A study of the effect of rod-bowing on critical heat flux

Nuclear Engineering and Design 42 (1977) 237-245 © North-Holland Publishing Company

237

A STUDY OF THE EFFECT OF ROD-BOWING ON CRITICAL HEAT FLUX I. NAKAJIMA, A. KIKUCHI and T. KOBORI Power Reactor and Nuclear Fuel Development Corporation, 4002 Narita, Oarai, Ibaraki, Japan Received 19 November 1976

An experimental study was carried out to determine the effect of rod-bowing on critical heat flux, using an electricallyheated rod cluster. In this experiment, rod-bow was set to occur in the severest subchannel and axially at the middle between the last two spacers, with uniform axial heat flux. The minimum gap between the outer and inner rods was reduced variously to 1.6 mm, 1.0 mm and zero from the nominal value of 2.1 mm. Other experimental conditions were as follows: pressure 7 MPa; mass velocity 640-2600 kg/m 2 sec; inlet subcooling 40-560 kJ/kg. Experimental results show only a slight rod-bowing effect, if any, compared with normal spacing, as confirmed by analysis of three-dimensional heat conduction around the rod-bowing area and by the local steam quality deviations calculated by subchannel analyses.

1. Introduction Critical heat flux is one o f the most important parameters for the thermal design o f water-cooled power reactors. Many experimental studies on burn-out phenomena and their effective parameters have gradually been established. But so far most experiments have been on nominal dimensions o f fuel clusters, despite the fact that the present need is for more precise assessment of nuclear reactor safety. To achieve a higher reliability of reactor components, several tests simulating abnormal conditions in the reactor core have been carried out to measure burn-out as one of the limiting factors in reactor safety. Lahey et al. [1,2] did experiments to confirm the effect of rod-bowing on critical power with BWR fuel. In their experiments, several adjacent r o d - r o d clearances were reduced or several rods were diagonally bowed. The results indicated that local abnormalities in rod geometry had no significant effect on critical power. Hill et al. [3] studied what effect a rod bowed to contact had on critical heat flux in PWR rod bundies. In this experiment, an inner rod was bowed diagonally in such a way that it was in contact with two unbowed rods. Results indicated that a bowing effect occurred only above a pressure-dependent heat flux threshold which was above normal operating conditions o f

a PWR. Rod-bowing has some slight effect on critical heat flux, according to these studies. Lund [4] made DNB experiments in a subcooled, low-pressure rod-bundle with various rod spacings. He concluded that the DNB under subcooled conditions depends primarily on the local hydrodynamics and that the critical heat flux is reduced as the spacing of the rods decreases. But available data are inadequate, and more experiments and analyses are needed to explain the experimental results. This paper describes results obtained for fuel clusters o f a heavy water reactor. Adjacent r o d - r o d clearances were reduced variously to 1.6 mm, 1.0 m m and zero from the nominal value of 2.1 mm to check the effect o f rod-bowing on CHF. The paper also includes analyses o f subchannel and heat conduction in order to explain the behavior of fuel surface temperature.

2. Experimental apparatus and methods 2.1. Apparatus The tests were conducted in a 14 MW heat transfer loop at the O-arai Engineering Center o f PNC. Fig. 1 is a general view of the test loop. Maximum pressure,

I. Naka/ima et al. / Effect of rod-bowing on critical heat flux

238

temperature and flow rate were 10 MPa, 583 K, and 22.2 kg/sec, respectively. The loop consists of steam drum, cooler, test sections, circulating pumps, high pressure condenser, heating power supply system and other components, as shown in fig. 2. Steam generated at the test section separates from the steam-water mixture in the steam drum, condenses in the high pressure condenser and returns to the steam drum. System pressure is adjusted by controlling the heat removal rate in the condenser,

COOLANTOUTLET,/: ~J

OUTLETBENDPIPE

UPPERELECTRODE

%

PRESSURE CONTAINER

8

•HIGH

Fig. 1. General view o f the 14 MW loop.

SCALEFUEL ABLY

CONDENSER

qu (STEAMO) J ~

~L'j PRESSURIZER I I

~ 200kW SUBCOOLERPREHEATER PUMP ®

~

c,

FLOWMETER

.r -'1__.~.~

200LING JACKET 3WER ELECTRODE

IPJPOWERSUPPLY 175V,80kA

r__q__j:~

60V,20kA

Fig.2. Flowdiagramof the loop.

