DNB experiments for high-conversion PWR core design

Nuclear Engineering and Design 126 (1991) 267-275 North-Holland

267

D N B experiments for high-conversion P W R core design *'** Y. A k i y a m a a, K. Hori b and S. Tsuda ~ a Mitsubishi Atomic Power Industries, Inc., 2-4-1 Shibakouen, Minato-ku, Tokyo, Japan Mitsubishi Heavy Industries Ltd, Takasago Research and Development Center, 2-1-1 Shinhama Arai-cho, ttyogo Prefl, Japan c Kansai Electric Power Company, 3-3-22 Nakanoshima Kita-ku, Osaka, Japan

Received April 1990

It is very important to clarify the departure from nucleate boiling (DNB) performance of core fuel assemblies for the high conversion pressurized water reactors (PWR) [1,2]. To investigate this, DNB experiments were performed in tight lattice rod bundles, using the model fluid Freon 12 and water under the actual operating conditions. In addition, DNB heat flux measurements in an annular-flow channel were carried out for the design of fertile rods, which are installed in thimble tubes.

I. Introduction The feasibility of a high conversion P W R depends not only on the core conversion ratio but also on the plant power capability, which should be maintained at least at the level of the reference conventional P W R plant. The achievable electric power output depends strongly on the D N B characteristics of the core, whose tightened fuel rods configuration may affect its thermal-hydraulic performance. In addition to that, in t h e case that a high conversion P W R is provided with fertile rods for spectral shift control to attain better core performance, the coolability of the fertile rods is also one of the major concerns in thermal-hydraulic design studies, where the annular coolant channel width between a fertile rod and its guide thimble should be optimized with respect to the coolability and the nuclear worth of the rod. To investigate the D N B characteristics associated with the core and the fertile rods, D N B experiments

* Paper presented at the Fourth International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, October 1013, 1989, Karlsruhe, Germany. ** The work that appears in section 2 was done under a joint research contract between the electric utilities group of Kansai, Hokkaido, Tokyo, Chubu, Shikoku and Kyushu, Japan Atomic Power, and Mitsubishi Heavy Industries, Ltd, while the work that appears in section 3 was done by Kansai Electric Power Company and Mitsubishi Heavy Industries, Ltd. 0029-5493/91/$03.50

simulating core fuel assemblies and fertile-rod coolant channels were performed. Freon 12 was used as working fluid for both the experiments. In addition, D N B experiments using water under the actual operating conditions were carried out to obtain absolute D N B heat fluxes of the core fuel assemblies.

2. DNB experiments for core fuel assembly design 2.1. Freon D N B experiments

Freon 12 was selected as the working fluid, because it is familiar in D N B experiments and its scaling factor to water is comparatively well known. The Freon D N B experimental apparatus is located in the Mitsubishi Takasago Research & Development Center. A schematic diagram of the loop is presented in fig. 1. The flow leaving the recirculating pump through the preheater enters the test housing, then flows down between the annulus and turns upward inside the flow channel containing the rod bundle. The resulting two-phase mixture from the test housing then flows into the subcooler, where the heat added by the rod bundle is removed. As shown in fig. 2, the rod bundle consists of seven electrically heated rods of 9.5 mm diameter for the testing of a typical cell, or six heated rods and one unheated rod of 11.8 m m diameter, which simulates a thimble tube for the testing of a thimble cell. The rods are arranged in a hexagonal array with a pitch of 12

© 1991 - E l s e v i e r S c i e n c e P u b l i s h e r s B.V. ( N o r t h - H o l l a n d )

268

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Y. Akiyama et al. / DNB experiments for high-conversion P WR core design

~

~

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.

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mm. The ratio of pitch to rod diameter is 1.26; it is then not a tight but semitight lattice geometry. The axial power distribution is uniform, and the heated length is 750 mm. A sheathed heater rod is provided rather than a directly heated rod, because the heat flux is small. The power input to the peripheral heater rods was reduced by about 10% to avoid D N B occurrence at these rods, which will be affected by the cold flow at the channel wall. Experiments were carried out for the four bundles shown in table 1. The experimental data ranges are: pressure,

2.0, 2.5, 2.9 MPa;

mass flux,

1.9, 2.8, 3.6 × 103 k g / m 2 s;

Inlet subcooling

3 - 3 6 K.

