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Experimental research progress on critical heat flux of Chinese PWR
Nuclear Engineering and Design 229 (2004) 213–222
Experimental research progress on critical heat flux of Chinese PWR Xiao Zejun∗ , Song Xianhui, Lang Xuemei, Bai Xuesong, Ma Jieliang, Wang Pengfei, Chen Bingde National Key Laboratory of Bubble Physics & Natural Circulation, Nuclear Power Institute of China, 25 South Third Section, Er Huan Street, P.O. Box 622, Chengdu 610041, PR China Received 23 July 2003; received in revised form 19 November 2003; accepted 9 December 2003
Abstract One of the most important requirements in the design of pressurized water reactor (PWR) is to avoid the occurrence of critical heat flux (CHF). The design criteria for PWR specify that they must be operated at a certain percentage below CHF at all times and locations so as to the cladding temperature of fuel element at safe values. So in the process of safety assessment, CHF is one of important thermal-hydraulic parameters limiting the available power, whose size directly affects safety and economy of PWR nuclear power plant. This paper deals with a summary of experimental research progress on CHF of Chinese PWR. It mainly presents CHF experimental researches of 10 fuel assembly, CHF experimental researches of standard fuel assembly, and CHF experimental progress of non-uniform heated rod bundles. It should be emphasized that it also presents experimental research programs on CHF of Chinese advanced fuel assembly with self-reliance copyright. All CHF data obtained will be used for design improvement of Chinese PWR and R&D program of New Generation 1000 MWe PWR. © 2004 Elsevier B.V. All rights reserved.
1. Introduction In a pressurized water reactor (PWR), the power generated is often limited by the value at which the rod surface is no longer wetted by the boiling liquid. At the point the heat transfer coefficient in core begins to deteriorate and finally results in fuel failure. The heat flux is commonly referred to as departure from nucleate boiling (DNB) or critical heat flux (CHF) (Hoyer, 1998; Katto, 1994). One of the most important requirements in the design of PWR is to avoid the occurrence of CHF. The design criteria for PWR specify that they must be operated at a certain percentage below CHF at all times and locations ∗
Corresponding author. Fax: +86-28-851-88498. E-mail address: [email protected] (X. Zejun).
so as to the cladding temperature of fuel element at safe values. So in the process of safety assessment, CHF is one of important thermal-hydraulic parameters limiting the available power, whose size directly affects safety and economy of nuclear power plant (NPP). For the sake of getting the higher CHF value, all large companies, such as Westinghouse, Framatome, Siemens, ABB, have spent a large quantity of manpower and financial power on improving the performance of fuel assembly. At present, representative high performance fuel assemblies are Performance+ of Westinghouse, AFA-3G of Framatome, HTP of Siemens and System80+ of ABB. In order to attract customer, all large companies are optimizing structure of fuel assembly continuously, and proceeding with CHF experimental researches.
0029-5493/$ – see front matter © 2004 Elsevier B.V. All rights reserved. doi:10.1016/j.nucengdes.2003.12.005
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In spite of a great quantity of experimental and theoretical researches, knowledge of the precise nature of CHF is still incomplete. The engineering importance of CHF has led to the development of a variety of empirical correlations. For engineering design, empirical correlations used have been developed by correlating the available CHF database obtained from particular rod bundles and parameter ranges (Cheng and Müller, 1998). Such correlations are valid over relatively narrow ranges and cannot be extrapolated to conditions far beyond the database. This paper deals with a summary of experimental research progress on CHF of Chinese PWR. It mainly presents CHF experimental researches of 10 fuel assembly, CHF experimental researches of standard fuel assembly, and CHF experimental progress of non-uniform heated rod bundles. It should be emphasized that it also presents experimental research programs on CHF of Chinese advanced fuel assembly (CAFA) with self-reliance copyright from Chinese government. 2. CHF experimental researches of 10 fuel assembly Since the end of 1970s, Nuclear Power Institute of China (NPIC) had proceeded with CHF experimental researches of 3 × 3 and 4 × 4 10 rod bundles with spacer grids. The main objectives of the experiments were as follows: (1) to get CHF correlation; (2) to investigate effect of different spacer grids and different thermal-hydraulic parameters on CHF of Qinshan I 300 MWe NPP; (3) to resolve the related technical problems of Chashma NPP, experimental researches of the effect of a rod bowed on CHF were developed. Furthermore, fuel assembly of Chinese low temperature nuclear heating reactor is the same as that of Qinshan I 300 MWe NPP, CHF experimental research at its operational parameter ranges was also conducted. It should be emphasized that the lengths of test sections with uniform wall thickness does not exceed 1400 mm in all the experimental researches. 2.1. CHF experiments of 4 × 4 rod bundles The experiments were carried out in the smallscale thermal-hydraulic test facility of NPIC. The
Table 1 Main parameters of small-scale thermal-hydraulic test facility Parameter
Value
Working medium Design pressure (MPa) Design temperature (◦ C) Pump head (MPa) Volume flow rate of the main loop (m3 /h) Volume flow rate of test branch (m3 /h) Power of test section (MW)
Deionized water 16 350 0.9 80 30 2.6 (dc 130 V, 20000 A)
specifications of the test loop are shown in Table 1. The test section consists of sixteen 10 mm diameter heater rods with heated length of 1200 mm, which simulates a typical cell in a 4 × 4 array with a rod pitch of 13.3 mm, using no mixing vane spacer grids on 450 mm spacing. The simple support grids were located in unheated section to prevent deflection of heater rods caused by mutual attraction in the electrical field. Two sorts of distances between the topmost spacer grid and the end of heated section were 220 and 20 mm, respectively. According to the two sorts of distances, two sets of thermocouples were located in different location at the end of heated section, and fixed to inside diameter of each tube. These thermocouples are used to detect the rapid and uncontrolled increase in wall temperature while CHF occurs. The main experimental parameters are: pressure of 15.3 MPa, mass velocity of 1111–3056 kg/m2 s, quality of −0.20 to 0.13. Before starting a set of experiments, heat balance tests were carried out to verify reliability of the measuring system. The heat balances were no less than 95%. Based on the experimental data, the following conclusions can be drawn: (1) For the distance between the topmost spacer grid and the end of heated section being 220 mm, 71 sets of experimental data have been obtained. For the distance between the topmost spacer grid and the end of heated section being 20 mm, 83 sets of experimental data have been obtained. By comparing with CHF data of different distance between the topmost spacer grid and the end of heated section, it shows that the distance has no effect on CHF (Fig. 1).
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2.2. Experimental researches of the effect of a rod bowed on CHF
3
220mm 20mm
2.5
2
Q e (MW/m )
2
1.5
1
0.5
0 - 0.3
-0.25
-0.2
-0.15
-0.1
-0.05
0
0.05
0.1
0.15
X
Fig. 1. Effect of spacer grid location on CHF.
