Experimental Study on Heat Transfer to Supercritical

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duffeyr@aecl.ca; khartabilh@aecl.ca. ABSTRACT. This paper presents selected results on heat transfer to supercritical water flowing upward in 1- and 4-m-long.
Proceedings of GLOBAL 2005 Tsukuba, Japan, Oct 9-13, 2005 Paper No. 518

Experimental Study on Heat Transfer to Supercritical Water Flowing in 1- and 4-m-Long Vertical Tubes Pavel KIRILLOV, Richard POMET'KO, Aleksandr SMIRNOV, Vera GRABEZHNAIA, Igor PIORO*1, Romney DUFFEY* and Hussam KHARTABIL* State Scientific Center of the Russian Federation − Institute of Physics and Power Engineering (IPPE) named after A.I. Leipunsky, Obninsk, Russia Tel.: +7-08439-9-8210, Fax: +7-08439-9-8805, E-mail: [email protected] * Chalk River Laboratories, AECL, Chalk River, ON, Canada K0J 1J0 1

Tel.: +1-613-584-8811 ext. 4805; 1Fax: +1-613-584-8213,E-mail: [email protected]; [email protected]; [email protected]

ABSTRACT This paper presents selected results on heat transfer to supercritical water flowing upward in 1- and 4-m-long vertical tubes. Supercritical water heat-transfer data were obtained at pressures of 24 – 25 MPa, mass fluxes of 200 – 1500 kg/m2s, heat fluxes up to 1050 kW/m2 and inlet temperatures from 300 to 380ºC for several combinations of wall and bulk fluid temperatures that were below, at or above the pseudocritical temperature. In general, the experiments confirmed that there are three heat transfer modes for water at supercritical pressures: (1) normal heat transfer characterized in general with heat transfer coefficients (HTCs) similar to those of subcritical convective heat transfer far from critical or pseudocritical regions, which are calculated according to the Dittus-Boelter type correlations, (2) deteriorated heat transfer with lower values of the HTC and hence higher values of wall temperature within some part of a test section compared to those of normal heat transfer and (3) improved heat transfer with higher values of the HTC and hence lower values of wall temperature within some part of a test section compared to those of normal heat transfer. These new heat-transfer data are applicable as a reference dataset for future comparison with supercritical water bundle data and for the verification of scaling parameters between water and modelling fluids. KEYWORDS: forced convection heat transfer, supercritical water, circular tube, nuclear reactor.

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plants. A major contribution to this energy cost reduction would come from boosting the outlet coolant temperature thereby increasing the thermal efficiency of the nuclear power plant. Compared to existing PWRs, the SCW nuclear reactors would involve increasing the coolant pressure from 10 – 16 MPa to about 25 MPa, the inlet temperature to about 350°C, and the outlet temperature to 625°C. The coolant would pass

INTRODUCTION

New reactor concepts (see Table 1) under development at AECL (Khartabil et al., 2005) and at IPPE (Baranaev et al., 2004) have the main design objective of achieving the major reduction in unit energy cost relative to existing pressure-water reactor (PWR) designs (Duffey et al., 2003). The approach builds on using existing operating supercritical water (SCW) experience in fossil-fired thermal power 1

through its pseudocritical reaching the channel outlet.

temperature1

before

deteriorated heat transfer, i.e., lower HTC values compared to those of normal heat transfer, may exist (Pioro and Duffey 2005; Pioro and Duffey 2004, 2003a,b). This new work investigates heat transfer at supercritical conditions using water. The main objective is to create a reference reliable supercritical water heat-transfer dataset for future comparison with supercritical water heat-transfer bundle data and to create a database for future verification of scaling parameters between water, CO2 and fluorocarbons. The supercritical pressure experiments with water were performed in the SKD-1 pumped loop at the Institute of Physics and Power Engineering (IPPE, Obninsk, Russia). This paper describes the heattransfer results and presents a selected set of data for comparison.

