Analyses of operation performance of advanced secondary passive residual heat removal system in PWR

Annals of Nuclear Energy 135 (2020) 106998

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Annals of Nuclear Energy journal homepage: www.elsevier.com/locate/anucene

Analyses of operation performance of advanced secondary passive residual heat removal system in PWR Liangguo Li a,⇑, Qianhua Su a, Jianming Yu a, Donghua Lu a, Peng Ju a, Chenyu Hao a, Xiaohang Wu a, Feng Zhu b a b

China Nuclear Power Technology Research Institute Co. Ltd., Shenzhen 518026, China China Nuclear Power Technology Engineering Co. Ltd., Shenzhen 518026, China

a r t i c l e

i n f o

Article history: Received 26 February 2019 Received in revised form 10 July 2019 Accepted 19 August 2019

Keywords: Advanced secondary passive residual heat removal system Operation performance Baring of heat exchanger tube Experimental analysis Sensitivity investigation

a b s t r a c t The newly designed advanced secondary passive heat removal system (ASP) was one of the most important cooling systems to cope with station blackout (SBO) accident in the generation II + 1000 MWe PWR nuclear power plant (NPP). Advanced secondary passive heat removal system test facility (ASPTF) was constructed to obtain the operation performance of ASP under SBO conditions. The geometrical scaling ratio of ASPTF is 1/4 in the height and 1/208 in the volume with respect to the generation II + 1000 MWe PWR. The drag coefficient experiments were conducted to match the prototypical resistance characteristics of ASP in this paper firstly. Then, a prolonged SBO accident sequence was experimentally investigated in ASPTF. The experimental results reveal that the secondary side pressure of steam generator (SG) maintained at the set value in the early stage with the reciprocating opening of vapor discharge to atmosphere valve (VDA). After the startup of ASP, the secondary side pressure of SG decreased and heat generated in core simulator was effectively removed by natural circulation through ASP within 3 h (h) of the design value. Three hours later, the flow rate of ASP became unstable with the baring of heat exchanger (HX) tube. The stable natural circulation of ASP was recovered after the injection of water in the water tank (WT) after 4 h. A sensitivity investigation on operation performance of ASP also was performed based on RELAP5 code. This study not only helps to improve the understanding of important thermal hydraulic phenomena in ASP, but also sheds some lights on operation optimization for the PWRs with ASP. Ó 2019 Elsevier Ltd. All rights reserved.

1. Introduction SBO accident is one of the most important design extension conditions (DECs), which has attracted wide international attention after the Fukushima Dai-ichi accident. Without any proper action during an SBO transient, a loss of the heat sink would leads to an uncovery and even a meltdown of the reactor core. ASP has been widely adopted to cope with SBO accident in the newly built generation III 1000 MWe PWR in recent years (Sun et al., 2017). Considering the supervision requirements, the newly designed system should be verified before its engineering application. Experimental and numerical methods were adopted to investigate the performance of ASP in different types of PWR in recent years. However, it is difficult to obtain the operation performance of the prototype ASP under SBO conditions in the NPP. Considering the geometrical size, flow rate and thermal power of prototype ASP, ⇑ Corresponding author. E-mail address: [email protected] (L. Li). https://doi.org/10.1016/j.anucene.2019.106998 0306-4549/Ó 2019 Elsevier Ltd. All rights reserved.

it is also feasible and uneconomical to construct a test facility of the same size as the NPP. Owing to the development of scaling theory, the integral effect test facilities with reduced size were widely used to observe the system responses and important thermal hydraulic phenomena in prototype NPP. Considering the different design feature of ASP in specific type of PWR, the integral effect test facilities were built with particular scaling criteria. Xiao et al. (2003) conducted an experimental research on air cooling type of secondary side emergency passive residual heat removal system (PRHRS) for AC600. Chung et al. (2006) studied the operation performance of PRHRS in VISTA facility. Krepper and Beyer (2010) investigated the natural circulation characteristics of a passive decay heat removal system for BWR-1000 and ESBWR experimentally and numerically. Yu et al. (2013) obtained the performance of APR1400 in ATLAS facility under SBO conditions. Min et al. (2014) conducted a experimental study on operation performance of PRHRS for SMART in SMART-ITL facility. Wu et al. (2015) conducted a small scale test facility to obtain the natural circulation characteristics of PRHRS for China generation III PWR. Sun et al.

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L. Li et al. / Annals of Nuclear Energy 135 (2020) 106998

Nomenclature ASP SBO NPP ASPTF SG VDA HX WT DECs PRHRS

Advanced secondary passive heat removal system Station blackout Nuclear power plant Advanced secondary passive heat removal system test facility Steam generator Vapor discharge to atmosphere valve Heat exchanger Water tank Design extension conditions Emergency passive residual heat removal system