(2) By comparing original and repeat runs at the same flow conditions, the repeatability of experimental data was satisfied. (3) The CHF data can be correlated as a function of inlet mass velocity and critical quality. The correlation was as follows: QCHF = A1 + B1 Wg − (C1 + D1 Wg )Xc Where A1 , B1 , C1 , D1 were experimental coefficients, and the unit of QCHF and Wg were MW/m2 and kg/m2 s, respectively. Relative errors of 95% experimental data were less than ±8% (Fig. 2). 3
2.5
2
+8%
-8%
2
Q c (MW/m )
215
1.5
1
0.5
0 0
0.5
1
1.5
2
2.5
2
Q e (MW/m )
Fig. 2. CHF comparison between experiment and calculation.
3
CHF experiments were conducted to measure the effect of a rod bowed with different gap closure both in a coldwall thimble cell and in a typical cell. The experiments were developed using electrically heated 4 × 4 rod bundles with 1200 mm heated length and with axially and radially uniform heat flux, locating no mixing vane spacer grids on a 450 mm spacing (Fig. 3). One of central heater rods was bowed in different gap closure at the topmost spacer grid span where most of CHF were observed in the unbowed geometry experiment (Nagino et al., 1978). Distance between the topmost spacer grid of heated section and the end of heated section was 220 mm. To attain the objectives of the researches, four experimental cases were carried out. The first case, CHF experiment with a typical cell was conducted using unbowed rod bundles. The main objectives of the experiment were as follows: (1) to get CHF correction of unbowed rod bundles with a typical cell; (2) to provide a base case for comparison to experiment data of later bowed rod bundles with a typical cell. The second case, CHF experiments with a typical cell was conducted using three sets of bowed rod bundles. The main objectives of the experiments were to investigate the effect of bowed rod bundles gap closure of 50, 92 and 100% on CHF. The third case, CHF experiment with a coldwall thimble cell was conducted using unbowed rod bundles. The main objectives of the experiment were as follows: (1) to investigate the effect of coldwall on CHF; (2) to provide a base case for comparison to experimental data of later bowed rod bundles with a coldwall thimble cell. The fourth case, CHF experiment with a coldwall thimble cell was conducted using bowed rod bundles having gap closure of 89%. The experiments were carried out in small-scale thermal-hydraulic test facility of NPIC. All the electrical signals from the sensors and transmitter are processed and analyzed by SI35951C data acquisition system and a control system. The control system can detect the rapid and uncontrolled increase in temperature while CHF occurs, and cut off power of heater rod bundles from 100% rate power to 80% rate power automatically.
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Fig. 3. Schematic diagram of test section.
The main experimental parameters are: pressure of 15.3 MPa, mass velocity of 1050–2800 kg/m2 s, quality of −0.20 to 0.15. Based on experimental data, the following conclusions can be drawn: (1) For unbowed rod bundles with a typical cell, 63 sets of experimental data have been obtained. CHF correction of unbowed rod bundles with a typical cell has been got. Relative errors of 95% experimental data were less than ±5%, whose standard deviation is equal to 2.1%. (2) For the typical cell with gap closure of 50%, 42 sets of experimental data have been obtained. For the typical cell with gap closure of 92%, 44 sets of experimental data have been obtained. Gap closure of 50% has no effect on CHF, but gap closure of over 92% has a considerable effect on CHF.
(3) For the typical cell with gap closure of 100%, 63 sets of experimental data have been obtained. There is a considerable effect of a rod bowed on CHF at high subcooling, but less effect at high quality (Fig. 4). The reduction of CHF due to a rod bowed can be correlated as a function of quality and heat flux. (4) For unbowed rod bundles with a coldwall thimble cell, 39 sets of experimental data have been obtained. By comparing coldwall rod bundles and typical rod bundles, coldwall effect brings CHF to decrease by 4.9% (Fig. 5). (5) For bowed rod bundles with a coldwall thimble cell, 62 sets of experimental data have been obtained. The reduction of CHF for a coldwall thimble cell is essentially similar to that of a typical cell.
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(1) to carry out basic CHF experimental research; (2) to investigate the effect of new type spacer grid on CHF; (3) to investigate the effect of bimetallic spacer grid on CHF. These researches provide the basis of thermal-hydraulic design of Qinshan II 600 MWe NPP. Furthermore, CHF Experiment of advanced Chinese 600 MWe PWR (AC600) under low mass flow rate has been completed. It should be emphasized that the lengths of test sections with uniform wall thickness does not exceed 2200 mm in all the experimental researches.
1.2
1.1
(QCHF)bow/(QCHF)nobow
217
1
0.9
0.8
0.7
0.6
3.1. CHF experiments of 4 × 4 rod bundles 0.5 - 0.3
- 0.2
- 0.1
0
0.1
0.2
X
Fig. 4. Ratio of measured (QCHF )bow to predicted (QCHF )no bow vs. local quality.
1.2
1.1
1
0.9
0.8
0.7
0.6
0.5 - 0.3
- 0.2
- 0.1
0
0.1
0.2
X
Fig. 5. Ratio of measured (QCHF )coldwall to predicted (QCHF )typical vs. local quality.
The experiments were carried out in small-scale thermal-hydraulic test facility of NPIC. The test sections consist of two sets of rod bundles. First set of rod bundles consists of sixteen 9.5 mm diameter heater rods with heated length of 1200 and 1000 mm, which simulates a typical cell in a 4 × 4 array with a rod pitch of 12.6 mm, using mixing vane spacer grids on a 270 mm spacing, with a simple support grid at the topmost of rod bundles. Distance between the topmost spacer grid of heated section and the end of heated section is 220 mm. Second set of rod bundles is similar to first set of rod bundles, but it consists of fifteen 9.5 mm diameter heater rods and a 12.2 mm diameter unheated rod. The main experimental parameters are: pressure of 14.7–15.5 MPa, mass velocity of 556–3056 kg/m2 s, quality of −0.12 to 0.26. Based on experimental data, the following conclusions can be drawn: (1) For rod bundles with a typical cell, 144 sets of experimental data have been obtained. CHF correlation is as follows: QCHF = A2 {B2 + C2 Wg − (D2 + E2 Wg )Xc } × (F2 − G2 P)
3. CHF experimental researches of standard fuel assembly Standard fuel assembly is adopted by Qinshan II 600 MWe NPP, whose characteristic is 17 × 17 9.5 fuel assembly. Since the beginning of 1980s, NPIC had proceeded with CHF experimental researches of 3×3 and 4×4 9.5 rod bundles with spacer grids. The main objectives of the experiments were as follows:
Where A2 , B2 , C2 , D2 , E2 , F2 , G2 are experimental coefficients, respectively. Relative errors of 95% experimental data were less than ±12% (Fig. 6), whose root mean square error is equal to 6.9%. (2) For rod bundles with a coldwall thimble cell, 68 sets of experimental data at pressure of 14.7 MPa have been obtained. CHF correlation is as follows: QCHF = A3 {B3 + C3 Wg − (D3 + E3 Wg )Xc }
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2
2
Qc(MW/m )
2.5
+12%
1.5
- 12%
1
0.5
0 0
0.5
1
1.5
2
2.5
3
2
Qe(MW/m )
Fig. 6. CHF comparison between experiment and calculation.
