizing a helical wire-wrap spacer system
,
1972
.
[8]
M. Azarian,et al.
Sodium thermal-hydraulics in the pool LMFBR primary vessel
,
1990
.
[9]
B. Launder,et al.
The numerical computation of turbulent flows
,
1990
.
[10]
M. Nishimura,et al.
Investigation of Core Thermohydraulics in Fast Reactors—Interwrapper Flow during Natural Circulation
,
2001
.
[12]
D. Tenchine,et al.
Some thermal hydraulic challenges in sodium cooled fast reactors
,
2010
.