These ranges were determined to cover the actual plant operating conditions by Ahmad's scaling law [3].

Fig. 2. Test section for Freon DNB experiments.

2.2. Water D N B experiments

The actual D N B heat flux of the high-conversion P W R core can not be predicted from the Freon data alone, because the uncertainty on the scaling law between Freon and water data affects the results. Thus, experiments using water under the actual operating conditions were performed.

Table l Rod bundles for Freon DNB experiments Bundle no.

Spacer grid type

Bundle configuration

Rod bowing

1

Simple support grid Actual grid Actual grid Actual grid

Typical cell

No

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No No Yes a

2 3 4

a rod bundle geometry is presented in fig. 10.

Y. Akiyama et al. / DNB experiments for high-conversion P W R core design

flux. The rod bundles described here are small and consist of two types of differently shaped and powered flow channels; the local fluid conditions are then strongly affected by the thermal mixing parameter between the hotter inner and the colder outer channels. These mixing parameters were obtained by measurements of the fluid temperature distribution at the end of the heated length. The program COBRA-IV-I [5] was applied to the evaluation of the Freon data, and T H I N C [6] for the water data. The thermal mixing coefficients obtained are about 0.005 for the all Freon bundles and 0.0035 for the water bundle. These values are very small compared with those of the existing P W R core geometries. It can not be concluded only from these experiments that these small values are proper to the tight lattice core geometry. The experiments using a bundle with a larger number of rods will be required to clarify this. The values obtained were applied to each subchannel analysis of the D N B data.

The experiments were done at the Heat Transfer Research Facility of the Columbia University [4]. The test bundle specification was almost the same as that of rod bundle 1 of the Freon D N B experiments, except that the heated length was extended to 2400 mm and directly heated rods were introduced. The experiments were performed under the following conditions: Pressure mass flux

12.3, 14.7, 16.7 MPa; 2.8, 3.8, 4.9 x 103 k g / m 2 s;

inlet subcooling

20-170 K.

The power input to peripheral heater rods was also reduced by 10%. Consequently D N B did not occur at these rods under any experimental conditions. Quadrant thermocouples, which were distributed in four circumferential orientations, were installed in the central heater rod to notify the circumferential D N B location. Evidence that D N B is liable to occur at a location facing the narrow gap, was not observed. F r o m this finding, it may be concluded that the influence of tightening the fuel rods on the D N B characteristics, at least for the pitch-to-rod diameter ratio used during the present investigation, is not so large.

2.4. D N B data comparison between Freon and water

Stevens et al. [7] found that when the critical quality was plotted against an empirically determined function E = GD 1/4 ( D / L ) 0"59, similar shaped curves occurred and that when these curves are displaced along the E-axis by a constant factor K, which is called the

2.3. Thermal mixing experiments

Knowledge of the local fluid conditions around the central heater rod are required to predict the D N B heat

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270

K Akiyama et al. / DNB experiments for high-conversion P WR core design

"scaling factor", F r e o n a n d water D N B data gave good agreement. After that, Staub [8] reported that the F r e o n data showed good agreement with the water data by Steven's m e t h o d of scaling, except that instead of the heated length, the length from where the equilibrium quality equals zero was used. In Staub's method, even if the ratio of inlet subcooling to latent heat is not equal, or the heated length is not same, the data can be correlated o n the same curve. Staub's m e t h o d was applied to the water D N B d a t a a n d the F r e o n data in the rod b u n d l e 1. Results are shown in fig. 3, where the mass flow scaling factor of 1.365 calculated by A h m a d ' s m e t h o d was used. Agreem e n t between water a n d F r e o n data is poor. The results in the case of the mass flow scaling factor of 1.0 are shown in fig. 4, where the data are in good agreement with each other, considering the substantial deviation of each data. The mass flow scaling factor for the rod bundles used in these experiments is considered to be smaller t h a n that used by A h m a d .

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2.5. Comparison with existing D N B correlations the basis of a large a m o u n t of C o l u m b i a University D N B data. A n d the K f K correlation was developed especially for tight lattice cores. Results are s h o w n in figs. 5, 6 a n d 7 respectively.