Where A3 , B3 , C3 , D3 , E3 are experimental coefficients, respectively. Relative errors of 95% experimental data were less than ±10% (Fig. 7), whose root mean square error is equal to 5.8%. (3) By comparing coldwall rod bundles and typical rod bundles, coldwall effect brings CHF to decrease by 2%.
(1) Two hundred and forty-one sets of experimental data have been obtained. CHF correlation is as follows: QCHF = {A4 + B4 Wg − (C4 + D4 Wg )Xc }
3.2. CHF experiment of AC600 under low mass flow rate
× (E4 − F4 P)
Though CHF correlation reliable to normal operating conditions of Qinshan II 600 MWe NPP, it has a 3
2. 5 +10%
2
2
Qc (MW/m )
relatively narrow range of validity. The correct characterization of CHF under low flow condition is of particular interest when predicting the reactor core behavior during a loss of coolant accident or steam line break accident without offsite power in PWR (Kim et al., 2000; Chun et al., 2001). In order to achieve an optimal design and to ensure a high degree of safety in AC600, CHF experiment under low mass flow rate was carried out. The experiments were carried out in small-scale thermal-hydraulic test facility of NPIC. The test section consists of sixteen 9.5 mm diameter heater rods with heated length of 1000 mm, which simulates a typical cell in a 4 × 4 array with a rod pitch of 12.6 mm, using a mixing vane spacer grid on a 270 mm spacing, with a simple support grid at the topmost of rod bundles. Distance between the topmost spacer grid and the end of heated section is 20 mm. The main experimental parameters are: pressure of 9.8–15.88 MPa, mass velocity of 278–1389 kg/m2 s, quality of −0.20 to 0.34. Based on experimental data, the following conclusions can be drawn:
- 10%
1. 5
Where A4 , B4 , C4 , D4 , E4 , F4 are experimental coefficients, respectively. Relative errors of 95% experimental data were less than ±10% (Fig. 8), whose standard deviation is equal to 5.81%. (2) A calculated DNB ratio (DNBR) value greater than this design limit provides assurance that there is at least a 95% probability at the 95% confidence level that CHF will not occur. DNBR is equal to 1.10.
1
4. CHF experimental progress of non-uniform heated rod bundles
0. 5
0 0
0. 5
1
1. 5
2
2. 5
3
4.1. Test facility
2
Qe (MW/m )
Fig. 7. CHF comparison between experiment and calculation.
At the beginning of 1970s, small-scale thermalhydraulic test facility had been constructed in NPIC. In
X. Zejun et al. / Nuclear Engineering and Design 229 (2004) 213–222 Table 2 Main parameters of Freon thermal-hydraulic test facility
3
2. 5
+10%
2
2
Qc (MW/m )
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- 10% 1. 5
1
Parameter
Value
Working medium Design pressure (MPa) Design temperature (◦ C) Operating pressure (MPa) Operating temperature (◦ C) Maximum power (kW) Maximum flow rate (kg/s) Pump head (MPa)
Freon 12 4.0 120 2.77 60 500 6.82 1
0. 5
0 0
0. 5
1
1. 5
2
2. 5
3
2
Qe (MW/m )
Fig. 8. CHF comparison between experiment and calculation.
order to develop CHF experimental researches of full length non-uniform heated rod bundles, NPIC have constructed Freon thermal-hydraulic test facility and large-scale thermal-hydraulic test facility. 4.1.1. Freon thermal-hydraulic test facility Freon thermal-hydraulic test facility is composed of main loop system, test section branches, purification system, refrigeration system, makeup and discharge system, pressurization system, vacuumized system
and safety relief system, etc. The main components include refrigeration equipment, preheater, mixing condenser, heat exchanger, pressurizer, canned pump, vacuum pump. Fig. 9 shows its schematic diagram. The main parameters are shown in Table 2. 4.1.2. Large-scale thermal-hydraulic test facility Large-scale thermal-hydraulic test facility is composed of a high-temperature high-pressure water test loop and a large heating resource. The loop consists of main loop system, test section branches, water purification system, makeup water system, pressurization system and safety relief system, etc. The main components include pump, heat exchanger, pressurizer, storage tank, mixing condenser, preheater. Fig. 10 shows its schematic diagram. The main parameters are shown in Table 3.
Fig. 9. Schematic diagram of Freon thermal-hydraulic test facility.
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Fig. 10. Schematic diagram of large-scale thermal-hydraulic test facility.