Fig. 3. Test section (length in mm).

239

L Nakafima et al. / Effect of rod-bowing on critical heat flux

Table 1 Dimensions of test section

PRESSURETUBE FUEL ROD O . D ~

NO.

~\

FUEL ROD I.D.

-

i ~-

i-

~117.8#

--53.48¢ 68.22 96.88 116.7 ~ *-8 ~(SPAC~ O.D.) -+°°' (PRESSURETUBEI.D.)

Fig. 4. Cross section of a spacer (diameters in mm). which is a vertical shell-and-tube heat exchanger, with secondary cooling water partially filling the shell side. The heat removal rate can be adjusted by controlling the secondary cooling water level. Water separated from the two-phase flowing mixture is heated by a 1.2 MW preheater to adjust its inlet subcooling and then enters the test section. The heating power supply system consists of a 15 MVA transformer with an on-load tap changer and

OF

FUEL RODS

;36

FUEL RODS OUTER/INNER O.D. ; 14.72/9.74mm ROD-ROD GAP ; 2.1 mm ECCENTRICITY ; 0.6 mm SPACERS O.D. ; 116.7mm PITCH ; 256 mm FLOW AREA; 42.5cm 2 HEATED LENGTH ; 3.7m AVERAGE EQUIVALENT DIAMETER ; 8.84 mm

two units of transformer and silicon-controlled rectifier, each with a maximum current of 40 KA and maximum voltage of 175 V. The maximum heating power is 14 MW, three to four times higher than the maxi-

ELECTRODE /~ASPACER NO. S_PAC -

8

®

[~ i

I ROD BOWING

ECCENTRICITY

1

I

-®

(~ ] .6 rnm

1.0mm 0 mm Fig. 5. Minimum rod-rod gaps.

UNIT; mm

BOH

@ Fig. 6. Axial location of rod-bowing (BOH = beginning of heater; EOH = end of heater).

240

~ R

I

N

G

L Naka/ima et al. /Effect of rod-bowing on critical heat flux

DIMPLE

/

'

"

MINIMUMGAP ; 1.0 mm Fig. 7. Rod-bowing device. mum rated power of a reactor rod cluster. The accuracies of the instrumentation are -+0.5% for the system pressure, -+0.5°C for the inlet subcooling, -+1.0% for the heating power, and -+3% for the flow rate. Fig. 8. Rod contact view.

2.2. Test section

A pressure vessel with the total length of 5.7 m (fig. 3) contains a simulated pressure tube with an inside diameter the same as an actual one. Ceramic liners are set in the pressure vessel to reduce electrolytic effects in the water, therefore actual spacers can be used for the experiments without modification, as shown in fig. 4. Fig. 5 illustrates a cross section of the fuel cluster which has a concentric rod double array, 18 rods in each layer, outer and inner, with diameters almost the same as in BWRs and PWRs, respectively. Table 1 shows the dimensions of the test section. Rod power distribution in the fuel cluster is simulated with different wall thicknesses to adjust the electric current passing through each fuel rod. In this test, rod-bowing was set to occur in the severest subchannel (fig. 5) and axially half-way between two spacers (fig. 6), with uniform axial heat flux. The minimum gaps between the outer and inner rods were re-

- - - 0 - - BEFORE EXPERIMENT - - 0 - - - AFTER EXPERIMENT

-g 2

O

/ t,

, 3.2

j

AXIAl.

I~' 3.4 LENGTH

J

(rn}

Fig. 9. Axial rod gap distribution.

j 3.6

~ j 3.7

L Naka]ima et al. / Effect of rod-bowing on critical heat flux

EOH

241

tion and many others just up-stream of the last spacer (fig. 11). All thermocouples were inserted from the inside of the heater tube and welded to the tube wall. Precisely manufactured stainless steel tubes (type: AISI-316) were used as simulated fuel rods. The accuracies of the tube dimensions for outside diameter and wall thickness were -+0.03 and +-0.04 mm of nominal value.