Table 1. Major parameters of SCW CANDU and SCW VVER-SCP nuclear reactor concepts. Parameters Reactor type Reactor spectrum Thermal power, MW Electric power, MW Thermal efficiency, % Pressure, MPa Inlet temperature, °C Outlet temperature, °C Flowrate, kg/s Number of fuel channels Number of fuel elements in a bundle Length of a bundle string, m Maximum cladding temperature, °C

SCW CANDU® PT Thermal 2540 1220 48 25 350 625 1320 300 43

VVERSCP RPV Fast 3830 1700 44 25 280 530 1860 241 252

6 850

4 630

2. SUPERCRITICAL WATER TEST FACILITY 2.1

The SKD-1 loop (see Fig. 1) is a hightemperature and high-pressure pumped loop. The operating pressure range is up to 28 MPa at outlet water temperatures up to 500°C. Distilled and deionized water is used in the loop. Water passes from the pump through a flow meter, a preheater, a test section, a mixer, main coolers and back to the pump. Pressurization is achieved with a high-pressure gas (N2). The test section is installed vertically with upward flow. Power is delivered to the test section by a 600 kW (AC) power supply, and cooling is achieved just downstream of the test section using a mixing cooler. While some of the heat from the test section is removed using this mixing cooler, a large amount is removed using the main loop heat exchangers in the discharge circuit of the pump.

Table 2 lists critical parameters of water. Table 2. Critical parameters of water. Parameter Critical pressure Critical temperature Critical density

Unit MPa ºC kg/m3

Water 22.1 374.1 315

Supercritical fluids2 have unique properties (Pioro et al. 2004; Pioro and Duffey 2003b). Beyond the critical point3, the fluid becomes a dense gas. Crossing from high-density fluid to low-density fluid does not involve a distinct phase change. Phenomena such as dryout (critical heat flux) are therefore not relevant. However, at supercritical conditions, 1

The pseudocritical temperature (Tpc > Tcr) corresponds to the point with the maximum specific heat at a given pseudocritical pressure (Ppc > Pcr).

2

Strictly speaking, a supercritical fluid is a fluid at pressures and temperatures that are higher than the critical pressure and critical temperature. A fluid that is at a pressure above the critical pressure but at a temperature below the critical temperature is considered to be a compressed fluid. However, in this paper, the term supercritical fluid includes both terms—a supercritical fluid and compressed fluid.

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The critical point is the point where the distinction between liquid and vapour regions disappears. It is characterized by the state parameters Tcr, vcr and Pcr, which have unique values for each pure substance and have been determined experimentally.

Loop and Water Supply

2

Figure 1. SKD-1 loop schematic: 1 − circulating pump, 2 − mechanical filter, 3 − regulating valves, 4 − electrical heater, 5 − flow meter, 6 − test section, 7 − throttle valve, 8 − mixer-cooler, 9 − discharge tank, 10 − heat exchangers – main coolers, 11 – feedwater tank, 12 − volume compensator, 13 − feedwater pump. 2.2

current passing through the tube wall from the inlet to the outlet power terminals (copper clamps). The test section is wrapped with thermal insulation to minimize heat loss.

Test-Section Design

The test section is a vertical stainless steel4 circular tube (10-mm-ID, 2-mm wall thickness and tube internal arithmetic average surface roughness Ra=0.63– 0.8 µm). The diameter of the test section is close to the equivalent-hydraulic diameter of a SCW reactor fuel bundle. Two heated lengths were used: 1-m-long and 4-m-long5. Water is heated by means of AC electrical 4

2.3

Instrumentation and test matrix

The following test-section parameters were measured or calculated during the experiments: • Test-section current and voltage were used to calculate the power. • Pressure at the test-section inlet. • Temperatures at the test-section inlet and outlet. These temperatures were measured using ungrounded sheathed thermocouples inserted into the fluid stream. The thermocouples were installed just downstream of the mixing chambers,

12Cr18Ni10Ti st. st. was used (content similar to SS-304).