4.2. Research review The prediction of CHF has been, in most cases, based on the empirical correlation. For PWR fuel assembly, the local parameter correlation requires the local thermal-hydraulic conditions usually calculated by Table 3 Main parameters of large-scale thermal-hydraulic test facility Parameter
Value
Working medium
Deionized water 20 370 17.5 355 500 80 1 220 300 70/15 10 (dc 250 V, 40000 A) 453
Design pressure (MPa) Design temperature (◦ C) Operating pressure (MPa) Operating temperature (◦ C) Volume flow rate of large pump (m3 /h) Volume flow rate of small pump (m3 /h) Pump head (MPa) Volume flow rate of the main loop (m3 /h) Maximum temperature of the main loop (◦ C) Volume flow rate of test branch (m3 /h) Power of test section (MW) Volume flow rate of cooling water (m3 /h)
a subchannel analysis code. The cross-sectional averaged fluid conditions of subchannel, however, are not sufficient for determining CHF, especially for the case of non-uniform axial heat flux distribution (Hwang et al., 2001). Many investigators have studied on the influence of upstream history of heat flux variation to CHF. But in China, because of some technical reasons, all CHF engineering experiments, were almost done through part length uniform heated rod bundles in the past, data of which were disposed through cross-sectional mean parameter on the basis of ‘local conditions’ hypothesis. The hypothesis assumes that there is a unique relationship between CHF and local thermal-hydraulic conditions. In reality, axial non-uniform power distribution of heated rod bundles has an important effect on critical point, CHF conditions and CHF value under any circumstances, which is close to axial power distribution of NPP fuel assemblies (Zejun et al., 2003). In the early of 1990s, NPIC started to investigate CHF experimental technique of full length non-uniform heated rod bundles, and successively proceeded with CHF research of single passage non-uniform heated rod, CHF experimentally technical researches
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of non-uniform heated rod bundles, manufacture of non-uniform heated rods and technical researches of critical detection. Because of the technical difficulty and support fund of scientific researches, experimental research progress on CHF of non-uniform heated rod bundles became slower and slower. In the middle of 1990s, NPIC undertook experimental research project on CHF of high performance fuel assembly from Chinese government, continued to proceed with CHF experimental research of non-uniform heated rod bundles. On the basis of absorption and digestion of the advanced design technique, NPIC mastered design key technique, for example, “O” type seal technique and insulation technique, and completed design and manufacture of test section. In the early of 2000s, NPIC had developed experimental research on CHF of 5 × 5 full length non-uniform heated rod bundles. The main objectives of the research were as follows: (1) to compare with the thermal-hydraulic performance between standard fuel assembly and fuel assembly with mid span mixing grid (MSMG); (2) to verify the conformability between CHF experimental data of rod bundles with MSMG and WRB-A correlation. 4.3. Experimental progress The experiments were carried out in large-scale thermal-hydraulic test facility of NPIC. The test sections consist of two sets of rod bundles. First set of rod bundles consists of sixteen 9.5 mm diameter heater rods with heated length of 3657 mm, which simulates a typical cell in a 5 × 5 array with a rod pitch of 12.6 mm, using mixing vane spacer grids on a 521 mm spacing, with simple support grids in the middle of the mixing vane spacer grids. Second set of rod bundles is similar to first set of rod bundles, but it consists of MSMG in the middle of the mixing vane spacer grids. The MSMG (three), placed in the upper of the rod bundles, enhance the mixing capabilities of the flow, thus delaying the occurrence of a vapor blanket causing CHF. The rod bundles have a non-uniform radial power distribution with sixteen peripheral rods having less power than nine central rods. It also has a non-uniform cosine axial power distribution which is made by smoothly changing the inside diameter of rods with axial position. For central rods, the thickness of the heater tube was about 0.44 mm at the maximum
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heat flux location and about 1.99 mm at the minimum heat flux location. For peripheral rods, the thickness of the heater tube was about 0.525 mm at the maximum heat flux location and about 2.6 mm at the minimum heat flux location. Seven sets of thermocouples for each tube are located in the ceramic pieces, which are fitted to inside diameter of each tube. The main experimental parameters are: pressure of 9.8–16.3 MPa, mass velocity of 1166–2613 kg/m2 s, quality of −0.005 to 0.306. Based on experimental data, the following conclusions can be drawn: (1) By comparing with the experimental data between standard fuel assembly and fuel assembly with MSMG, it shows that thermal-hydraulic performance of fuel assembly with MSMG is better than that of standard fuel assembly. (2) By calculating the local thermal-hydraulic parameters using FLICA code, it shows that CHF experimental data of rod bundles with MSMG are conform with WRB-A correlation so much.
5. Summary Since the end of 1970s, NPIC had completed a great amount of CHF experiments. With the construction of Freon thermal-hydraulic test facility and large-scale thermal-hydraulic test facility, on the basis of improvement of design technique and experimental technique continuously, NPIC have had the top capability of developing CHF experimental research of full length non-uniform heated rod bundles, and completed CHF experimental research of high performance fuel assembly. At present, NPIC are undertaking experimental research project on CHF of CAFA with self-reliance copyright from Chinese government, and developing a large quantity of experimental researches on CHF. The main research programs are as follows: (1) While water flowing through rod bundles with different kinds of spacer grids, its velocity fields will be calculated using CFX5.5 code. By comparing with thermal-hydraulic performance among different kinds of spacer grids, three kinds of better performance of spacer grids will be collected. (2) In Freon thermal-hydraulic test facility, CHF experiments of part length uniform heated rod
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(3)
(4)
(5)
(6)
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bundles with three kinds of spacer grids will be developed to filter spacer grids. By comparing with CHF among rod bundles with three kinds of spacer grids, the best performance spacer grid will be collected. In small-scale thermal-hydraulic test facility, hydraulic experiment of spacer grids will be conducted to measure form resistance coefficient of structural grid and MSMG. In large-scale thermal-hydraulic test facility, subchannel mixing experiment of rod bundles will be developed to measure subchannel mixing coefficient. In large-scale thermal-hydraulic test facility, CHF experiments of uniform heated rod bundles will be developed. By using FLICA code to calculate local thermal-hydraulic parameters, CHF correlation of uniform heated rod bundles will be got. In large-scale thermal-hydraulic test facility, CHF experiment of full length non-uniform heated rod bundles will be developed. By using FLICA code to calculate local thermal-hydraulic parameters, CHF correlation of full length non-uniform heated rod bundles will be got. By comparing with CHF between CAFA and the other advanced fuel assemblies, better thermal-hydraulic performance of CAFA will be verified.
All CHF data obtained will be used for design improvement of Chinese PWR and R&D program of New Generation 1000 MWe PWR. References Cheng, X., Mˆuller, U., 1998. Critical heat flux and turbulent mixing in hexagonal tight rod bundles. Int. J. Multiphase Flow 24, 1245–1263. Chun, S.-Y., Chung, H.-J., Moon, S.-K., Yang, S.-K., Chung, M.-K., Schoesse, T., Aritomi, M., 2001. Effect of pressure on critical heat flux in uniformly heated vertical annulus under low flow conditions. Nucl. Eng. Des. 203, 159–174. Hoyer, N., 1998. Calculation of dryout and post-dryout heat transfer for tube geometry. Int. J. Multiphase Flow 24 (2), 319– 334. Hwang, D.-H., Park, C., Zee, S.-Q., 2001. A phenomenological approach to correcting DNB-type critical heat flux for non-uniform axial power shapes. Int. J. Heat Mass Transfer 44, 4483–4492. Katto, Y., 1994. Critical heat flux. Int. J. Multiphase Flow 20, 53–90. Kim, H.C., Baek, W.-P., Chang, S.H., 2000. Critical heat flux water in vertical round tubes at low pressure and low flow conditions. Nucl. Eng. Des. 199, 49–73. Nagino, Y., Shodai, N., Matsumoto, C., Ueji, M., 1978. Rod bowed to contact departure from nucleate boiling tests in coldwall thimble cell geometry. J. Nucl. Sci. Technol. 15 (8), 568– 573. Zejun, X., Bingde, C., Zhongpeng, L., Dounan, J., 2003. Research progress of experiment on critical heat flux. Nucl. Power Eng. 24 (1), 24–27 (in Chinese).