JJJlJJJ SPACERNo. -'~ T.C ~1 CONTACT POINT -- T.C. ~2

@

2.3. Experimental methods ~

~,

1

I FLOW Fig. 1 0. Axial thermocouple locations.

duced variously from the nominal value of 2.1 mm to 1.6 mm, 1.0 mm and zero in the severest subchannels. In the test section with rod-bowing, the outer and inner rods were bowed by changing the height of spacer dimples, as shown in fig. 7. Fig. 8 shows the rod contact method; the inner rod was bowed by means of a pin welded to the inner rod and the cbntral support tube. Fig. 9 shows the axial distribution of minimum gap before and after experiments. Thus, it was confirmed that no changes occurred in rod gaps in any of the tests. It was found that burn-out takes place, without exception, upstream of the spacer set near the end of the heated section. Thermocouples were therefore attached at two axial positions (fig. 10) to measure the behavior of heater wall temperature, two at the rod-bowed loca-

19-1A

__ 19-1B

19-2

The critical heat flux with rod-bowing was measured in the same way as in the usual steady state burn-out experiment; heating power was increased gradually under constant system pressure, flow rate and inlet temperature until one or more thermocouples indicated temperature rise due to burn-out. At the moment of burn-out heating was partially reduced within 100 msec. Experimental conditions were as follows; pressure, P, 7 MPa, mass velocity, G, 640-2600 kg/m2sec, inlet subcooling, AHin, 40-560 kJ/kg.

3. Experimental results Figs. 12, 13 and 14 plot the experimental burn-out data for mass velocities of 1300, 2000 and 2600 kg/ m2sec, respectively. By visual inspection, fig. 13 shows the 1.6 mm gap yielded the lowest qCHF. Fig. 14 shows the 1.0 mm gap lowest, and in fig. 12 all gaps roughly agree. Superim position shows that the data points in all curves agree equally well, the standard deviation

_

'~

2.0.

1.5

P=7 MPa ] G=1300kg/m2st #O~a

i¢1

ROD-RODGAP O 2.1ram~ • 1.6ram 1 # 1.0ram /

60 T.C. ~;1

T.C. g2

Fig. 11. Radial thermocouple locations.

AVERAGE EXITQUALITY Xe (%)

Fig. 12. Rod-bowing effect (G = 1300 kg/m2sec).

70

242

I. Naka/ima et al. / Effect of rod-bowing on critical heat flux 2 . 0 - -

×10'

ROD NO.

A

6

[

1-1 o x

jHi~ = 170kJ/kg 1.5

q =HF~ 166~10~W/m 2o

~•

0.5 20

310

L 40

,

i

•

61

19-1A ~

36

4. Discussion

4.1. Subchannel analysis A subchannel analysis was done to determine the rod-bowing effect on CHF by use of COBRA lI [5],

P=7 MPa G=2600 kg/m~s ROD-ROD GAP O 2,1mm

~(~j~

~ I k

WOo

L

•

1.6ram

1.0ram 0mm

A ° 1.5

1.010

0 2

I

30

6

~<

4

I

I

40

AVERAGE EXIT QUALITY Xe (%)

Fig. 14. Rod-bowing effect (G = 2600 kg/m 2 sec).

4

2

01

among all being +6%. Evidently the decrease of CHF by rod-bowing is too small to be distinguished from the experimental error band. Burn-out occurred mainly at rod numbers 19, 20 or 36, where the subchannel quality was highest. Fig. 15 gives an example of recorded wall temperatures at various locations for the thermocouple numbers in fig. 11. From the signals registered, it is clear that wall temperatures at contact points did not rise even in the case of inner rod contacts with outer rods.

2.0

~

~-

AVERAGE EXlI QUALITY Xe (%)

'o x

2

6O

Fig. 13. Rod-bowing effect (G --- 2 0 0 0 kg/m2sec).

