HEAT TRANSFER AT SUPERCRITICAL PRESSURES (SURVEY) 1 Igor Pioro*, Hussam Khartail and Romney Duffey Chalk River Laoratories, AECL, Chalk River, ON, Canada K0J 1J0 Keywords: Supercritical pressure, forced convective heat transfer, water, caron dioxide. Ojectives The ojectives are to assess the work that was done in the area of heat transfer at supercritical pressures, to understand the specifics of heat transfer at these conditions, to compare different prediction methods for supercritical heat transfer in tues and undles, and to choose the most reliale ones. Preliminary Findings The exhaustive literature search, which included hundreds of papers, showed that the majority of correlations were otained in tues and just few of them in other flow geometries including undles. The use of supercritical steam-water in nuclear reactors (Generation IV Nuclear Energy Systems Report 2001) will: Significantly increase thermal efficiency up to 40 45%; Decrease reactor coolant pumping power; Lower containment loadings during loss-of-coolant accidents; Eliminate dryout; and Eliminate steam dryers, steam separators, re-circulation pumps, and steam generators. 1 The presentation is ased on the following papers: 1. Pioro, I.L., Khartail, H.F. and Duffey, R.B., Heat Transfer at Supercritical Pressures (Survey), Proceedings of the 11 th International Conference on Nuclear Engineering (ICONE-11), Shinjuku, Tokyo, Japan, April 20 23, 2003, Paper No. 36454, 13 pages. 2. Duffey, R.B., Khartail, H.F., Pioro, I.L. and Hopwood, J.M., The Future of Nuclear: SCWR Generation IV High Performance Channels, Proceedings of the 11 th International Conference on Nuclear Engineering (ICONE-11), Shinjuku, Tokyo, Japan, April 20 23, 2003, Paper No. 36222, 8 pages.
THERMOPHYSICAL PROPERTIES AT CRITICAL AND SUPERCRITICAL PRESSURES 700 600 p=22.1 MPa p=25.0 MPa Density, kg/m 3 500 400 300 200 100 0 350 360 370 380 390 400
Specific Enthalpy, kj/kg 3000 2500 2000 p=22.1 MPa p=25.0 MPa 1500 350 360 370 380 390 400
600 p=22.1 MPa p=25.0 MPa Specific Heat, kj/kg K 500 400 300 200 100 0 350 360 370 380 390 400
Volume Expansivity, 1/K 1 0.1 0.01 p=22.1 MPa p=25.0 MPa 0.001 300 350 400 450 500 550 600 650
0.7 Thermal Conductivity, W/m K 0.6 0.5 0.4 0.3 0.2 0.1 p=22.1 MPa p=25.0 MPa 0.0 350 360 370 380 390 400
Dynamic Viscosity * 10 5, Pa s 8 7 6 5 4 3 2 350 360 370 380 390 400 p=22.1 MPa p=25.0 MPa
35 p=22.1 MPa p=25.0 MPa 30 Prandtl Numer 25 20 15 10 5 0 350 360 370 380 390 400
KRASNOSHCHEKOV AND PROTOPOPOV (1959, 1960) FOR TUBES 0.35 0.33 0.11 0 = p p w w c c k k Nu Nu µ µ 2 10 3 2 0 64 1 82 1 1 07 1 1 8 7 12 8 ). Re log. (. ) Pr (. Pr Re Nu = + = ξ ξ ξ w w k T T H H µ ) ( ) ( Pr = w w p T T H H c =.
DYADYAKIN AND POPOV (1977) FOR TIGHT 7-ROD BUNDLE WITH HELICAL FINS Nu x = 0.021 Re 0.8 x Pr 0.7 x ρ w ρ 0.45 x µ µ in 0.2 x ρ ρ in 0.1 x 1 + 2.5 D x hy
Heat Transfer Coefficient, kw/m 2 K 50 45 40 35 30 25 20 15 10 5 Bulk 280 300 320 340 360 Experiment (Shitsman, 1963) Correlation (Dittus-Boelter) Correlation (Shitsman, 1959) Correlation (Kondrat'ev, 1969) Correlation (Krasnoshchekov- Protopopov, 1960) Correlation (Ornatsky et al., 1970) Correlation for finned undle (Dyadyakin and Popov, 1977) Correlation (Bishop et al., 1964) Correlation (Kitoh et al., 1999) Water, circular vertical tue, D=8 mm, L=1.5 m, P=23.3 MPa, q=1084 kw/m 2, G=1500 kg/m 2 s, t pc =378.6 o C, H pc =2148 kj/kg 0 1200 1300 1400 1500 1600 1700 1800 Fluid Enthalpy, kj/kg 0.0 0.2 0.4 0.6 0.8 1.0 1.2 1.4 Heated Length, m
Sheath Wall 900 850 800 750 700 650 600 550 500 450 Goran' et al., 1990 Kondrat'ev, 1969 Krasnoshchekov-Protopopov, 1960 Dyadyakin-Popov, 1977 Bishop et al., 1964 Kitoh et al., 1999 Kirillov et al., 1990 CANDU-X Pressure 25 MPa, Mass flux 860 kg/m 2 s Heat flux 670 kw/m 2 Uniform axially and radially D hy =7.71 mm Heated length 5.772 m 43-element undle 12 undles in string Bulk Fluid Temperature 400 Pseudocritical Temperature 350 0.00.51.01.52.02.53.03.54.04.55.05.56.0 Heated Length, m
FINAL REMARKS AND CONCLUSIONS A comparison of various correlations for supercritical heat transfer showed that several correlations can e used for preliminary estimations of heat transfer in tues and undles. However, no one correlation is ale to descrie deteriorated heat transfer in tues. Preliminary calculations of heat transfer and temperature profiles in a CANDU- X supercritical water-cooled reactor operating conditions showed that the proposed concept of this reactor is feasile for future development.
CURRENT EXPERIMENTAL DATA FOR CO 2 LOOP (NORMAL HEAT TRANSFER) Caron dioxide, P out =8.36 MPa, P=1.5 kpa, G=726 kg/m 2 s, Q=1.5 kw, q=26.8 kw/m 2 (uniform heat flux) 70 60 50 40 30 20 10 0 Fluid Bulk Enthalpy, kj/kg 250 260 270 280 290 Inside wall temperature (recalculated from T w ext ) T in Heat transfer coefficient (calculated) HTC (Krasnoshchekov- Protopopov, 1960) cal Tpc = 36. 5 Bulk fluid temperature (calculated) T in, T out, T out mixer, T w ext are measured values Heated length 0 500 1000 1500 2000 2500 Axial Location, mm o C 3000 2500 2000 1500 1000 T out T out mixer Heat Transfer Coefficient, W/m 2 K
(NORMAL, DETERIORATED AND IMPROVED HEAT TRANSFER) 250 300 350 400 450 500 550 280 240 200 160 120 80 40 0 Caron dioxide, P out =8.37 MPa, P=1.7 kpa, G=823 kg/m 2 s, Q=12.0 kw, q=214.3 kw/m 2 (uniform heat flux) T in Fluid Bulk Enthalpy, kj/kg DHT, q/g= 0.26 kj/kg IHT Heat transfer coefficient (calculated) Inside wall temperature (recalculated from T w ext ) T in, T out, T out mixer, T w ext are measured values Bulk fluid temperature (calculated) Heated length cal Tpc = 36. 5 T out 0 500 1000 1500 2000 2500 Axial Location, mm o C 2500 2000 1500 1000 T out mixer Heat Transfer Coefficient, W/m 2 K