Experimental research progress on critical heat flux of Chinese PWR Xiao Zejun∗ , Song Xianhui, Lang Xuemei, Bai Xuesong, Ma Jieliang, Wang Pengfei, Chen Bingde National Key Laboratory of Bubble Physics & Natural Circulation, Nuclear Power Institute of China, 25 South Third Section, Er Huan Street, P.O. Box 622, Chengdu 610041, PR China Received 23 July 2003; received in revised form 19 November 2003; accepted 9 December 2003
Abstract One of the most important requirements in the design of pressurized water reactor (PWR) is to avoid the occurrence of critical heat flux (CHF). The design criteria for PWR specify that they must be operated at a certain percentage below CHF at all times and locations so as to the cladding temperature of fuel element at safe values. So in the process of safety assessment, CHF is one of important thermal-hydraulic parameters limiting the available power, whose size directly affects safety and economy of PWR nuclear power plant. This paper deals with a summary of experimental research progress on CHF of Chinese PWR. It mainly presents CHF experimental researches of 10 fuel assembly, CHF experimental researches of standard fuel assembly, and CHF experimental progress of non-uniform heated rod bundles. It should be emphasized that it also presents experimental research programs on CHF of Chinese advanced fuel assembly with self-reliance copyright. All CHF data obtained will be used for design improvement of Chinese PWR and R&D program of New Generation 1000 MWe PWR. © 2004 Elsevier B.V. All rights reserved.
1. Introduction In a pressurized water reactor (PWR), the power generated is often limited by the value at which the rod surface is no longer wetted by the boiling liquid. At the point the heat transfer coefficient in core begins to deteriorate and finally results in fuel failure. The heat flux is commonly referred to as departure from nucleate boiling (DNB) or critical heat flux (CHF) (Hoyer, 1998; Katto, 1994). One of the most important requirements in the design of PWR is to avoid the occurrence of CHF. The design criteria for PWR specify that they must be operated at a certain percentage below CHF at all times and locations ∗
Corresponding author. Fax: +86-28-851-88498. E-mail address: [email protected] (X. Zejun).
so as to the cladding temperature of fuel element at safe values. So in the process of safety assessment, CHF is one of important thermal-hydraulic parameters limiting the available power, whose size directly affects safety and economy of nuclear power plant (NPP). For the sake of getting the higher CHF value, all large companies, such as Westinghouse, Framatome, Siemens, ABB, have spent a large quantity of manpower and financial power on improving the performance of fuel assembly. At present, representative high performance fuel assemblies are Performance+ of Westinghouse, AFA-3G of Framatome, HTP of Siemens and System80+ of ABB. In order to attract customer, all large companies are optimizing structure of fuel assembly continuously, and proceeding with CHF experimental researches.
0029-5493/$ – see front matter © 2004 Elsevier B.V. All rights reserved. doi:10.1016/j.nucengdes.2003.12.005
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In spite of a great quantity of experimental and theoretical researches, knowledge of the precise nature of CHF is still incomplete. The engineering importance of CHF has led to the development of a variety of empirical correlations. For engineering design, empirical correlations used have been developed by correlating the available CHF database obtained from particular rod bundles and parameter ranges (Cheng and Müller, 1998). Such correlations are valid over relatively narrow ranges and cannot be extrapolated to conditions far beyond the database. This paper deals with a summary of experimental research progress on CHF of Chinese PWR. It mainly presents CHF experimental researches of 10 fuel assembly, CHF experimental researches of standard fuel assembly, and CHF experimental progress of non-uniform heated rod bundles. It should be emphasized that it also presents experimental research programs on CHF of Chinese advanced fuel assembly (CAFA) with self-reliance copyright from Chinese government. 2. CHF experimental researches of 10 fuel assembly Since the end of 1970s, Nuclear Power Institute of China (NPIC) had proceeded with CHF experimental researches of 3 × 3 and 4 × 4 10 rod bundles with spacer grids. The main objectives of the experiments were as follows: (1) to get CHF correlation; (2) to investigate effect of different spacer grids and different thermal-hydraulic parameters on CHF of Qinshan I 300 MWe NPP; (3) to resolve the related technical problems of Chashma NPP, experimental researches of the effect of a rod bowed on CHF were developed. Furthermore, fuel assembly of Chinese low temperature nuclear heating reactor is the same as that of Qinshan I 300 MWe NPP, CHF experimental research at its operational parameter ranges was also conducted. It should be emphasized that the lengths of test sections with uniform wall thickness does not exceed 1400 mm in all the experimental researches. 2.1. CHF experiments of 4 × 4 rod bundles The experiments were carried out in the smallscale thermal-hydraulic test facility of NPIC. The
Table 1 Main parameters of small-scale thermal-hydraulic test facility Parameter
Value
Working medium Design pressure (MPa) Design temperature (◦ C) Pump head (MPa) Volume flow rate of the main loop (m3 /h) Volume flow rate of test branch (m3 /h) Power of test section (MW)
Deionized water 16 350 0.9 80 30 2.6 (dc 130 V, 20000 A)
specifications of the test loop are shown in Table 1. The test section consists of sixteen 10 mm diameter heater rods with heated length of 1200 mm, which simulates a typical cell in a 4 × 4 array with a rod pitch of 13.3 mm, using no mixing vane spacer grids on 450 mm spacing. The simple support grids were located in unheated section to prevent deflection of heater rods caused by mutual attraction in the electrical field. Two sorts of distances between the topmost spacer grid and the end of heated section were 220 and 20 mm, respectively. According to the two sorts of distances, two sets of thermocouples were located in different location at the end of heated section, and fixed to inside diameter of each tube. These thermocouples are used to detect the rapid and uncontrolled increase in wall temperature while CHF occurs. The main experimental parameters are: pressure of 15.3 MPa, mass velocity of 1111–3056 kg/m2 s, quality of −0.20 to 0.13. Before starting a set of experiments, heat balance tests were carried out to verify reliability of the measuring system. The heat balances were no less than 95%. Based on the experimental data, the following conclusions can be drawn: (1) For the distance between the topmost spacer grid and the end of heated section being 220 mm, 71 sets of experimental data have been obtained. For the distance between the topmost spacer grid and the end of heated section being 20 mm, 83 sets of experimental data have been obtained. By comparing with CHF data of different distance between the topmost spacer grid and the end of heated section, it shows that the distance has no effect on CHF (Fig. 1).
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2.2. Experimental researches of the effect of a rod bowed on CHF
3
220mm 20mm
2.5
2
Q e (MW/m )
2
1.5
1
0.5
0 - 0.3
-0.25
-0.2
-0.15
-0.1
-0.05
0
0.05
0.1
0.15
X
Fig. 1. Effect of spacer grid location on CHF.