2.5

4

uJ 0

1.0mm 0rnm

5O

Xa~=341 ' 0

I P=7MPo i G=2000 kg/m2s ROD-ROD GAP O 2.1mm • 1.6mm

1.0

ffJ

G= 2600kg/m2s

50

1-2

6 4

CONTACT POINT

2

0

1%2

CONTACT POINT

42 0

- 5 ~c. TIME

Fig. 15. Thermocouple signals of contact case.

applicability having been verified by comparing it with our experiment [6,7] as regards the effect of rod cluster eccentricity. Fig. 5 illustrates eccentricity of a rod cluster in a pressure tube, assumed in this calculation to be 0.6 ram. The analytical correlation agrees well with the data obtained in the 14 MW loop. The mixing parameter,/3, was used as a function of steam quality as per table 2. As the rod-rod gap was reduced, each subchannel flow area and the input power transferred to each subchannel varied along the axis of the fuel channel. The code was modified to embody the axial variation of the rod-rod gap, flow rate and input power of each subchannel. In particular, the flow area of the hottest subchannel is reduced to 96, 91 and 86% of the nominal value at the bowing location, corresponding with each rod-bowing condition. It is considered that the increase in steam quality of the hottest subchannel is due to these decreased flow areas. Figs. 16(a) and (b) show the distribution of steam quality deviation from the average value for normal and r o d rod contact cases. The steam quality of the hottest subchannel increases from 8.4 to 8.7%. Fig. 17 shows the steam quality deviation of the hottest subchannel from average steam quality. Accordingly, the subchan-

L Nakafima et al. / Effect of rod.bowing on critical heat flux Table 2 Subchannel code

243

I P=7 MPa I G = 2000 kg/m 2s

x z

CODE

; COBRA-IT [5~

o

SUBCOOLED VOID

; LEVY EQ. [8]

>

BULK VOID

; MODIFIED ARMAND EQ.[9~

TWO-PHASE MULTIPLIER ; ARMAND EQ. [9~

10

<

MIXING PARAMETER (/3') ; ~10] (11]

ROD-ROD GAP •

O

--,~ .

.

.

..... X

20 ILl

.

2

.,,

mm

1.6mm 1,0mm Omm

~ \ x

I

L

30

40

50

AVERAGE EXIT QUALITY Xe (~)

Fig. 17. Subchannel analysis. I

0 0

I

I

I

[

I

I

I

I

50

100

STEAM QUALITY X (%) ROD BOWING AXIAL CHANGE • SUBCHANNEL FLOW AREA • ROD-ROD GAP • POWER SUPPLY FROM ROD

this experiment is 0.035 X 106 W/m 2 for each 1% increase in steam quality• It becomes clear from a calculation utilizing the gradient of this experimental correlation that CHF can decrease by only a few percent of its normal value. The decrease of CHF is thus too small to be distinguished from the experimental error band, and the experimental result would show only a small difference between normal and bowed-rod conditions• 4.2. Heat conduction analysis

nel analysis indicates that the increase of the local steam quality deviation due to rod bowing is only 0.2-0.5%. In general, the CHF decreases monotonically with steam quality. The rate of decrease of the CHF in ECCENTRICITY

The temperature distribution was analysed with the three-dimensional heat conduction code TAC-3D [12], in which the following fundamental heat conduction equation is solved by the implicit numerical method:

ECCENTRICITY

-+

_

UNIT ; % P=7MPa

v.kv

2.1 83

aT

where

G =2000kg/m2s

k T q'" O Cp t

Q=6.83MW (I .34 X I 0 W/m 2) ~Hin=210 kJ/kg Xav=40.3 %

(a) 2.1tam

.... q

(b)

Omm

ROD-ROD GAP

Fig. 16. Distribution of percent deviation of steam quality.

= = = = = =

thermal conductivity local temperature volumetric heat generation rate density specific heat time.

Fig. 18 shows a simplified model in which the contact region is assumed to be adiabatic, and the heat transfer at the other surface is assumed to be nucleate boiling, the heat transfer coefficient of which is given by the Jens-Lottes correlation [13], and is of the order of l0 s W/m2°C for the range of experimental

244

L Nakajima et al. / Effect of rod-bowing on critical heat flux ,o .oo %

o~

0 ~ -°6~~''''''~

~"~"2

0~,*"_o*/ vG/ / ~ ,oyV / ~9/ / oval I

d°l ~

~

o0

BOILING

",,?~," o,.,9\'~oc:,,,...o; \ ~,.*o~Y

\

^:.

I

~

.

\

\~ W ~

"°"r .