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It is important to perform supercritical heat transfer experiments in one sufficiently long heated test section, which can represent full bundle length (for further explanations, see Pioro and Duffey (2005)). However, a CANDU 12-bundle fuel string may be considered as twelve 0.5-m long independent test sections.

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In general, the data reduction procedure is based on local parameters, which were measured or calculated at each cross section corresponding to the external wall thermocouples. The external wall temperatures, inlet and outlet bulk fluid temperatures and electrical current were used as the basis for local parameters calculations. These local parameters include thermal conductivity and electrical resistivity of the wall material, electrical resistance, power, heat flux, volumetric heat flux, internal wall temperature, heat loss, bulk fluid temperature and pressure. The general and local parameters are defined as follows.

which were used to minimize non-uniformity in the cross-sectional temperature distribution. The thermocouples were calibrated in situ. Wall temperatures at equal intervals (50 mm) along the test section. Twenty-one thermocouples for the 1-m-long tube and 81 thermocouples for the 4-m-long tube were contact welded to the tube outside wall. All these thermocouples were calibrated in situ. Water mass-flow rate was calculated based on the measured pressure drop over a small orifice plate, which was monitored with a differential-pressure (DP) cell. Ambient temperature.

General parameters:

π D2

The instrumentation used to measure the loop parameters was thoroughly checked and calibrated. Uncertainties of primary parameters are summarized in Table 4.



Flow area:

A flow =



Mass flux:

G=

Table 4. Uncertainties of primary parameters.

• •

Total heated area: Ah = π D L. Measured power: POW = V I, where V is the testsection voltage drop, and I is the electrical current.



Average heat flux:



Outlet pressure: Pout = Pin – ∆PTS, where ∆PTS is the total pressure drop across the test section.

Parameter Test-section power Inlet pressure Wall temperatures Mass-flow rate Heat loss

Maximum Uncertainty ±1.0% ±0.25% ±3.0°C ±1.5%

≤ 3%

Table 5. Test matrix. Tin ºC 300– 380

Tout °C 360– 390

Tw °C Tpc and (a) Tb < Tpc; or

(b) Tb < Tpc, Tb = Tpc and Tb > Tpc; 2)

Twin > Tpc and (a) Tb < Tpc and Tb = Tpc; or (b) Tb
Tpc. Typically at the entrance region (i.e.,

CONCLUSIONS

L ≤ 30 ), D

the wall temperature rises sharply (Figs. 2 and 3). In general, this temperature profile is due to the thermal boundary layer development. At the outlet, the colder power clamp lowers heated wall temperature nearby (see Fig. 2). Similar power clamp effect may contribute to the inlet temperature profile, however, only within a short distance. Therefore, any data, i.e., affected with colder clamp effect, should be eliminated from the consideration. Experimental data of supercritical water obtained at higher mass fluxes (G = 1500 kg/m2s) (see Fig. 3) 6

The data obtained at high heat fluxes were chosen for presentation.

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Bulk Fluid Enthalpy, kJ/kg 1350

1800 7000

Water, vertical, ID=10 mm, Pin=24.9 MPa, G=200 kg/m s, Q=2.7 kW, qave=88 kW/m

6000

2

5000

Heat transfer coefficient (calculated)

4000 3000

Tout

lated) perature (calcu Bulk fluid tem

325 Tin, Tout, Tw ext are measured values

400

re ratu mpe id te

350

300

Tin

0

(a)

Tout

100 200 300 400 500 600 700 800 900 1000

(b) Bulk Fluid Enthalpy, kJ/kg

6000 5000

2

ave

2

2

G=201 kg/m s, Q=8.6 kW, qave=275 kW/m

4000 Heat transfer coefficient (calculated)

1700 1800 1900 2000 2100 2200 2300 2400 2500 6000 Water, vertical, ID=10 mm, Pin=24.9 MPa, 5000 G=201 kg/m2s, Q=11.7 kW, q =375 kW/m2