The D N B data o b t a i n e d for water were c o m p a r e d with the W-3 [9], EPRI-1 [10] a n d K f K correlations [11]. T h e W-3 correlation is well k n o w n as the P W R design correlation. The EPRI-1 correlation was developed on

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The agreement with the W-3 correlation is poor. The EPRI-1 correlation was converted to a local-condition form to be applicable to the actual core design. The original EPRI-1 correlation slightly underpredicted the present experimental data, then it was multiplied by 1.025. This correlation gives a better agreement, as shown in fig. 6. The EPRI-1 correlation was developed ×10 3 5

4

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on the basis of a wide range of bundle data for a light-water reactor design. Thus, it can be concluded that the D N B characteristics of the tight lattice core presented here are almost same the as those of a conventional P W R core. The K f K correlation slightly overpredicts the data, but the degree of scattering is very small, as shown in fig. 7. A comparison of Freon D N B data in rod bundle 1 with the modified EPRI-1 correlation is presented in fig. 8. In this comparison, to transform the Freon data into water values, a mass flow scaling factor of 1.0 was applied in accordance with the previous discussion. The agreement is good, but the deviation is large compared to the case of water.

2.6. Parameter effect on D N B heat flux

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To investigate the grid structure influence on the D N B characteristics, the data of rod bundle 2, where actual grids were used, were compared with those of rod bundle 1. The deviations arising from these grid structure differences were almost same as those of the test repeatability error band. This comes from the fact that the actual grids referred to here do not have mixing vanes. The Freon D N B data in rod bundle 3, which simulates the thimble cell configuration, were also compared with the modified EPRI-1 correlation. As presented in

Y. Akiyama et a L / DNB experiments for high-conversion P W R core design

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fig. 9, the d a t a were p r e d i c t e d as well as the typical cell data. F i g u r e 10 s h o w s the c r o s s section a n d axial g e o m e t r y of r o d b u n d l e 4, w h i c h s i m u l a t e s r o d b o w i n g . T h e r m o c o u p l e s w e r e installed at the c o n t a c t level a n d also 30

Rod Bundle DNB Power with Rod Bowing(N) Fig. 11. Effect of rod bowing on DNB power.

s h o w n in fig. 11. It m a y be said t h a t the D N B at l o w - q u a l i t y c o n d i t i o n s is m o r e affected b y local g e o m e try c h a n g e s . C o m p a r i s o n o f D N B p o w e r in this r o d b u n d l e w i t h t h o s e in r o d b u n d l e 2 u n d e r the s a m e test

m m d o w n s t r e a m f r o m there. T h e D N B u s u a l l y o c c u r r e d at the c o n t a c t level in s u b c o o l e d o r l o w - q u a l i t y c o n d i tions; o n the o t h e r h a n d , it t e n d e d to o c c u r d o w n s t r e a m f r o m the c o n t a c t level in h i g h - q u a l i t y c o n d i t i o n s as

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Fig. 12. Test section of Freon DNB experiments for fertile-rod design.

Y. Akiyama et aL / DNB experiments for high-conoersion P W R core design

conditions is also shown in fig. 11. The rod bundle with rod bowing gives a slightly higher power under many conditions. This may be explained as follows. The adverse effect on DNB performance caused by rod bowing is small compared with that caused by a grid located downstream from the DNB location, or the enhancement effect caused by a grid located upstream from the DNB location is larger in the bundle with rod bowing, or both effects. In any case, it can be said that the effect of rod bowing is small.

3. DNB experiments for fertile-rod design The DNB experiments for fertile-rod design were performed using the same experimental apparatus described in section 2.1. The test section consists of a heater rod of 9.5 mm diameter and a thimble tube, which form an annular coolant channel as shown in fig. 12. The inner diameter of the thimble tube was varied from 11.5 mm to 13.0 mm to investigate the gap width effect on the DNB characteristics. In addition, experiments consisting of a heated rod in contact with a thimble tube along the axial length were performed. The experiments were performed in the following parameter ranges: pressure,

2.0, 2.5, 2.9 MPa;

Mass flux Inlet subcooling

1.0, 2.0, 3.1 × 103 k g / m 2 s; 5-80 K.

