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An experimental study on the validation of cooling capability for the Passive Auxiliary Feedwater System (PAFS) condensation heat exchanger
Nuclear Engineering and Design 260 (2013) 54–63
Contents lists available at SciVerse ScienceDirect
Nuclear Engineering and Design journal homepage: www.elsevier.com/locate/nucengdes
An experimental study on the validation of cooling capability for the Passive Auxiliary Feedwater System (PAFS) condensation heat exchanger Seok Kim a , Byoung-Uhn Bae a , Yun-Je Cho a , Yu-Sun Park a , Kyoung-Ho Kang a , Byong-Jo Yun b,∗ a b
Thermal Hydraulics Safety Research Division, Korea Atomic Energy Research Institute, 1045 Daedeok-daero, Yuseong-gu, Daejeon, 305-353, Republic of Korea School of Mechanical Engineering, Pusan National University, 30 Jangjeon-dong, Geumjeong-gu, Busan, 609-735, Republic of Korea
h i g h l i g h t s • • • •
PAFS is designed to replace a conventional active Auxiliary Feedwater System. A SET facility is constructed for investigating the thermal-hydraulic behavior of the PAFS system. Experimental results proved that the PCHX design satisfied the heat removal requirements. Results of the MARS-KS code provided a conservative prediction of the heat transfer phenomena.
a r t i c l e
i n f o
Article history: Received 4 May 2012 Received in revised form 25 February 2013 Accepted 13 March 2013
a b s t r a c t The Passive Auxiliary Feedwater System (PAFS) is one of the advanced safety features adopted in the Advanced Power Reactor Plus (APR+). PAFS is designed to replace a conventional active Auxiliary Feedwater System (AFWS). The PAFS cools down the steam generator secondary side and eventually removes the decay heat from the reactor core by a natural circulation mechanism, i.e., condensing steam in nearly horizontal U-tubes submerged inside a pool. A separate effect test facility was constructed with the aim of validating the cooling and operational performance of the PAFS. The PAFS Condensing Heat Removal Assessment Loop (PASCAL) was constructed by simulating a single Passive Condensation Heat Exchanger (PCHX) tube submerged in the Passive Condensation Cooling Tank (PCCT) according to the volumetric scaling methodology. Quasi-steady state (SS) test cases and PCCT level decrease (PL) were sequentially performed with the steam generator heater power set at 540 kW to investigate the thermal-hydraulic behavior of the PAFS system and the characteristics of the natural circulation in the loop. The experimental results proved that the current PCHX design satisfied the heat removal requirement for cooling down the reactor core during an accident condition. Therefore, the PAFS can replace a conventional active AFWS in the APR+ by utilizing the two-phase natural circulation flow. The Multi-dimensional Analysis of Reactor Safety, KINS Standard Version (MARS-KS), a thermal hydraulic system analysis code, was utilized to validate the present experimental data. The results of the MARS-KS code provided a conservative prediction of the heat transfer phenomena for the PCHX cooling performance. © 2013 Elsevier B.V. All rights reserved.
1. Introduction The Advanced Power Reactor Plus (APR+) is a Generation III+ nuclear power plant being developed in Korea. The Passive Auxiliary Feedwater System (PAFS) is one of the advanced safety features being adopted in the APR+. PAFS is intended to completely replace the conventional active auxiliary feedwater system (Song et al., 2010; Cheon et al., 2010). PAFS can improve the reliability of the safety system and reduce operator error, which are the fundamental weak points outlined in the Probability Safety Assessment (PSA).
∗ Corresponding author. Tel.: +82 51 510 2484. E-mail address: [email protected] (B.-J. Yun). 0029-5493/$ – see front matter © 2013 Elsevier B.V. All rights reserved. http://dx.doi.org/10.1016/j.nucengdes.2013.03.016
The PAFS cools down the secondary side of the steam generator (SG) and eventually removes the decay heat from the reactor core by adopting a natural circulation mechanism, i.e., condensing steam in the nearly-horizontal Passive Condensation Heat Exchanger (PCHX) tubes submerged inside the PCCT. When the water level in the steam generator becomes lower than 25% of the wide range of the water level transmitter during an accident situation, the actuation valve at the return-water line opens and then the PAFS begins its natural circulation flow. To satisfy a single failure criterion, the PAFS is composed of two independent trains, but a single PAFS train is capable of removing the whole decay heat from the reactor core during the anticipated accident transients. Fig. 1 shows a schematic diagram of the PAFS for the APR+. The diagram comprises a steamsupply line, a PCHX, a return-water line, and a passive condensation
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Nomenclature h k L ˙ m q r
enthalpy [J/kg] thermal conductivity [W/mK] length [m] mass flow rate [kg/s] heat flux [W/m2 ] radius [m]
cooling tank (PCCT), as shown in Figs. 2 and 3. The PCHX comprises 4 bundles and 240 tubes with a nearly-horizontal U-tube shape not only to fulfill the heat removal requirement but also to prevent the occurrence of a water hammer inside the PCHX (Bae et al., 2011). The PCCT was sized and designed as a heat sink, evaporating the water pool for 8 h after a PAFS actuation (Bae et al., 2012). The experimental separate effect test program is in progress at the Korea Atomic Energy Research Institute (KAERI). The program aims to validate the cooling and operational performance of the PAFS. The PAFS Condensing Heat Removal Assessment Loop test facility (PASCAL) was constructed to experimentally investigate the condensation heat transfer and natural circulation phenomena in the PAFS (Kang et al., 2012; Yun et al., 2010). This study utilized a single, nearly-horizontal PCHX tube that had the same dimensions and design material as the APR+ PAFS prototype. In this experiment, the major thermal-hydraulic parameters (e.g., the local or overall heat transfer coefficients, the fluid temperature inside the tube, the wall temperature of the tube, and the pool temperature distribution in the PCCT) were measured to evaluate the current condensation heat transfer model and also to provide a database for validating the calculation performance of the PAFS safety analysis codes. A single PAFS train is required to remove a maximum heat removal rate of 129.8 MW by 240 PCHX tubes. Therefore, 540 kW of thermal power was supplied to generate a steam flow into the single tube of the PASCAL facility according to the facility’s scaling ratio (1/240). There are two PAFS trains for two SGs in the APR+, but to satisfy the single failure criterion, a single PAFS train should be designed to remove the total decay heat during an accident.
Fig. 2. The design of a PCHX bundle in APR+ PAFS.
Therefore, in this study, the condensation heat exchanger for PASCAL has a nominal heat removal capacity of 540 kW. In this study, two test cases were performed to validate the PAFS cooling performance. The quasi-steady state condition of the 540 kW thermal power was simulated according to volumetric scaling methodology. With the thermal power generated by electrical heaters in the steam generator, a heat removal rate in the PCHX was measured, and the characteristics of the natural loop circulation were investigated. In the experiment, a quasi-steady state condition of the system was simulated when a heat balance between the steam generator and the PCHX was achieved. After the quasi-steady state condition test, the PCCT water level test was performed with the goal of investigating thermal-hydraulic behavior during a water level decrease in the PCCT. After reaching the predetermined goals of the thermal power in the steam generator and the heat removal rate in the PCHX, the characteristics of the natural circulation in the loop were measured. A MARS-KS (Multi-dimensional Analysis for Reactor Safety, KINS Standard Version) code analysis was conducted to validate the experimental data. The results will contribute
Fig. 1. A schematic diagram of APR+ PAFS.
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Fig. 3. The design of APR+ PCCT.
to the improvement of the condensation and boiling heat transfer models.