"~\

.

\ .

ADIABATIC

o;¢.\ \

\

/

/ .Afe.

~-0~"---4--J~

z0;; o.,-oo

~UCLaTE ' -

I

REG O IN

~, / ,{'~

.

by rod bowing would be about 1 5 - 5 0 K. On the other hand, the thermocouple signals at the conlact point in the experiment show almost no temperature difference, as stated. Therefore, the width of the adiabatic area surrounding the contact point is considered to be less than 1 mm, and considerably higher heat transfer could be expected where the gap is very small.

_." - -

o,,,NG

5. Conclusions

.o

Fig. 18. Heat conduction model.

conditions. The radial contact length is assumed to be small, 1 - 2 mm, which would be equivalent to 0 . 0 3 - 0 . 1 4 mm of the rod-rod gap, respectively. Fig. 8 shows that the axial contact length was less than about 10 mm. The contact area is therefore assumed to be 1 × 1 mm, 2 × 2 mm or 1 X 10 mm as a parameter. Fig. 19 indicates the circumferential temperature distribution at the contact position, and leads to the conclusion that the maximum temperature difference between the contact point and other surfaces not affected

ROD SURFACE TEMPERATUREDISTRIBUTION

420 P=7 MPa q=1.83X10 ~ W/rn2 ADIABATIC AREA

40G

(Sxz) ~--

2mmx2mm

~

lrnm×lOrnm

The following points became clear from this experimental study on rod-bowing. (1) Experimental results show that rod-bowing and mass velocity have only a slight effect on critical heat flux in the ranges of experimental variables covered. (2) Subchannel analyses indicate that the decrease of CHF calculated by local steam quality deviation is of the order of 1-2%, which is too small to be distinguished from the experimental error band. (3) Heat conduction analyses show that a considerably higher heat transfer coefficient can be expected where the gap is very narrow, compared with the actual behavior of wall temperature in the experiment.

Acknowledgement The authors wish to thank the members of the PNC Heat Transfer Laboratory for their assistance in carrying out these experiments and for many helpful suggestions. This paper is expanded from the one presented at the ANS/ENS 1976 International Conference [14].

38C

36C .<

~ m l mXm lm 8

34C CONTACT POINT

3X-4o 4o-:,o-io

4,

o

lb :~o 3b 4b

DEGREE

mm CIRCUMFERENTIAL LOCATION

Fig. 19. Rod surface temperature distribution.

References [1] R.T. Lahey Jr., E.E. Polomik and G.E. Dix, Proceedings of an International Meeting on Reactor Heat Transfer (American Nuclear Society, Kerntechnische Gagelschaft im Fiir Kernforschung mbH., Karlsruge, 1973). [2] R.B. Nixon, B. Matzner and R.T. Lahey Jr., ASME Paper No. 75-HT-77 (1976). [3] K.W. Hill, F.E. Motley, F.F. Cadek and J.E. Casterline, ASME Paper No. 75-WA/HT-77 (1975). [4] K.O. Lund, ASME Paper No. 75-HT-49 (1976). [5 ] D.S. Rowe, BNWL-1229, Battelle/Pacific Northwest Laboratory (1970). [6] T. Kobori et al., 5th All Union Heat & Mass Transfer Conference, Paper No. 3-37 (1976).

L Naka/ima et al. / Effect of rod.bowing on critical heat flux [7] T. Kobori, Bull. JSME 19 (131) (1976). [8] S. Levy, GEAP-5157 (1967). [9] A.A. Armand, AERE Trans. 828, Izv. Vsesajuznogo Teplotekh. Inst. 1 (1946) 16. [10] D.S. Rowe, BNWL-371, Mar. (1967). [11] K.F. Rudzinski, K. Singh and C.C. Pierre, Canad. J. Chem. Eng. 50, Apr. (1972).

245

[12] J.F. Peterson, AEC Research and Development Report, GA-9263 UC-32 (1969). [13] W.H. Jens and P.A. Lottes, USAEC Report ANS-4627, Argonne National Laboratory (1951). [14] I. Nakajima, A. Kikuchi and T. Kobori, Trans. of the ANS/ENS International Conference, Washington D.C., Nov. (1976).