2200

HTC, W/m K

2100

Water, vertical, ID=10 mm, Pin=24.7 MPa,

Hpc

3000 2000

4000 Heat transfer coefficient (calculated) Hpc

1000

Inside wall temperature (calculated from Tw ext)

550

d) ulate

Axial Location, mm

Bulk Fluid Enthalpy, kJ/kg 2000

(calc

Heated length

Axial Location, mm

1900

2000

flu Bulk Tin, Tout, Tw ext are measured values

325

100 200 300 400 500 600 700 800 900 1000

1800

4000

Tpccal = 384.5o C

375

275

1700

5000

Inside wall temperature (calculated from Tw ext)

Heated length 0

6000

3000 2000

Inside wall temperature (calculated from Tw ext)

650

2

275

7000 2

HTC, W/m K

300

1800

3000

Temperature, oC

o

Temperature, C

350

1725

425

T pccal = 384.5o C Tin

1650

Heat transfer coefficient (calculated)

450

Inside wall temperature (calculated from Tw ext)

375

1575

G=200 kg/m s, Q=7.2 kW, qave=227 kW/m

2000

400

1500 2

2

2

1425

Water, vertical, ID=10 mm, Pin=24.9 MPa,

HTC, W/m K

1725

HTC, W/m2K

Bulk Fluid Enthalpy, kJ/kg 1650

1000

o

Temperature, C

o

Temperature, C

600

500 450

Tin, Tout, Tw ext are measured values

400

Tpccal = 383.8o C Tin

350 0

Tout

lated) perature (calcu Bulk fluid tem Heated length

550 500

400 350 300

100 200 300 400 500 600 700 800 900 1000

Tin, Tout, Tw ext are measured values

450 Tin

0

T pccal = 384.5o C (calc Bulk fluid temperature Heated length

ulated)

Tout

100 200 300 400 500 600 700 800 900 1000 Axial Location, mm

Axial Location, mm

(c)

(d)

Figure 2. Temperature and HTC variations along a 1-m circular tube at various heat fluxes and inlet temperatures (mainly deteriorated heat-transfer mode): Nominal flow conditions – Pin=24.9 MPa and G=200 kg/m2s; Hpc=2149 kJ/kg.

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Bulk Fluid Enthalpy, kJ/kg

Hpc Pin=24.0 MPa, G=1500 kg/m2s

o

Temperature, C

400 375

tempera Bulk fluid

350 325 300

0.0

lated) ture (calcu

1.0

1.5

2.0

2.5

3.0

3.5

450 425 400

325

4.0

Tin

0.0

) ext

Tout

T pccal = 381.6o C lated) perature (calcu id Bulk flu tem

375 350

Heated length 0.5

Tw from lated u lc (ca ture pera m e t ll e wa Insid

30 28 26 24 22 20 18 16 14 12 10

2

2

Tout

Tin, Tout, Tw ext are measured values

Tin

Q=28.8 kW, qave= 884 kW/m2

475

) T w ext

r eratu temp wall e id Ins Tpccal = 381.2o C

425

Pin=24.1 MPa, G=1500 kg/m2s

500

2

ro m ted f lcula e (c a

450

Hpc

o

Q=28.6 kW, qave= 874 kW/m

Heat transfer coefficient

Temperature, C

475

34 32 30 28 26 24 22 20 18 16

HTC, kW/m K

Heat transfer coefficient

1600 1700 1800 1900 2000 2100 2200 2300 2400 2500 2600

HTC, kW/m K

Bulk Fluid Enthalpy, kJ/kg

1400 1500 1600 1700 1800 1900 2000 2100 2200 2300 2400

Tin, Tout, Tw ext are measured values Heated length 0.5

1.0

1.5

2.0

2.5

3.0

3.5

4.0

Axial Location, m

Axial Location, m

(a)

(b)

Figure 3. Temperature and HTC variations along a 4-m circular tube: Nominal operating conditions – Pin=24.0 MPa, G=1500 kg/m2s and q=880 kW/m2; Hpc=2159 kJ/kg. cr el ext f flow h in int ℓ out pc TS v w