2. Description of the PASCAL facility The PASCAL facility was designed according to a volumetric scaling methodology (Nahavandi et al., 1979). The methodology preserves the elevation change between a heat source and a heat sink in a natural circulation loop under the same pressure and temperature conditions. Before conducting the experiments, the scaling methodology applied in the design of the PASCAL facility was validated by the MARS-KS code analysis in a previous study (Cho et al., 2012). Fig. 4 shows the schematic diagram of the PASCAL facility. A steam generator supplies saturated steam to the PCHX tube. Electrical heaters in the steam generator provided the heat source that scaled down the heat transfer rate at the U-tube surface. To preserve the driving force of the natural circulation in the loop, a distance equivalent to that of the prototype was maintained between the mixture level in the steam generator and the PCHX tube was maintained. The steam generator was connected to the PCHX tube with a steam-supply line and a return-water line. To determine the PCHX design parameters for the PAFS, the requirements for the heat exchanger must be satisfied. The PAFS condensation heat exchanger can remove the decay heat from the reactor core during accidents. Therefore, the heat removal rate has to balance
the anticipated decay heat level obtained from the safety analysis of the APR+. Another important requirement is preventing the condensation-induced water hammer inside the tube. Particularly, a horizontal heat exchanger was considered for the PAFS because the heat exchanger should be designed to prevent the water hammer phenomenon to ensure structural integrity. In this experimental study, the PCHX of the PASCAL facility simulates a single tube among 240 tubes in the prototype; hence, the volumetric scaling ratio of the facility is 1/240. The PCCT pool volume was also reduced to 1/240 of the prototype. From the top view of the PCCT as shown in Fig. 3, total 4 bundles were immerged in the PCCT. One bundle consisted of 60 tubes and 120 tubes are located in the half side. Then, the length of the PCCT was scaled by 1/2 and the width of that was scaled by 1/120. The length, the width, and height of the PCCT are 6.7 m, 0.112 m, and 11.484 m, respectively. This aspect ratio was designed to preserve the natural convection flow inside the PCCT. Table 1 compares the major design and scaling parameters of the PASCAL facility with those of the prototype. The PCCT of the PASCAL facility was designed before the prototype PCCT was changed so a discrepancy appears in the PCCT length and the width, as shown in Table 1. The increase in the PCCT length and width resulted in a conservative prediction for the cooling performance of the PCHX due to the weakness of the natural circulation. In Table 1, the PCCT height means the minimum water level of the PCCT as the design requirements of the PAFS. The MARS-KS code analysis is performed to verify the influence of the water level. The influence of the PCCT water level is negligible as shown in Fig. 5.
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Table 1 The geometry and scaling parameters of the PAFS and the PASCAL facility. Parameter
APR+
PASCAL
Ratio
PCHX tube
Inner/outer diameter Length Number of tubes Material Operating condition
44.8 mm/50.8 mm 8.4 m 240 Stainless steel 304 7.4 MPa 290 ◦ C
44.8 mm/50.8 mm 8.4 m 1 Stainless Steel 304 7.4 MPa 290 ◦ C
1/1 1/1 1/240 – –
PCCT
Pool height Pool length Pool width Elevation from SG water level
7.62 m (Previously 8.9 m) 18.29 m (Previously 13.4 m) 13.56 m (Previously 13.4 m) 16.2 m
8.9 m 6.7 m 0.112 m 16.2 m
1/0.86 (previously 1/1) 1/2.7 (previously 1/2) 1/121 (previously 1/120) 1/1
In the PASCAL test, the major measuring parameters are the flow rate of the steam and condensate liquid, the loop temperature and pressure, the differential pressure on the steam-supply line and the return-water line. Additionally, the collapsed water level and the liquid temperature were measured in the PCCT pool. To induce the local distribution of the heat flux and the heat transfer coefficient of the PCHX, the surface temperatures at the inside
Fig. 5. The influence of the PCCT water level (MARS analysis).
and the outside walls were measured at eleven points along the tube length (as shown in Fig. 6). At each point, the fluid temperature profile inside the tube was measured in a vertical direction so that the distribution of condensate liquid flow inside the PCHX tube could be inferred. Uncertainty (of the measured experimental data) was analyzed in accordance with a 95% confidence level. According to the ASME performance test code 19.1, the uncertainty level of the present results was estimated by an unsystematic rootmean-square random error and a systematic error (Kim, 2011). The random and systematic errors were evaluated from the data acquisition hardware specifications and the calibration results performed once every year. Table 2 shows the analyzed uncertainty levels of each group of instruments. 3. Experimental conditions and procedure 3.1. Experimental conditions The nominal experimental condition was determined according to the PAFS maximum heat removal requirement. Because a single Table 2 Instrument uncertainty levels.
Fig. 4. A schematic diagram of the PASCAL.
Items
Unit
Uncertainty
Static pressure Differential pressure Collapsed water level Temperature Flow rate Heater power Heat flux (tube, pool side) Heat transfer coefficient (tube side) Heat transfer coefficient (pool side)
MPa kPa m ◦ C kg/s W W/m2 W/m2 K W/m2 K
±0.365% ±6.7% ±4.0% Maximum 4.4 ◦ C ±0.93% ±0.595% ±15.3% ±3.1% ±2.6%
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Fig. 6. The Measurement of wall and fluid temperature.
train of the PAFS is required to remove 129.8 MW as a maximum heat removal rate by 240 tubes of the PCHX, 540 kW of thermal power was supplied to generate a steam flow into the single tube of the PASCAL facility according to the scaling ratio of the facility (1/240). These test cases were regarded as the SS/PL-540-P1 test, where SS/PL means a combined experiment of the continuous simulation for a quasi-steady state condition (SS) and a decrease of the PCCT water level (PL). The quasi-steady state condition focused on characteristics of the natural circulation flow when the heat balance between the steam generator and the PCHX was achieved at a nominal PCCT water level.
3.2. Experimental procedure The overall procedure, including the SS and the PL tests, is listed below. Step 1: Pre-heating and flashing of the steam generator system After ventilating the non-condensable gas, the steam generator, the steam-supply line, the PCHX tube, and the return-water line were filled with water. The electrical heater of the steam generator heated the system until the fluid temperature reached to 180 ◦ C with a forced convection flow driven by a circulation pump in the loop. The system pressure was maintained higher than a saturation pressure of the fluid system to prevent boiling inside the steam generator. After the fluid system temperature reached 180 ◦ C, the forced convection flow was stopped. Then the steam generator system vent valve was opened to the atmosphere to decrease an inventory and make a mixture level in the steam generator 1.2 m by flashing the steam flow. Step 2: Starting the data acquisition and achieving the quasisteady state condition
When the collapsed water level in the steam generator was reduced by 1.2 m, the vent valve of the steam generator system was closed to stop flashing. That closure made a natural circulation flow of the steam in the loop. Data acquisition began at that moment. The thermal power of the electrical heaters in the steam generator was gradually increased to the thermal power of the experimental condition, while maintaining an increasing rate of the system temperature as 1 K/min. The water in the PCCT reached a saturated state at the atmospheric pressure via the PCHX heat transfer. The evaporation of the PCCT pool water decreased the water level to 9.3 m, which is the nominal water level at the saturated state. Because the initial PCCT water level in the prototype is 8.9 m at room temperature, the moment that the PCCT water level reached 9.3 m at the saturated condition (due to a reduced density) was considered to be a quasi-steady state. When the pressure, temperature, and flow rate reached a quasi-steady state at the constant thermal power condition, the heat removal rate and the natural circulation flow provided the data for the SS test. Step 3: Decreasing the PCCT water level to 3.5 m The continuous supply of the thermal power in the steam generator made the PCCT water level decrease by evaporation. The experimental data measured during the decrease of the PCCT water level provided the data for the PL test. The data acquisition continued until the PCCT water level decreased to 3.5 m, which was near the top of the PCHX. 4. Experimental results and MARS-KS code analysis 4.1. Experimental results of the SS/PL-540-P1 test Table 3 summarizes the overall sequence of events for the SS/PL-540-P1 test according to the test procedure. From the data acquisition time of 0–4200 s, the thermal power of the electrical
Table 3 The sequence of events for the SS/PL-540-P1 test. Phase
Time
Sequence
Start of data acquisition
0s
Heat-up phase Transient phase
0–4200 s 4200–13,200 s
Data of SS-540-P1 case Data of PL-540-P1 case
13,200–13,300 s 13,300–32,600 s
Flashing is stopped after maintaining the initial SG system inventory. Natural circulation flow is generated in the loop Stepwise increase of SG heater power up to 540 kW Time period until the system pressure and temperature reached a steady-state condition and the PCCT water level was 9.3 m Quasi-steady state condition at the continuous thermal power Decrease of PCCT water level
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Fig. 7. The heat removal rate of the SS/PL-540-P1 test.