(2) By comparing original and repeat runs at the same flow conditions, the repeatability of experimental data was satisfied. (3) The CHF data can be correlated as a function of inlet mass velocity and critical quality. The correlation was as follows: QCHF = A1 + B1 Wg − (C1 + D1 Wg )Xc Where A1 , B1 , C1 , D1 were experimental coefficients, and the unit of QCHF and Wg were MW/m2 and kg/m2 s, respectively. Relative errors of 95% experimental data were less than ±8% (Fig. 2). 3
2.5
2
+8%
-8%
2
Q c (MW/m )
215
1.5
1
0.5
0 0
0.5
1
1.5
2
2.5
2
Q e (MW/m )
Fig. 2. CHF comparison between experiment and calculation.
3
CHF experiments were conducted to measure the effect of a rod bowed with different gap closure both in a coldwall thimble cell and in a typical cell. The experiments were developed using electrically heated 4 × 4 rod bundles with 1200 mm heated length and with axially and radially uniform heat flux, locating no mixing vane spacer grids on a 450 mm spacing (Fig. 3). One of central heater rods was bowed in different gap closure at the topmost spacer grid span where most of CHF were observed in the unbowed geometry experiment (Nagino et al., 1978). Distance between the topmost spacer grid of heated section and the end of heated section was 220 mm. To attain the objectives of the researches, four experimental cases were carried out. The first case, CHF experiment with a typical cell was conducted using unbowed rod bundles. The main objectives of the experiment were as follows: (1) to get CHF correction of unbowed rod bundles with a typical cell; (2) to provide a base case for comparison to experiment data of later bowed rod bundles with a typical cell. The second case, CHF experiments with a typical cell was conducted using three sets of bowed rod bundles. The main objectives of the experiments were to investigate the effect of bowed rod bundles gap closure of 50, 92 and 100% on CHF. The third case, CHF experiment with a coldwall thimble cell was conducted using unbowed rod bundles. The main objectives of the experiment were as follows: (1) to investigate the effect of coldwall on CHF; (2) to provide a base case for comparison to experimental data of later bowed rod bundles with a coldwall thimble cell. The fourth case, CHF experiment with a coldwall thimble cell was conducted using bowed rod bundles having gap closure of 89%. The experiments were carried out in small-scale thermal-hydraulic test facility of NPIC. All the electrical signals from the sensors and transmitter are processed and analyzed by SI35951C data acquisition system and a control system. The control system can detect the rapid and uncontrolled increase in temperature while CHF occurs, and cut off power of heater rod bundles from 100% rate power to 80% rate power automatically.
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Fig. 3. Schematic diagram of test section.
The main experimental parameters are: pressure of 15.3 MPa, mass velocity of 1050–2800 kg/m2 s, quality of −0.20 to 0.15. Based on experimental data, the following conclusions can be drawn: (1) For unbowed rod bundles with a typical cell, 63 sets of experimental data have been obtained. CHF correction of unbowed rod bundles with a typical cell has been got. Relative errors of 95% experimental data were less than ±5%, whose standard deviation is equal to 2.1%. (2) For the typical cell with gap closure of 50%, 42 sets of experimental data have been obtained. For the typical cell with gap closure of 92%, 44 sets of experimental data have been obtained. Gap closure of 50% has no effect on CHF, but gap closure of over 92% has a considerable effect on CHF.
(3) For the typical cell with gap closure of 100%, 63 sets of experimental data have been obtained. There is a considerable effect of a rod bowed on CHF at high subcooling, but less effect at high quality (Fig. 4). The reduction of CHF due to a rod bowed can be correlated as a function of quality and heat flux. (4) For unbowed rod bundles with a coldwall thimble cell, 39 sets of experimental data have been obtained. By comparing coldwall rod bundles and typical rod bundles, coldwall effect brings CHF to decrease by 4.9% (Fig. 5). (5) For bowed rod bundles with a coldwall thimble cell, 62 sets of experimental data have been obtained. The reduction of CHF for a coldwall thimble cell is essentially similar to that of a typical cell.
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(1) to carry out basic CHF experimental research; (2) to investigate the effect of new type spacer grid on CHF; (3) to investigate the effect of bimetallic spacer grid on CHF. These researches provide the basis of thermal-hydraulic design of Qinshan II 600 MWe NPP. Furthermore, CHF Experiment of advanced Chinese 600 MWe PWR (AC600) under low mass flow rate has been completed. It should be emphasized that the lengths of test sections with uniform wall thickness does not exceed 2200 mm in all the experimental researches.
1.2
1.1
(QCHF)bow/(QCHF)nobow
217
1
0.9
0.8
0.7
0.6
3.1. CHF experiments of 4 × 4 rod bundles 0.5 - 0.3
- 0.2
- 0.1
0
0.1
0.2
X
Fig. 4. Ratio of measured (QCHF )bow to predicted (QCHF )no bow vs. local quality.
1.2
1.1
1
0.9
0.8
0.7
0.6
0.5 - 0.3
- 0.2
- 0.1
0
0.1
0.2
X
Fig. 5. Ratio of measured (QCHF )coldwall to predicted (QCHF )typical vs. local quality.
The experiments were carried out in small-scale thermal-hydraulic test facility of NPIC. The test sections consist of two sets of rod bundles. First set of rod bundles consists of sixteen 9.5 mm diameter heater rods with heated length of 1200 and 1000 mm, which simulates a typical cell in a 4 × 4 array with a rod pitch of 12.6 mm, using mixing vane spacer grids on a 270 mm spacing, with a simple support grid at the topmost of rod bundles. Distance between the topmost spacer grid of heated section and the end of heated section is 220 mm. Second set of rod bundles is similar to first set of rod bundles, but it consists of fifteen 9.5 mm diameter heater rods and a 12.2 mm diameter unheated rod. The main experimental parameters are: pressure of 14.7–15.5 MPa, mass velocity of 556–3056 kg/m2 s, quality of −0.12 to 0.26. Based on experimental data, the following conclusions can be drawn: (1) For rod bundles with a typical cell, 144 sets of experimental data have been obtained. CHF correlation is as follows: QCHF = A2 {B2 + C2 Wg − (D2 + E2 Wg )Xc } × (F2 − G2 P)
3. CHF experimental researches of standard fuel assembly Standard fuel assembly is adopted by Qinshan II 600 MWe NPP, whose characteristic is 17 × 17 9.5 fuel assembly. Since the beginning of 1980s, NPIC had proceeded with CHF experimental researches of 3×3 and 4×4 9.5 rod bundles with spacer grids. The main objectives of the experiments were as follows:
Where A2 , B2 , C2 , D2 , E2 , F2 , G2 are experimental coefficients, respectively. Relative errors of 95% experimental data were less than ±12% (Fig. 6), whose root mean square error is equal to 6.9%. (2) For rod bundles with a coldwall thimble cell, 68 sets of experimental data at pressure of 14.7 MPa have been obtained. CHF correlation is as follows: QCHF = A3 {B3 + C3 Wg − (D3 + E3 Wg )Xc }
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2
2
Qc(MW/m )
2.5
+12%
1.5
- 12%
1
0.5
0 0
0.5
1
1.5
2
2.5
3
2
Qe(MW/m )
Fig. 6. CHF comparison between experiment and calculation.