NOMENCLATURE A D G H HL I k L m P POW Q q qv Ra Rel T V v z

area, m2 inner tube diameter, m mass flux, kg/m2s bulk fluid enthalpy, J/kg heat loss, W current, A thermal conductivity, W/m·K heated length, m mass-flow rate, kg/s pressure, MPa power to the test section, W heat-transfer rate, W heat flux, W/m2 volumetric heat flux, W/m3 arithmetic average surface roughness, µm electrical resistance, Ohm temperature, oC voltage, V volume, m3 axial location, mm

Abbreviations

difference

Subscripts ave b cal

AC AECL

Alternative Current Atomic Energy of Canada Limited

DAS DP HTC ID IPPE NIST

data acquisition system differential pressure heat transfer coefficient inside diameter Institute of Physics and Power Engineering National Institute of Standards and Technology (USA) outside diameter Pressure Tube Pressurized Water Reactor Reactor Pressure Vessel

CANDU® CANada Deuterium Uranium reactor is a registered trademark of AECL

Greek letters ∆

critical electrical external fluid flow heated inlet conditions internal local outlet conditions pseudocritical test section volumetric wall

OD PT PWR RPV

average bulk calculated 7

SCP SCW SKD SS VVER

Pioro, I.L., Khartabil, H.F. and Duffey, R.B., 2004. Heat Transfer to Supercritical Fluids Flowing in Channels – Empirical Correlations (Survey), Nuclear Engineering and Design, Vol. 230, No. 1–3, pp. 69–91.

supercritical pressure supercritical water Supercritical Pressure (in Russian abbreviations) Stainless Steel Water-Water Power Reactor (in Russian abbreviations)

REFERENCES Baranaev, Yu.D., Kirillov, P.L., Poplavskii, V.M. and Sharapov, V.N., 2004. Supercritical-pressure water nuclear reactors, Atomic Energy (Атомная Энергия, (5), стр. 374–380), Vol. 96, No. 4, pp. 345–351. Duffey, R.B., Khartabil, H.F., Pioro, I.L. and Hopwood, J.M., 2003. The Future of Nuclear: SCWR Generation IV High Performance Channels, Proc. of the ICONE-11, Shinjuku, Tokyo, Japan, April 20–23, 2003, Paper No. 36222, 8 pages. Incropera, F.P. and DeWitt, D.P., 2002. Fundamentals of Heat and Mass Transfer, 5th edition, J. Wiley & Sons, New York, NY, USA, p. 124. Khartabil, H.F., Duffey, R.B., Spinks, N. et al., 2005. The Pressure-Tube Concept of Generation IV Supercritical Water-Cooled Reactor (SCWR): Overview and Status, Proc. of the ICAPP-05, Seoul, Korea, May, 2005, Paper #5564. NIST Reference Fluid Thermodynamic and Transport Properties–REFPROP, 2002. NIST Standard Reference Database 23 (on diskette: Executable with Source), Ver. 7.0, E.W. Lemmon, M.O. McLinden, M.L. Huber, U.S. Dept. of Commerce, August. Pioro, I. and Duffey, R., 2005. Experimental Heat Transfer to Supercritical Water Flowing Inside Channels (Survey), Nuclear Engineering and Design, to appear. Pioro, I. and Duffey, R., 2003a. Experimental Heat Transfer to Water Flowing in Channels at Supercritical Pressures (Survey), Proc. of the ANS/ENS Int. Winter Meeting and Nuclear Technology Expo, Embedded Topical Meeting GLOBAL 2003 “Advanced Nuclear Energy and Fuel Cycle Systems”, New Orleans, Louisiana, USA, November 16–20. Pioro, I.L. and Duffey, R.B., 2003b. Literature Survey of the Heat Transfer and Hydraulic Resistance of Water, Carbon Dioxide, Helium and Other Fluids at Supercritical and Near-Critical Pressures, Report AECL-12137/FFC-FCT-409, CRL AECL, April, ISSN 0067-0367, 182 pages.

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