heaters in the SG increased incrementally by 1 K/min to preserve the system temperature. After the heater power reached 540 kW, the pressure and the temperature of the SG system increased until a heat balance was achieved between the heat source (the heater) and the PCHX heat removal. At approximately 8000 s, the water level of the PCCT began decreasing due to the evaporation caused by the boiling heat transfer at an outer wall of the PCHX. Experimental data from the system pressure, the mass flow rate, and the loop temperature for the quasi-steady state condition were gathered during the period from 13,200 s to 13,300 s. This data were used to perform the SS-540-P1 test. When the water level in the PCCT decreased to 3.5 m, the experimental data acquired after the steady-state condition was used for the PL-540-P1 test. Figs. 7–10 presented the heat removal rates, the steam generator system pressure, the steam generator system temperature, and the mass flow rate for the SS/PL-540-01 test. In the PASCAL test, the heat removal rates are listed in Table 4, which includes the supplied thermal power of electrical heaters in steam generator (HR SG), the summation of heat transfer rate of the PCHX tube surface (HR Tube), the enthalpy decrease between the inlet and the outlet of the PCHX (HR SS and HR RW). Among the calculated heat removal rates, HR Tube was considered to be the most reliable value due to an integration of the local heat transfer rate along the PCHX tube length. When the thermal power of steam generator was small in the heat-up phase, a natural circulation flow rate was
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Fig. 9. The steam generator system temperature of the SS/PL-540-P1 test.
Fig. 10. The steam generator system flow rate of the SS/PL-540-P1 test. Table 4 The definition of heat removal rate. Notation
Description
HR SG
The thermal power supplied by the electrical heaters in SG
HR Tube
Heat removal rate of PCHX calculated from summation of wall heat transfer rate HR Tube =
ip
" 2r L qavg,o o ip
" " " qavg,o = qtop,o + qbottom,o /2 kTube (Tw,i −Tw,o ) k (T −Tw,o )bottom top " " , qbottom,o = Tube w,i qtop,o = ro ln(ro /ri ) ro ln(ro /ri ) ri : Inner radius of tube r0 : Outer radius of tube Lip : The length of the tube segment (ip is the measurement position number from 1 to 11)
Fig. 8. The steam generator system pressure of the SS/PL-540-P1 test.
HR SS
The heat removal rate calculated from the flow rate in the steam-supply line ˙ ss (hin − hout ) HRSS = m ˙ ss : mass flow rate of the steam-supply line m hin : fluid enthalpy of the PCHX inlet hout : Fluid enthalpy of the PCHX outlet
HR RW
The heat removal rate calculated from the flow rate in return-water line ˙ RW (hin − hout ) HRRW = m ˙ RW : Mass flow rate of the return-water line m hin and hout : Same to those in HR SS
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Fig. 11. The wall temperature of the PCHX tube.
small so that a flow oscillation in the return-water line occurred and it affected the oscillation of HR RW calculation as shown in Fig. 7. While the thermal power of electrical heaters in steam generator increased stepwise in the heat-up phase, the heat removal rate of PCHX also increased following the supplied thermal power. During the heat-up phase, the difference between the HR SG and the HR Tube in the transient phase is the transient conduction heat transfer of the PCHX tube. Because the supplied thermal power was larger than the removed heat transfer rate at PCHX during these periods, the pressure and temperature of the steam generator system increased continuously, as shown in Figs. 8 and 9. During the quasi-steady state condition from 13,200 to 13,300 s, the heat removal rate of the PCHX tube was equivalent to the supplied power. At the quasi-steady state, the pressure of the steam generator system reached 3.22 MPa and the steam temperature was 242 ◦ C. The PCHX of the prototype PAFS was designed to remove the decay heat at the steam condition of 7.4 MPa and 290 ◦ C. The experimental result for the quasi-steady state condition indicated that the current design of the PCHX had enough of a thermal margin to cool the reactor core without an active safety injection system. Fig. 10 shows the natural circulation flow rate, where the mass flow rate measured in the steam-supply line and the return-water line presented similar behavior during the quasi-steady state, which means that the current PAFS design can cool the secondary system of the steam generator stably in a passive manner. After the quasisteady state, the heat removal rate increased due to an increase of boiling heat transfer in the PCCT water pool. Therefore, the pressure, the temperature, and the natural circulation flow rate of the steam generator system decreased, as shown in Figs. 8–10. Figs. 11–14 show variations in the surface temperature, the heat flux, the heat transfer coefficient and the fluid temperature along the tube length during the quasi-steady state condition. As the steam flowed toward the outlet, the heat flux decreased due to the lower wall temperature, which was induced by the subcooled liquid flow. The condensation heat transfer coefficient at the top region of the inner wall was larger than that of bottom region because the top part of the tube was filled with the steam flow and the condensate liquid flowed in the bottom region inside the tube. The phase distribution could be interpreted by the fluid temperature measurement shown in Fig. 14, which was also plotted as a contour in Fig. 15. The temperature at point E (the bottom region as shown in Fig. 5) was lower than the saturation temperature along the whole tube length. Near the outlet of the PCHX tube, the fluid temperature at point D also revealed the subcooled water (by an increase of the liquid fraction). The round region of the tube tends to
Fig. 12. The heat flux of the PCHX tube.
Fig. 13. The heat transfer coefficient of the PCHX tube.
mix the condensate liquid flow in nature, so that the fluid temperature showed a uniform profile. From the distribution of the fluid temperature inside the PCHX tube, it was confined that a stratified flow appeared along the whole length of the tube without a bubbly flow or a slug flow. The NUREG report suggested several
Fig. 14. The fluid temperature in the PCHX tube.