Where A3 , B3 , C3 , D3 , E3 are experimental coefficients, respectively. Relative errors of 95% experimental data were less than ±10% (Fig. 7), whose root mean square error is equal to 5.8%. (3) By comparing coldwall rod bundles and typical rod bundles, coldwall effect brings CHF to decrease by 2%.
(1) Two hundred and forty-one sets of experimental data have been obtained. CHF correlation is as follows: QCHF = {A4 + B4 Wg − (C4 + D4 Wg )Xc }
3.2. CHF experiment of AC600 under low mass flow rate
× (E4 − F4 P)
Though CHF correlation reliable to normal operating conditions of Qinshan II 600 MWe NPP, it has a 3
2. 5 +10%
2
2
Qc (MW/m )
relatively narrow range of validity. The correct characterization of CHF under low flow condition is of particular interest when predicting the reactor core behavior during a loss of coolant accident or steam line break accident without offsite power in PWR (Kim et al., 2000; Chun et al., 2001). In order to achieve an optimal design and to ensure a high degree of safety in AC600, CHF experiment under low mass flow rate was carried out. The experiments were carried out in small-scale thermal-hydraulic test facility of NPIC. The test section consists of sixteen 9.5 mm diameter heater rods with heated length of 1000 mm, which simulates a typical cell in a 4 × 4 array with a rod pitch of 12.6 mm, using a mixing vane spacer grid on a 270 mm spacing, with a simple support grid at the topmost of rod bundles. Distance between the topmost spacer grid and the end of heated section is 20 mm. The main experimental parameters are: pressure of 9.8–15.88 MPa, mass velocity of 278–1389 kg/m2 s, quality of −0.20 to 0.34. Based on experimental data, the following conclusions can be drawn:
- 10%
1. 5
Where A4 , B4 , C4 , D4 , E4 , F4 are experimental coefficients, respectively. Relative errors of 95% experimental data were less than ±10% (Fig. 8), whose standard deviation is equal to 5.81%. (2) A calculated DNB ratio (DNBR) value greater than this design limit provides assurance that there is at least a 95% probability at the 95% confidence level that CHF will not occur. DNBR is equal to 1.10.
1
4. CHF experimental progress of non-uniform heated rod bundles
0. 5
0 0
0. 5
1
1. 5
2
2. 5
3
4.1. Test facility
2
Qe (MW/m )
Fig. 7. CHF comparison between experiment and calculation.
At the beginning of 1970s, small-scale thermalhydraulic test facility had been constructed in NPIC. In
X. Zejun et al. / Nuclear Engineering and Design 229 (2004) 213–222 Table 2 Main parameters of Freon thermal-hydraulic test facility
3
2. 5
+10%
2
2
Qc (MW/m )
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- 10% 1. 5
1
Parameter
Value
Working medium Design pressure (MPa) Design temperature (◦ C) Operating pressure (MPa) Operating temperature (◦ C) Maximum power (kW) Maximum flow rate (kg/s) Pump head (MPa)
Freon 12 4.0 120 2.77 60 500 6.82 1
0. 5
0 0
0. 5
1
1. 5
2
2. 5
3
2
Qe (MW/m )
Fig. 8. CHF comparison between experiment and calculation.
order to develop CHF experimental researches of full length non-uniform heated rod bundles, NPIC have constructed Freon thermal-hydraulic test facility and large-scale thermal-hydraulic test facility. 4.1.1. Freon thermal-hydraulic test facility Freon thermal-hydraulic test facility is composed of main loop system, test section branches, purification system, refrigeration system, makeup and discharge system, pressurization system, vacuumized system
and safety relief system, etc. The main components include refrigeration equipment, preheater, mixing condenser, heat exchanger, pressurizer, canned pump, vacuum pump. Fig. 9 shows its schematic diagram. The main parameters are shown in Table 2. 4.1.2. Large-scale thermal-hydraulic test facility Large-scale thermal-hydraulic test facility is composed of a high-temperature high-pressure water test loop and a large heating resource. The loop consists of main loop system, test section branches, water purification system, makeup water system, pressurization system and safety relief system, etc. The main components include pump, heat exchanger, pressurizer, storage tank, mixing condenser, preheater. Fig. 10 shows its schematic diagram. The main parameters are shown in Table 3.
Fig. 9. Schematic diagram of Freon thermal-hydraulic test facility.
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Fig. 10. Schematic diagram of large-scale thermal-hydraulic test facility.