Representative results of the experiments are shown in fig. 13. The data obtained show the following tendencies with respect to the DNB heat flux. In the case of the large gap or the line contact configuration, the DNB heat flux changes continuously with DNB quality. However, in the case of the smaller gaps, the DNB heat flux increases to a high value below a certain quality. This may be explained as follows. In the bubbly flow regime, the interaction between the core liquid flow and the bubbly layer is increased as the gap width becomes smaller in the concentric configuration. This increase of interaction increases the value of the DNB heat flux. On the other hand, in the annular-flow regime, the dependence of the DNB heat flux on gap width is small. Thus, a remarkable transition of DNB heat flux occurs at a certain quality in the concentric-flow channel with a smaller gap. These observations are similar of those for the small round-tube data found by Katto and Yokoya [12]. The pressure drops were also measured near the DNB conditions and compared with the predictions

273

made by the homogeneous model. Representative resuits are shown in fig. 14. The concentric-flow channels with a smaller gap give a higher pressure drop than those predicted by the homogeneous model, and the concentric configuration gives a higher pressure drop than the contact configuration of the same diametral gap. It is considered that this pressure drop behaviour is closely related to the DNB performance. In other words, the larger pressure drop gives a better DNB performance. Using these experimental data, the allowable fertilerod power was evaluated, considering the actual thimble tube structures and the pressure drop of the reference

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274

Y. Akiyama et al. / DNB experiments for high-conversion P W R core design

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core. As shown in fig. 15, the allowable power sharply decreases below the gap w i d t h of 1.0 mm, a n d the difference of allowable power between the concentric configuration and the contact configuration is small at the gap width of 1.0 mm. According to these results a n d the nuclear calculations we have optimized the gap w i d t h to be 1.0 ram, a n d decided n o t to introduce any

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The present e x p e r i m e n t a l investigations of the D N B p e r f o r m a n c e for the high-conversion P W R core design show that the D N B heat flux d e p e n d e n c e s on the operational conditions are almost the same as those of c o n v e n t i o n a l P W R s a n d that the existing D N B correlation of EPRI-1 c a n b e applied with m i n o r modifications to the high-conversion P W R geometry. Consequently, it was decided t h a t the high-conversion P W R core design can b e b a s e d o n the well-established P W R thermal-hydraulic design p r o c e d u r e with sufficient reliability. T h e coolant c h a n n e l w i d t h between a fertile rod and a thimble tube does n o t greatly degrade the D N B p e r f o r m a n c e a n d is optimized to b e 1 m m with respect to its nuclear w o r t h a n d coolability.

References

[1] T. Umeoka et al., Current status of high conversion pressurized water reactor studies, Nucl. Technol. 80 (1988). [2] A. Iizuka et al., Feasibility design studies on HCPWR's with semi-tight core configuration-ll, ANS Meeting, Jackson Hole, September 1988. [3] S.Y. Ahmad, Fluid to fluid modeling of critical heat flux: a compensated distortion model, AECL-3663 (1971). [4] C.F. Fighetti et al., Parametric study of CHF data: compilation of rod bundle CHF data available at the Columbia University heat transfer research facility, EPRI-NP-2609, Vol. 1 (1982). [5] C.L. Wheeler et aL, COBRA-IV-I: an interim version of COBRA for thermal-hydraulic analysis of rod bundle nuclear fuel elements and core, BNWL-1962 (1976). [6] H. Chelemer et al., THINC-IV: an improved program for thermal-hydraulic analysis of rod bundle cores, WCAP7956 (1973). [7] G. F. Stevens et al., A quantitative comparison between burnout data for water at 1000 l b / i n 2 and Freon-12 at 155 l b / i n 2 uniformly heated round tubes vertical upwards, AEWW-R327 (1964). [8] F.W. Staub, Two-phase fluid modeling: the critical heat flux, Nucl. Sci. Eng. 35 (1969). [9] L.S. Tong, An evaluation of the departure from nucleate boiling in bundles of reactor fuel rods, Nucl. Sci. Eng. 33 (1968). [10] D.G. Reddy et al., Parametric study of CHF data: a generalized subchannel CHF correlation for PWR and BWR fuel assemblies, EPRI-NP-2609, Vol. 2 (1983).

Y. Akiyama et al. / DNB experiments for high-conversion P W R core design [11] M. Dalle Donne and W. Hame, Critical heat flux correlation for triangular arrays of rod bundles with tight lattices, including the spiral spacer effect, Nucl. Technol. 71 (1985).

275

[12] Y. Katto and S. Yokoya, A data set of critical heat flux of boiling R-12 in uniformly heated vertical tubes including very large length-to-diameter ratios (1987).