S. Kim et al. / Nuclear Engineering and Design 260 (2013) 54–63
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Fig. 15. The fluid temperature contour inside the tube.
guidelines for inhibiting the water hammer in a horizontal channel (Griffith, 1997). Excluding slug flow is one of the requirements, which prevented a condensation-induced water hammering. The experimental results are in agreement with the criterion for inhibiting a condensation-induced water hammer phenomenon inside the tube. Fig. 16 shows a variation of the PCCT water level in the SS/PL540-P1 test. As the pool temperature increased during the heat-up phase, the water level increased due to the volume expansion of the PCCT pool. When the pool water reached a saturated state, the collapsed water level in the PCCT began to continuously decrease. The lower water level of the PCCT means the lower static pressure and the lower saturation temperature of the pool water around the PCHX tube, so that a decrease of the PCCT water level resulted in much of the nucleate boiling on the tube surface. The heat transfer coefficients at points 1 and 11 (as shown in Fig. 6) were presented in Fig. 17. In this study, the boiling heat transfer coefficient at the PCHX tube surface increased in response to the decrease of the PCCT water level, while the condensation heat transfer coefficient showed a uniform behavior with the variation of the PCCT water level. Fig. 18 compares a collapsed water level of the steam generator with the return-water line. During the period from heat-up until 4200 s, the water level difference between the SG and the returnwater line increased due to an increase of the supplied thermal power and a drop in pressure. After the quasi-steady state condition was achieved, the water level of the return-water line showed increased slightly due to the generation of a larger condensate liquid flow in the loop. Due to level difference between the steam generator and return-water line, condensate water can be returned back to the SG with a natural driving force.
Considering that the heat exchanger of the PAFS was designed to remove the decay heat of APR+, the experimental result proved that the current design of the PCHX had enough of a margin to cool down the reactor system without any active safety injection system. The mass flow rate measured on the steam-supply line and the return-water line presented similar behavior in the steady state. This indicates that the current PAFS design can supply the auxiliary feedwater and passively but effectively cool down the secondary steam generator system.
Fig. 16. The collapsed water level of PCCT.
Fig. 18. The water level in the loop system.
Fig. 17. The heat transfer coefficient during the PCCT water level decrease.
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Fig. 19. A nodalization diagram for the PASCAL facility.
4.2. MARS-KS code analysis The experimental data of the quasi-steady state condition, SS540-P1 test case, was utilized to validate the prediction capability of the MARS-KS code. The MARS-KS code is a best-estimate system analysis code based on the two-fluid model (Cho et al., 2012; Jeong et al., 1999). The MARS-KS code has the capability of analyzing a one-dimensional or a three-dimensional best-estimated thermal hydraulic system and the fuel responses of the light-water reactor transients. It treats the two-phase flow phenomena with six equations for the transfer of mass, momentum, and energy. The MARS-KS code models the condensation heat transfer coefficient as the maximum of the values yielded by the Nusselt’s model and
Shah’s model (Schaffrath et al., 1999). The geometry of the PASCAL facility was modeled as shown in Fig. 19. The PCHX tube was modeled as a single pipe with heat structures and divided by 13 volumes along the length. The PCCT pool was modeled using the multi-D component in the MARS-KS code to effectively simulate a multi-dimensional natural convection flow in the rectangular pool. To simulate the experimental condition of the SS-540-P1 test, the heater power in the steam generator was supplied with 540 kW. The initial condition of a coolant inventory and a system pressure in the steam generator and the PCCT was maintained equivalent to that of the SS-540-P1 test. In the MARS-KS code analysis, a PCCT pool water level decreased to 9.3 m by evaporation at approximately 2250 s (as shown in Fig. 20). The time difference for when PCCT pool water level was 9.3 m was caused by the experimental procedure of increasing the heater power. Fig. 21 presents the
Fig. 20. The MARS-KS calculation of the PCCT water level.
Fig. 21. The MARS-KS calculation of steam generator pressure.
S. Kim et al. / Nuclear Engineering and Design 260 (2013) 54–63
analysis result for the steam generator pressure. Compared to the experimental result at the quasi-steady state condition, the original MARS-KS code overestimates the system pressure in the quasisteady state condition. This overestimation means that the code underestimated a heat transfer rate at the PCHX, which proves that the analysis result of the MARS-KS code is sufficiently conservative in calculating the amount of condensation heat transfer in the PCHX of the PAFS. The use of condensation heat transfer coefficient measured in the experiment, however, dramatically improves the calculation result. There is a need to improve the condensation and boiling heat transfer model in MARS-KS code. 5. Conclusion This study presents the experimental results from the PASCAL facility. The PASCAL facility was constructed to validate the cooling performance of the PAFS and to investigate the condensation heat transfer and natural circulation phenomena in the PAFS. In this study, test cases of SS-540-P1 and PL-540-P1 were performed. The objective of the SS-540-P1 test is to investigate the cooling performance and natural circulation characteristics of the PAFS by simulating a quasi-steady state condition of the thermal power according to the volumetric scaling methodology. The PL-540-P1 test was conducted after the SS-540-P1 test, when the collapsed water level in the PCCT was continuously decreasing from evaporation. The test results proved that the current design of the PCHX satisfied the heat removal requirement for cooling down the reactor core during an accident condition. Therefore, it can be preliminarily concluded that the PAFS can replace a conventional active auxiliary feedwater system in the APR+, utilizing the natural circulation of two-phase flow. However, a simulation of SS-540-P1 test by the MARS-KS code showed a conservative result with respect to the heat removal rate of the horizontal PCHX. The present experimental results will contribute toward improving the model of the condensation and boiling heat transfer and also toward providing
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the benchmark data for validating the calculation performance of a thermal hydraulic system analysis code with respect to the PAFS. Acknowledgment The authors would like to acknowledge the financial support provided by the Ministry of Knowledge and Economy of the Korean government. References Bae, B.U., et al., 2011. Design of horizontal condensation heat exchanger for PAFS (Passive Auxiliary Feedwater System) in APR+. In: International Conference of Multiphase Flow in Industrial Plants, September 21–23, Ischia, Italy. Bae, B.U., et al., 2012. Design of condensation heat exchanger for the PAFS (Passive Auxiliary Feedwater System) of APR+ (Advanced Power Reactor Plus). Ann. Nucl. Energy 46, 134–143. Cheon, J., et al., 2010. The development of a passive auxiliary feedwater system in APR+ track 1: water-cooled reactor programs & issues. In: Proceedings of ICAPP’10, June 13–17, San Diego, CA, USA. Cho, Y.J., et al., 2012. Analytical studies of the heat removal capability of a passive auxiliary feedwater system (PAFS). Nucl. Eng. Des. 248, 306–316. Griffith, B., 1997. Screening Reactor Steam/Water Piping Systems for Water Hammer. NUREG/CR-6519. Jeong, J.J., Ha, K.S., Chung, B.D., Lee, W.J., 1999. Development of a multi-dimensional thermal-hydraulic system code, MARS 1.3.1. Ann. Nucl. Energy 26 (18), 16111642. Kang, K.H., et al., 2012. Separate and integral effect tests for validation of cooling and operational performance of the APR+ passive auxiliary feedwater system. Nucl. Eng. Technol. 44 (6). Kim, S., 2011. Data Analysis Methodology of Separate Effect Test Facility for Passive Auxiliary Feedwater System. APR+-PAFS-DD-01. Nahavandi, A.N., et al., 1979. Scaling laws for modeling nuclear reactor systems. Nucl. Sci. Eng. 72, 75-83. Schaffrath, A., Hicken, E.F., Jaegers, H., Prasser, H.M., 1999. Experimental and analytical investigation of the operation mode of the emergency condenser of the SWR1000”. Nucl. Technol. 126, 123–142. Song, C.H., et al., 2010. Thermal-hydraulic R&Ds for the APR+ developments in Korea. In: Proceedings of the 18th International Conference on Nuclear Engineering, May 17–21, Xi’an, China. Yun, B.J. et al., 2010. Construction Report of Separate Effect Test Facility for PAFS, KAERI/TR-4085/2010.