4.2. Research review The prediction of CHF has been, in most cases, based on the empirical correlation. For PWR fuel assembly, the local parameter correlation requires the local thermal-hydraulic conditions usually calculated by Table 3 Main parameters of large-scale thermal-hydraulic test facility Parameter
Value
Working medium
Deionized water 20 370 17.5 355 500 80 1 220 300 70/15 10 (dc 250 V, 40000 A) 453
Design pressure (MPa) Design temperature (◦ C) Operating pressure (MPa) Operating temperature (◦ C) Volume flow rate of large pump (m3 /h) Volume flow rate of small pump (m3 /h) Pump head (MPa) Volume flow rate of the main loop (m3 /h) Maximum temperature of the main loop (◦ C) Volume flow rate of test branch (m3 /h) Power of test section (MW) Volume flow rate of cooling water (m3 /h)
a subchannel analysis code. The cross-sectional averaged fluid conditions of subchannel, however, are not sufficient for determining CHF, especially for the case of non-uniform axial heat flux distribution (Hwang et al., 2001). Many investigators have studied on the influence of upstream history of heat flux variation to CHF. But in China, because of some technical reasons, all CHF engineering experiments, were almost done through part length uniform heated rod bundles in the past, data of which were disposed through cross-sectional mean parameter on the basis of ‘local conditions’ hypothesis. The hypothesis assumes that there is a unique relationship between CHF and local thermal-hydraulic conditions. In reality, axial non-uniform power distribution of heated rod bundles has an important effect on critical point, CHF conditions and CHF value under any circumstances, which is close to axial power distribution of NPP fuel assemblies (Zejun et al., 2003). In the early of 1990s, NPIC started to investigate CHF experimental technique of full length non-uniform heated rod bundles, and successively proceeded with CHF research of single passage non-uniform heated rod, CHF experimentally technical researches
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of non-uniform heated rod bundles, manufacture of non-uniform heated rods and technical researches of critical detection. Because of the technical difficulty and support fund of scientific researches, experimental research progress on CHF of non-uniform heated rod bundles became slower and slower. In the middle of 1990s, NPIC undertook experimental research project on CHF of high performance fuel assembly from Chinese government, continued to proceed with CHF experimental research of non-uniform heated rod bundles. On the basis of absorption and digestion of the advanced design technique, NPIC mastered design key technique, for example, “O” type seal technique and insulation technique, and completed design and manufacture of test section. In the early of 2000s, NPIC had developed experimental research on CHF of 5 × 5 full length non-uniform heated rod bundles. The main objectives of the research were as follows: (1) to compare with the thermal-hydraulic performance between standard fuel assembly and fuel assembly with mid span mixing grid (MSMG); (2) to verify the conformability between CHF experimental data of rod bundles with MSMG and WRB-A correlation. 4.3. Experimental progress The experiments were carried out in large-scale thermal-hydraulic test facility of NPIC. The test sections consist of two sets of rod bundles. First set of rod bundles consists of sixteen 9.5 mm diameter heater rods with heated length of 3657 mm, which simulates a typical cell in a 5 × 5 array with a rod pitch of 12.6 mm, using mixing vane spacer grids on a 521 mm spacing, with simple support grids in the middle of the mixing vane spacer grids. Second set of rod bundles is similar to first set of rod bundles, but it consists of MSMG in the middle of the mixing vane spacer grids. The MSMG (three), placed in the upper of the rod bundles, enhance the mixing capabilities of the flow, thus delaying the occurrence of a vapor blanket causing CHF. The rod bundles have a non-uniform radial power distribution with sixteen peripheral rods having less power than nine central rods. It also has a non-uniform cosine axial power distribution which is made by smoothly changing the inside diameter of rods with axial position. For central rods, the thickness of the heater tube was about 0.44 mm at the maximum
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heat flux location and about 1.99 mm at the minimum heat flux location. For peripheral rods, the thickness of the heater tube was about 0.525 mm at the maximum heat flux location and about 2.6 mm at the minimum heat flux location. Seven sets of thermocouples for each tube are located in the ceramic pieces, which are fitted to inside diameter of each tube. The main experimental parameters are: pressure of 9.8–16.3 MPa, mass velocity of 1166–2613 kg/m2 s, quality of −0.005 to 0.306. Based on experimental data, the following conclusions can be drawn: (1) By comparing with the experimental data between standard fuel assembly and fuel assembly with MSMG, it shows that thermal-hydraulic performance of fuel assembly with MSMG is better than that of standard fuel assembly. (2) By calculating the local thermal-hydraulic parameters using FLICA code, it shows that CHF experimental data of rod bundles with MSMG are conform with WRB-A correlation so much.
5. Summary Since the end of 1970s, NPIC had completed a great amount of CHF experiments. With the construction of Freon thermal-hydraulic test facility and large-scale thermal-hydraulic test facility, on the basis of improvement of design technique and experimental technique continuously, NPIC have had the top capability of developing CHF experimental research of full length non-uniform heated rod bundles, and completed CHF experimental research of high performance fuel assembly. At present, NPIC are undertaking experimental research project on CHF of CAFA with self-reliance copyright from Chinese government, and developing a large quantity of experimental researches on CHF. The main research programs are as follows: (1) While water flowing through rod bundles with different kinds of spacer grids, its velocity fields will be calculated using CFX5.5 code. By comparing with thermal-hydraulic performance among different kinds of spacer grids, three kinds of better performance of spacer grids will be collected. (2) In Freon thermal-hydraulic test facility, CHF experiments of part length uniform heated rod
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(3)
(4)
(5)
(6)
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bundles with three kinds of spacer grids will be developed to filter spacer grids. By comparing with CHF among rod bundles with three kinds of spacer grids, the best performance spacer grid will be collected. In small-scale thermal-hydraulic test facility, hydraulic experiment of spacer grids will be conducted to measure form resistance coefficient of structural grid and MSMG. In large-scale thermal-hydraulic test facility, subchannel mixing experiment of rod bundles will be developed to measure subchannel mixing coefficient. In large-scale thermal-hydraulic test facility, CHF experiments of uniform heated rod bundles will be developed. By using FLICA code to calculate local thermal-hydraulic parameters, CHF correlation of uniform heated rod bundles will be got. In large-scale thermal-hydraulic test facility, CHF experiment of full length non-uniform heated rod bundles will be developed. By using FLICA code to calculate local thermal-hydraulic parameters, CHF correlation of full length non-uniform heated rod bundles will be got. By comparing with CHF between CAFA and the other advanced fuel assemblies, better thermal-hydraulic performance of CAFA will be verified.
All CHF data obtained will be used for design improvement of Chinese PWR and R&D program of New Generation 1000 MWe PWR. References Cheng, X., Mˆuller, U., 1998. Critical heat flux and turbulent mixing in hexagonal tight rod bundles. Int. J. Multiphase Flow 24, 1245–1263. Chun, S.-Y., Chung, H.-J., Moon, S.-K., Yang, S.-K., Chung, M.-K., Schoesse, T., Aritomi, M., 2001. Effect of pressure on critical heat flux in uniformly heated vertical annulus under low flow conditions. Nucl. Eng. Des. 203, 159–174. Hoyer, N., 1998. Calculation of dryout and post-dryout heat transfer for tube geometry. Int. J. Multiphase Flow 24 (2), 319– 334. Hwang, D.-H., Park, C., Zee, S.-Q., 2001. A phenomenological approach to correcting DNB-type critical heat flux for non-uniform axial power shapes. Int. J. Heat Mass Transfer 44, 4483–4492. Katto, Y., 1994. Critical heat flux. Int. J. Multiphase Flow 20, 53–90. Kim, H.C., Baek, W.-P., Chang, S.H., 2000. Critical heat flux water in vertical round tubes at low pressure and low flow conditions. Nucl. Eng. Des. 199, 49–73. Nagino, Y., Shodai, N., Matsumoto, C., Ueji, M., 1978. Rod bowed to contact departure from nucleate boiling tests in coldwall thimble cell geometry. J. Nucl. Sci. Technol. 15 (8), 568– 573. Zejun, X., Bingde, C., Zhongpeng, L., Dounan, J., 2003. Research progress of experiment on critical heat flux. Nucl. Power Eng. 24 (1), 24–27 (in Chinese).