Contents lists available at SciVerse ScienceDirect
Nuclear Engineering and Design journal homepage: www.elsevier.com/locate/nucengdes
An experimental study on the validation of cooling capability for the Passive Auxiliary Feedwater System (PAFS) condensation heat exchanger Seok Kim a , Byoung-Uhn Bae a , Yun-Je Cho a , Yu-Sun Park a , Kyoung-Ho Kang a , Byong-Jo Yun b,∗ a b
Thermal Hydraulics Safety Research Division, Korea Atomic Energy Research Institute, 1045 Daedeok-daero, Yuseong-gu, Daejeon, 305-353, Republic of Korea School of Mechanical Engineering, Pusan National University, 30 Jangjeon-dong, Geumjeong-gu, Busan, 609-735, Republic of Korea
h i g h l i g h t s • • • •
PAFS is designed to replace a conventional active Auxiliary Feedwater System. A SET facility is constructed for investigating the thermal-hydraulic behavior of the PAFS system. Experimental results proved that the PCHX design satisfied the heat removal requirements. Results of the MARS-KS code provided a conservative prediction of the heat transfer phenomena.
a r t i c l e
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Article history: Received 4 May 2012 Received in revised form 25 February 2013 Accepted 13 March 2013
a b s t r a c t The Passive Auxiliary Feedwater System (PAFS) is one of the advanced safety features adopted in the Advanced Power Reactor Plus (APR+). PAFS is designed to replace a conventional active Auxiliary Feedwater System (AFWS). The PAFS cools down the steam generator secondary side and eventually removes the decay heat from the reactor core by a natural circulation mechanism, i.e., condensing steam in nearly horizontal U-tubes submerged inside a pool. A separate effect test facility was constructed with the aim of validating the cooling and operational performance of the PAFS. The PAFS Condensing Heat Removal Assessment Loop (PASCAL) was constructed by simulating a single Passive Condensation Heat Exchanger (PCHX) tube submerged in the Passive Condensation Cooling Tank (PCCT) according to the volumetric scaling methodology. Quasi-steady state (SS) test cases and PCCT level decrease (PL) were sequentially performed with the steam generator heater power set at 540 kW to investigate the thermal-hydraulic behavior of the PAFS system and the characteristics of the natural circulation in the loop. The experimental results proved that the current PCHX design satisfied the heat removal requirement for cooling down the reactor core during an accident condition. Therefore, the PAFS can replace a conventional active AFWS in the APR+ by utilizing the two-phase natural circulation flow. The Multi-dimensional Analysis of Reactor Safety, KINS Standard Version (MARS-KS), a thermal hydraulic system analysis code, was utilized to validate the present experimental data. The results of the MARS-KS code provided a conservative prediction of the heat transfer phenomena for the PCHX cooling performance. © 2013 Elsevier B.V. All rights reserved.
1. Introduction The Advanced Power Reactor Plus (APR+) is a Generation III+ nuclear power plant being developed in Korea. The Passive Auxiliary Feedwater System (PAFS) is one of the advanced safety features being adopted in the APR+. PAFS is intended to completely replace the conventional active auxiliary feedwater system (Song et al., 2010; Cheon et al., 2010). PAFS can improve the reliability of the safety system and reduce operator error, which are the fundamental weak points outlined in the Probability Safety Assessment (PSA).
∗ Corresponding author. Tel.: +82 51 510 2484. E-mail address: [email protected] (B.-J. Yun). 0029-5493/$ – see front matter © 2013 Elsevier B.V. All rights reserved. http://dx.doi.org/10.1016/j.nucengdes.2013.03.016
The PAFS cools down the secondary side of the steam generator (SG) and eventually removes the decay heat from the reactor core by adopting a natural circulation mechanism, i.e., condensing steam in the nearly-horizontal Passive Condensation Heat Exchanger (PCHX) tubes submerged inside the PCCT. When the water level in the steam generator becomes lower than 25% of the wide range of the water level transmitter during an accident situation, the actuation valve at the return-water line opens and then the PAFS begins its natural circulation flow. To satisfy a single failure criterion, the PAFS is composed of two independent trains, but a single PAFS train is capable of removing the whole decay heat from the reactor core during the anticipated accident transients. Fig. 1 shows a schematic diagram of the PAFS for the APR+. The diagram comprises a steamsupply line, a PCHX, a return-water line, and a passive condensation
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Nomenclature h k L ˙ m q r
enthalpy [J/kg] thermal conductivity [W/mK] length [m] mass flow rate [kg/s] heat flux [W/m2 ] radius [m]
cooling tank (PCCT), as shown in Figs. 2 and 3. The PCHX comprises 4 bundles and 240 tubes with a nearly-horizontal U-tube shape not only to fulfill the heat removal requirement but also to prevent the occurrence of a water hammer inside the PCHX (Bae et al., 2011). The PCCT was sized and designed as a heat sink, evaporating the water pool for 8 h after a PAFS actuation (Bae et al., 2012). The experimental separate effect test program is in progress at the Korea Atomic Energy Research Institute (KAERI). The program aims to validate the cooling and operational performance of the PAFS. The PAFS Condensing Heat Removal Assessment Loop test facility (PASCAL) was constructed to experimentally investigate the condensation heat transfer and natural circulation phenomena in the PAFS (Kang et al., 2012; Yun et al., 2010). This study utilized a single, nearly-horizontal PCHX tube that had the same dimensions and design material as the APR+ PAFS prototype. In this experiment, the major thermal-hydraulic parameters (e.g., the local or overall heat transfer coefficients, the fluid temperature inside the tube, the wall temperature of the tube, and the pool temperature distribution in the PCCT) were measured to evaluate the current condensation heat transfer model and also to provide a database for validating the calculation performance of the PAFS safety analysis codes. A single PAFS train is required to remove a maximum heat removal rate of 129.8 MW by 240 PCHX tubes. Therefore, 540 kW of thermal power was supplied to generate a steam flow into the single tube of the PASCAL facility according to the facility’s scaling ratio (1/240). There are two PAFS trains for two SGs in the APR+, but to satisfy the single failure criterion, a single PAFS train should be designed to remove the total decay heat during an accident.
Fig. 2. The design of a PCHX bundle in APR+ PAFS.
Therefore, in this study, the condensation heat exchanger for PASCAL has a nominal heat removal capacity of 540 kW. In this study, two test cases were performed to validate the PAFS cooling performance. The quasi-steady state condition of the 540 kW thermal power was simulated according to volumetric scaling methodology. With the thermal power generated by electrical heaters in the steam generator, a heat removal rate in the PCHX was measured, and the characteristics of the natural loop circulation were investigated. In the experiment, a quasi-steady state condition of the system was simulated when a heat balance between the steam generator and the PCHX was achieved. After the quasi-steady state condition test, the PCCT water level test was performed with the goal of investigating thermal-hydraulic behavior during a water level decrease in the PCCT. After reaching the predetermined goals of the thermal power in the steam generator and the heat removal rate in the PCHX, the characteristics of the natural circulation in the loop were measured. A MARS-KS (Multi-dimensional Analysis for Reactor Safety, KINS Standard Version) code analysis was conducted to validate the experimental data. The results will contribute
Fig. 1. A schematic diagram of APR+ PAFS.
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Fig. 3. The design of APR+ PCCT.
to the improvement of the condensation and boiling heat transfer models.
2. Description of the PASCAL facility The PASCAL facility was designed according to a volumetric scaling methodology (Nahavandi et al., 1979). The methodology preserves the elevation change between a heat source and a heat sink in a natural circulation loop under the same pressure and temperature conditions. Before conducting the experiments, the scaling methodology applied in the design of the PASCAL facility was validated by the MARS-KS code analysis in a previous study (Cho et al., 2012). Fig. 4 shows the schematic diagram of the PASCAL facility. A steam generator supplies saturated steam to the PCHX tube. Electrical heaters in the steam generator provided the heat source that scaled down the heat transfer rate at the U-tube surface. To preserve the driving force of the natural circulation in the loop, a distance equivalent to that of the prototype was maintained between the mixture level in the steam generator and the PCHX tube was maintained. The steam generator was connected to the PCHX tube with a steam-supply line and a return-water line. To determine the PCHX design parameters for the PAFS, the requirements for the heat exchanger must be satisfied. The PAFS condensation heat exchanger can remove the decay heat from the reactor core during accidents. Therefore, the heat removal rate has to balance
the anticipated decay heat level obtained from the safety analysis of the APR+. Another important requirement is preventing the condensation-induced water hammer inside the tube. Particularly, a horizontal heat exchanger was considered for the PAFS because the heat exchanger should be designed to prevent the water hammer phenomenon to ensure structural integrity. In this experimental study, the PCHX of the PASCAL facility simulates a single tube among 240 tubes in the prototype; hence, the volumetric scaling ratio of the facility is 1/240. The PCCT pool volume was also reduced to 1/240 of the prototype. From the top view of the PCCT as shown in Fig. 3, total 4 bundles were immerged in the PCCT. One bundle consisted of 60 tubes and 120 tubes are located in the half side. Then, the length of the PCCT was scaled by 1/2 and the width of that was scaled by 1/120. The length, the width, and height of the PCCT are 6.7 m, 0.112 m, and 11.484 m, respectively. This aspect ratio was designed to preserve the natural convection flow inside the PCCT. Table 1 compares the major design and scaling parameters of the PASCAL facility with those of the prototype. The PCCT of the PASCAL facility was designed before the prototype PCCT was changed so a discrepancy appears in the PCCT length and the width, as shown in Table 1. The increase in the PCCT length and width resulted in a conservative prediction for the cooling performance of the PCHX due to the weakness of the natural circulation. In Table 1, the PCCT height means the minimum water level of the PCCT as the design requirements of the PAFS. The MARS-KS code analysis is performed to verify the influence of the water level. The influence of the PCCT water level is negligible as shown in Fig. 5.
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Table 1 The geometry and scaling parameters of the PAFS and the PASCAL facility. Parameter
APR+
PASCAL
Ratio
PCHX tube
Inner/outer diameter Length Number of tubes Material Operating condition
44.8 mm/50.8 mm 8.4 m 240 Stainless steel 304 7.4 MPa 290 ◦ C
44.8 mm/50.8 mm 8.4 m 1 Stainless Steel 304 7.4 MPa 290 ◦ C
1/1 1/1 1/240 – –
PCCT
Pool height Pool length Pool width Elevation from SG water level
7.62 m (Previously 8.9 m) 18.29 m (Previously 13.4 m) 13.56 m (Previously 13.4 m) 16.2 m
8.9 m 6.7 m 0.112 m 16.2 m
1/0.86 (previously 1/1) 1/2.7 (previously 1/2) 1/121 (previously 1/120) 1/1
In the PASCAL test, the major measuring parameters are the flow rate of the steam and condensate liquid, the loop temperature and pressure, the differential pressure on the steam-supply line and the return-water line. Additionally, the collapsed water level and the liquid temperature were measured in the PCCT pool. To induce the local distribution of the heat flux and the heat transfer coefficient of the PCHX, the surface temperatures at the inside
Fig. 5. The influence of the PCCT water level (MARS analysis).
and the outside walls were measured at eleven points along the tube length (as shown in Fig. 6). At each point, the fluid temperature profile inside the tube was measured in a vertical direction so that the distribution of condensate liquid flow inside the PCHX tube could be inferred. Uncertainty (of the measured experimental data) was analyzed in accordance with a 95% confidence level. According to the ASME performance test code 19.1, the uncertainty level of the present results was estimated by an unsystematic rootmean-square random error and a systematic error (Kim, 2011). The random and systematic errors were evaluated from the data acquisition hardware specifications and the calibration results performed once every year. Table 2 shows the analyzed uncertainty levels of each group of instruments. 3. Experimental conditions and procedure 3.1. Experimental conditions The nominal experimental condition was determined according to the PAFS maximum heat removal requirement. Because a single Table 2 Instrument uncertainty levels.
Fig. 4. A schematic diagram of the PASCAL.
Items
Unit
Uncertainty
Static pressure Differential pressure Collapsed water level Temperature Flow rate Heater power Heat flux (tube, pool side) Heat transfer coefficient (tube side) Heat transfer coefficient (pool side)
MPa kPa m ◦ C kg/s W W/m2 W/m2 K W/m2 K
±0.365% ±6.7% ±4.0% Maximum 4.4 ◦ C ±0.93% ±0.595% ±15.3% ±3.1% ±2.6%
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Fig. 6. The Measurement of wall and fluid temperature.
train of the PAFS is required to remove 129.8 MW as a maximum heat removal rate by 240 tubes of the PCHX, 540 kW of thermal power was supplied to generate a steam flow into the single tube of the PASCAL facility according to the scaling ratio of the facility (1/240). These test cases were regarded as the SS/PL-540-P1 test, where SS/PL means a combined experiment of the continuous simulation for a quasi-steady state condition (SS) and a decrease of the PCCT water level (PL). The quasi-steady state condition focused on characteristics of the natural circulation flow when the heat balance between the steam generator and the PCHX was achieved at a nominal PCCT water level.
3.2. Experimental procedure The overall procedure, including the SS and the PL tests, is listed below. Step 1: Pre-heating and flashing of the steam generator system After ventilating the non-condensable gas, the steam generator, the steam-supply line, the PCHX tube, and the return-water line were filled with water. The electrical heater of the steam generator heated the system until the fluid temperature reached to 180 ◦ C with a forced convection flow driven by a circulation pump in the loop. The system pressure was maintained higher than a saturation pressure of the fluid system to prevent boiling inside the steam generator. After the fluid system temperature reached 180 ◦ C, the forced convection flow was stopped. Then the steam generator system vent valve was opened to the atmosphere to decrease an inventory and make a mixture level in the steam generator 1.2 m by flashing the steam flow. Step 2: Starting the data acquisition and achieving the quasisteady state condition
When the collapsed water level in the steam generator was reduced by 1.2 m, the vent valve of the steam generator system was closed to stop flashing. That closure made a natural circulation flow of the steam in the loop. Data acquisition began at that moment. The thermal power of the electrical heaters in the steam generator was gradually increased to the thermal power of the experimental condition, while maintaining an increasing rate of the system temperature as 1 K/min. The water in the PCCT reached a saturated state at the atmospheric pressure via the PCHX heat transfer. The evaporation of the PCCT pool water decreased the water level to 9.3 m, which is the nominal water level at the saturated state. Because the initial PCCT water level in the prototype is 8.9 m at room temperature, the moment that the PCCT water level reached 9.3 m at the saturated condition (due to a reduced density) was considered to be a quasi-steady state. When the pressure, temperature, and flow rate reached a quasi-steady state at the constant thermal power condition, the heat removal rate and the natural circulation flow provided the data for the SS test. Step 3: Decreasing the PCCT water level to 3.5 m The continuous supply of the thermal power in the steam generator made the PCCT water level decrease by evaporation. The experimental data measured during the decrease of the PCCT water level provided the data for the PL test. The data acquisition continued until the PCCT water level decreased to 3.5 m, which was near the top of the PCHX. 4. Experimental results and MARS-KS code analysis 4.1. Experimental results of the SS/PL-540-P1 test Table 3 summarizes the overall sequence of events for the SS/PL-540-P1 test according to the test procedure. From the data acquisition time of 0–4200 s, the thermal power of the electrical
Table 3 The sequence of events for the SS/PL-540-P1 test. Phase
Time
Sequence
Start of data acquisition
0s
Heat-up phase Transient phase
0–4200 s 4200–13,200 s
Data of SS-540-P1 case Data of PL-540-P1 case
13,200–13,300 s 13,300–32,600 s
Flashing is stopped after maintaining the initial SG system inventory. Natural circulation flow is generated in the loop Stepwise increase of SG heater power up to 540 kW Time period until the system pressure and temperature reached a steady-state condition and the PCCT water level was 9.3 m Quasi-steady state condition at the continuous thermal power Decrease of PCCT water level
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Fig. 7. The heat removal rate of the SS/PL-540-P1 test.
heaters in the SG increased incrementally by 1 K/min to preserve the system temperature. After the heater power reached 540 kW, the pressure and the temperature of the SG system increased until a heat balance was achieved between the heat source (the heater) and the PCHX heat removal. At approximately 8000 s, the water level of the PCCT began decreasing due to the evaporation caused by the boiling heat transfer at an outer wall of the PCHX. Experimental data from the system pressure, the mass flow rate, and the loop temperature for the quasi-steady state condition were gathered during the period from 13,200 s to 13,300 s. This data were used to perform the SS-540-P1 test. When the water level in the PCCT decreased to 3.5 m, the experimental data acquired after the steady-state condition was used for the PL-540-P1 test. Figs. 7–10 presented the heat removal rates, the steam generator system pressure, the steam generator system temperature, and the mass flow rate for the SS/PL-540-01 test. In the PASCAL test, the heat removal rates are listed in Table 4, which includes the supplied thermal power of electrical heaters in steam generator (HR SG), the summation of heat transfer rate of the PCHX tube surface (HR Tube), the enthalpy decrease between the inlet and the outlet of the PCHX (HR SS and HR RW). Among the calculated heat removal rates, HR Tube was considered to be the most reliable value due to an integration of the local heat transfer rate along the PCHX tube length. When the thermal power of steam generator was small in the heat-up phase, a natural circulation flow rate was
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Fig. 9. The steam generator system temperature of the SS/PL-540-P1 test.
Fig. 10. The steam generator system flow rate of the SS/PL-540-P1 test. Table 4 The definition of heat removal rate. Notation
Description
HR SG
The thermal power supplied by the electrical heaters in SG
HR Tube
Heat removal rate of PCHX calculated from summation of wall heat transfer rate HR Tube =
ip
" 2r L qavg,o o ip
" " " qavg,o = qtop,o + qbottom,o /2 kTube (Tw,i −Tw,o ) k (T −Tw,o )bottom top " " , qbottom,o = Tube w,i qtop,o = ro ln(ro /ri ) ro ln(ro /ri ) ri : Inner radius of tube r0 : Outer radius of tube Lip : The length of the tube segment (ip is the measurement position number from 1 to 11)
Fig. 8. The steam generator system pressure of the SS/PL-540-P1 test.
HR SS
The heat removal rate calculated from the flow rate in the steam-supply line ˙ ss (hin − hout ) HRSS = m ˙ ss : mass flow rate of the steam-supply line m hin : fluid enthalpy of the PCHX inlet hout : Fluid enthalpy of the PCHX outlet
HR RW
The heat removal rate calculated from the flow rate in return-water line ˙ RW (hin − hout ) HRRW = m ˙ RW : Mass flow rate of the return-water line m hin and hout : Same to those in HR SS
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Fig. 11. The wall temperature of the PCHX tube.
small so that a flow oscillation in the return-water line occurred and it affected the oscillation of HR RW calculation as shown in Fig. 7. While the thermal power of electrical heaters in steam generator increased stepwise in the heat-up phase, the heat removal rate of PCHX also increased following the supplied thermal power. During the heat-up phase, the difference between the HR SG and the HR Tube in the transient phase is the transient conduction heat transfer of the PCHX tube. Because the supplied thermal power was larger than the removed heat transfer rate at PCHX during these periods, the pressure and temperature of the steam generator system increased continuously, as shown in Figs. 8 and 9. During the quasi-steady state condition from 13,200 to 13,300 s, the heat removal rate of the PCHX tube was equivalent to the supplied power. At the quasi-steady state, the pressure of the steam generator system reached 3.22 MPa and the steam temperature was 242 ◦ C. The PCHX of the prototype PAFS was designed to remove the decay heat at the steam condition of 7.4 MPa and 290 ◦ C. The experimental result for the quasi-steady state condition indicated that the current design of the PCHX had enough of a thermal margin to cool the reactor core without an active safety injection system. Fig. 10 shows the natural circulation flow rate, where the mass flow rate measured in the steam-supply line and the return-water line presented similar behavior during the quasi-steady state, which means that the current PAFS design can cool the secondary system of the steam generator stably in a passive manner. After the quasisteady state, the heat removal rate increased due to an increase of boiling heat transfer in the PCCT water pool. Therefore, the pressure, the temperature, and the natural circulation flow rate of the steam generator system decreased, as shown in Figs. 8–10. Figs. 11–14 show variations in the surface temperature, the heat flux, the heat transfer coefficient and the fluid temperature along the tube length during the quasi-steady state condition. As the steam flowed toward the outlet, the heat flux decreased due to the lower wall temperature, which was induced by the subcooled liquid flow. The condensation heat transfer coefficient at the top region of the inner wall was larger than that of bottom region because the top part of the tube was filled with the steam flow and the condensate liquid flowed in the bottom region inside the tube. The phase distribution could be interpreted by the fluid temperature measurement shown in Fig. 14, which was also plotted as a contour in Fig. 15. The temperature at point E (the bottom region as shown in Fig. 5) was lower than the saturation temperature along the whole tube length. Near the outlet of the PCHX tube, the fluid temperature at point D also revealed the subcooled water (by an increase of the liquid fraction). The round region of the tube tends to
Fig. 12. The heat flux of the PCHX tube.
Fig. 13. The heat transfer coefficient of the PCHX tube.
mix the condensate liquid flow in nature, so that the fluid temperature showed a uniform profile. From the distribution of the fluid temperature inside the PCHX tube, it was confined that a stratified flow appeared along the whole length of the tube without a bubbly flow or a slug flow. The NUREG report suggested several
Fig. 14. The fluid temperature in the PCHX tube.
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Fig. 15. The fluid temperature contour inside the tube.
guidelines for inhibiting the water hammer in a horizontal channel (Griffith, 1997). Excluding slug flow is one of the requirements, which prevented a condensation-induced water hammering. The experimental results are in agreement with the criterion for inhibiting a condensation-induced water hammer phenomenon inside the tube. Fig. 16 shows a variation of the PCCT water level in the SS/PL540-P1 test. As the pool temperature increased during the heat-up phase, the water level increased due to the volume expansion of the PCCT pool. When the pool water reached a saturated state, the collapsed water level in the PCCT began to continuously decrease. The lower water level of the PCCT means the lower static pressure and the lower saturation temperature of the pool water around the PCHX tube, so that a decrease of the PCCT water level resulted in much of the nucleate boiling on the tube surface. The heat transfer coefficients at points 1 and 11 (as shown in Fig. 6) were presented in Fig. 17. In this study, the boiling heat transfer coefficient at the PCHX tube surface increased in response to the decrease of the PCCT water level, while the condensation heat transfer coefficient showed a uniform behavior with the variation of the PCCT water level. Fig. 18 compares a collapsed water level of the steam generator with the return-water line. During the period from heat-up until 4200 s, the water level difference between the SG and the returnwater line increased due to an increase of the supplied thermal power and a drop in pressure. After the quasi-steady state condition was achieved, the water level of the return-water line showed increased slightly due to the generation of a larger condensate liquid flow in the loop. Due to level difference between the steam generator and return-water line, condensate water can be returned back to the SG with a natural driving force.
Considering that the heat exchanger of the PAFS was designed to remove the decay heat of APR+, the experimental result proved that the current design of the PCHX had enough of a margin to cool down the reactor system without any active safety injection system. The mass flow rate measured on the steam-supply line and the return-water line presented similar behavior in the steady state. This indicates that the current PAFS design can supply the auxiliary feedwater and passively but effectively cool down the secondary steam generator system.
Fig. 16. The collapsed water level of PCCT.
Fig. 18. The water level in the loop system.
Fig. 17. The heat transfer coefficient during the PCCT water level decrease.
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Fig. 19. A nodalization diagram for the PASCAL facility.
4.2. MARS-KS code analysis The experimental data of the quasi-steady state condition, SS540-P1 test case, was utilized to validate the prediction capability of the MARS-KS code. The MARS-KS code is a best-estimate system analysis code based on the two-fluid model (Cho et al., 2012; Jeong et al., 1999). The MARS-KS code has the capability of analyzing a one-dimensional or a three-dimensional best-estimated thermal hydraulic system and the fuel responses of the light-water reactor transients. It treats the two-phase flow phenomena with six equations for the transfer of mass, momentum, and energy. The MARS-KS code models the condensation heat transfer coefficient as the maximum of the values yielded by the Nusselt’s model and
Shah’s model (Schaffrath et al., 1999). The geometry of the PASCAL facility was modeled as shown in Fig. 19. The PCHX tube was modeled as a single pipe with heat structures and divided by 13 volumes along the length. The PCCT pool was modeled using the multi-D component in the MARS-KS code to effectively simulate a multi-dimensional natural convection flow in the rectangular pool. To simulate the experimental condition of the SS-540-P1 test, the heater power in the steam generator was supplied with 540 kW. The initial condition of a coolant inventory and a system pressure in the steam generator and the PCCT was maintained equivalent to that of the SS-540-P1 test. In the MARS-KS code analysis, a PCCT pool water level decreased to 9.3 m by evaporation at approximately 2250 s (as shown in Fig. 20). The time difference for when PCCT pool water level was 9.3 m was caused by the experimental procedure of increasing the heater power. Fig. 21 presents the
Fig. 20. The MARS-KS calculation of the PCCT water level.
Fig. 21. The MARS-KS calculation of steam generator pressure.
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analysis result for the steam generator pressure. Compared to the experimental result at the quasi-steady state condition, the original MARS-KS code overestimates the system pressure in the quasisteady state condition. This overestimation means that the code underestimated a heat transfer rate at the PCHX, which proves that the analysis result of the MARS-KS code is sufficiently conservative in calculating the amount of condensation heat transfer in the PCHX of the PAFS. The use of condensation heat transfer coefficient measured in the experiment, however, dramatically improves the calculation result. There is a need to improve the condensation and boiling heat transfer model in MARS-KS code. 5. Conclusion This study presents the experimental results from the PASCAL facility. The PASCAL facility was constructed to validate the cooling performance of the PAFS and to investigate the condensation heat transfer and natural circulation phenomena in the PAFS. In this study, test cases of SS-540-P1 and PL-540-P1 were performed. The objective of the SS-540-P1 test is to investigate the cooling performance and natural circulation characteristics of the PAFS by simulating a quasi-steady state condition of the thermal power according to the volumetric scaling methodology. The PL-540-P1 test was conducted after the SS-540-P1 test, when the collapsed water level in the PCCT was continuously decreasing from evaporation. The test results proved that the current design of the PCHX satisfied the heat removal requirement for cooling down the reactor core during an accident condition. Therefore, it can be preliminarily concluded that the PAFS can replace a conventional active auxiliary feedwater system in the APR+, utilizing the natural circulation of two-phase flow. However, a simulation of SS-540-P1 test by the MARS-KS code showed a conservative result with respect to the heat removal rate of the horizontal PCHX. The present experimental results will contribute toward improving the model of the condensation and boiling heat transfer and also toward providing
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the benchmark data for validating the calculation performance of a thermal hydraulic system analysis code with respect to the PAFS. Acknowledgment The authors would like to acknowledge the financial support provided by the Ministry of Knowledge and Economy of the Korean government. References Bae, B.U., et al., 2011. Design of horizontal condensation heat exchanger for PAFS (Passive Auxiliary Feedwater System) in APR+. In: International Conference of Multiphase Flow in Industrial Plants, September 21–23, Ischia, Italy. Bae, B.U., et al., 2012. Design of condensation heat exchanger for the PAFS (Passive Auxiliary Feedwater System) of APR+ (Advanced Power Reactor Plus). Ann. Nucl. Energy 46, 134–143. Cheon, J., et al., 2010. The development of a passive auxiliary feedwater system in APR+ track 1: water-cooled reactor programs & issues. In: Proceedings of ICAPP’10, June 13–17, San Diego, CA, USA. Cho, Y.J., et al., 2012. Analytical studies of the heat removal capability of a passive auxiliary feedwater system (PAFS). Nucl. Eng. Des. 248, 306–316. Griffith, B., 1997. Screening Reactor Steam/Water Piping Systems for Water Hammer. NUREG/CR-6519. Jeong, J.J., Ha, K.S., Chung, B.D., Lee, W.J., 1999. Development of a multi-dimensional thermal-hydraulic system code, MARS 1.3.1. Ann. Nucl. Energy 26 (18), 16111642. Kang, K.H., et al., 2012. Separate and integral effect tests for validation of cooling and operational performance of the APR+ passive auxiliary feedwater system. Nucl. Eng. Technol. 44 (6). Kim, S., 2011. Data Analysis Methodology of Separate Effect Test Facility for Passive Auxiliary Feedwater System. APR+-PAFS-DD-01. Nahavandi, A.N., et al., 1979. Scaling laws for modeling nuclear reactor systems. Nucl. Sci. Eng. 72, 75-83. Schaffrath, A., Hicken, E.F., Jaegers, H., Prasser, H.M., 1999. Experimental and analytical investigation of the operation mode of the emergency condenser of the SWR1000”. Nucl. Technol. 126, 123–142. Song, C.H., et al., 2010. Thermal-hydraulic R&Ds for the APR+ developments in Korea. In: Proceedings of the 18th International Conference on Nuclear Engineering, May 17–21, Xi’an, China. Yun, B.J. et al., 2010. Construction Report of Separate Effect Test Facility for PAFS, KAERI/TR-4085/2010.