(2017, 2018) designed a full height test facility to study the safety performance of PRHRS in HPR1000. The best estimate codes were also widely used in the simulation of secondary passive residual heat removal system. Chung et al. (2008) simulated the PRHRS cooldown performance of a Korean 65 MW integral reactor based on TASS/SMR code. Zio et al. (2010) calculated the safety margins of PRHRS for the high temperature reactor-pebble modular under different accidents conditions. Zhang et al. (2011, 2012) studied the PRHRS for CPR1000 under air cooling and water cooling conditions using RELAP5. Wang et al. (2012) studied the water cooling type of PRHRS in case of SBO accident for CPR1000 based on RELAP5. Then Wang et al. (2014) proposed an conceptual design of water cooling type PRHRS for CPR1000 and studied the transient characteristics of PRHRS under SBO and loss of feed water accidents using RELAP5. As mentioned above, the ASP has been widely adopted by different reactors. However, most nuclear plants in service and built are generation Ⅱ or Ⅱ+ reactors with less passive safety systems. The operation performance of different secondary passive heat removal systems mainly affected by the structure of SG and HX, the system arrangement, the system resistance characteristics and so on. Therefore, the performance of ASP applied on the newly built generationⅡ+ reactors should be studied. ASP designed in one of the domestic generationⅡ+ nuclear reactor was studied in this paper. The ASPTF based on scaling criteria was constructed and the long-time cooling simulation experiment under prolonged SBO conditions was conducted. The test loop, test procedures, experimental results and sensitivity simulation results based on RELAP5 will be presented as follows.

RCS EWT PIRT H M L H2TS DP FP WR

Reactor coolant system Elevated water tank Phenomena identification and ranking table High Moderate Low Hierarchical two-tiered approach to scaling Pressure difference Full power Wide range

through the inlet pipeline, and to the feed water pipeline through the outlet pipeline. The VDA is connected to the main steam pipeline. The schematic of ASP is shown in Fig. 1. The ASP is designed to operate automatically from standby mode when SBO accident occurs. At the initial of SBO accident, the pressure of secondary side of SG dramatically increased with the closing of main steam isolation valve. When the pressure of secondary side of SG achieved the set value, the VDA would open automatically. The discharge capability of VDA is enough to remove the decay heat. With the release of vapor, the liquid level of secondary side of SG decreased. When the liquid level of secondary side of SG reached the set value, the isolation valves attached to the HX inlet pipeline and outlet pipeline open automatically in turn. The steam flows through the inlet pipeline and enters the tube side of HX. The steam is then condensed and the released heat is transferred to water in WT. The condensed water from HX flows back to the secondary side of SG, and evaporates after absorbing the heat from the primary side of SG. With the condensation of steam in HX tube, the pressure of secondary side of SG is gradually decreased. The VDA is closed when the system pressure is lower than the set value. The temperature of water in WT increases after absorbing the heat from the tube side of HX and the water in WT evaporates finally. Then, the liquid level of WT continuously decreased with the evaporation of water in WT. When HX tube is bareness at 6 h, the water supply valve would open automatically and the water from EWT injected into WT. Therefore, the decay heat in the core is efficiently transferred to the water in WT through natural circulation, which could protect the core from overheating. 2.2. Description of ASPTF

2. Description of prototype PWR and ASPTF 2.1. Description of prototype PWR This nuclear reactor, which is a three-loop 1000 MWe pressurized water reactor, has been developed as a new passive safety and generationⅡ+ reactor based on the mature PWR technology. The ASP of this generationⅡ+ 1000 MWe PWR is comprised of three loops each corresponding to one of three reactor coolant system (RCS) loops. Each ASP loop includes a SG, a HX located in WT, a VDA and corresponding valves and pipes. As the final heat sink of ASP, WT is full of water at room temperature, where the three HXs located in. The total volume of WT is 600 m3, which is designed to meet the requirement of the system operation in the preliminary 6 h considering the special design of main pump in RCS. After 6 h, water from the passive high elevated water tank (EWT) was injected into WT, which was driven by gravity. With the recharge of WT, ASP can efficiently remove the heat generated from core in 72 h. The HX is connected to the main steam pipeline

ASPTF is an integral test facility scaled from one loop of ASP in generationⅡ+ 1000 MWe PWR, which has a 1/4 height and 1/208 vol scale of the prototype. It is constructed to investigate the integral effect of residual heat removal performance of ASP and separate effect of HX. The ASP and RCS of generationⅡ+ 1000 MWe PWR are simulated in ASPTF as accurate as possible. A detailed scaling analysis was done for ASPTF to simulate the system response and important thermal hydraulic phenomena identified by the phenomena identification and ranking table (PIRT) (Wilson and Boyack, 1998). The PIRT of ASP is listed in Table 1. If the phenomenon can significantly impact the peak cladding temperature, it is ranked high (H). Similarly, M and L represents moderate and low, respectively. The basic scaling analysis was performed based on the hierarchical two-tiered approach to scaling (H2TS) method (Zuber et al., 1998) which has already been used by APEX (Reyes and Hochreiter, 1998), ATLAS (Park, et al., 2007), and ACME (Li et al., 2016). The scaling analysis of natural circulation in ASP and RCS

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L. Li et al. / Annals of Nuclear Energy 135 (2020) 106998

Elevated water tank

Water supply valve

Vapor discharge to atmosphere valve

Heat exchanger

Water tank Pressurizer Steam generator

Reactor

Main pump Fig. 1. Schematic of ASP in PWR.

was developed based on modified drift flux model (Lu et al., 2010). The scaling ratios are listed in Table 2. The schematic of ASPTF is shown in Fig. 2. The working fluid flow diagram of ASPTF is similar to prototype PWR. The water from feed water system evaporates in the secondary side of SG and the steam flows to the condenser under normal conditions. With the closing of isolation valve of condenser and opening of valves in steam pipeline and backwater pipeline, the steam generated in SG condenses in HX and transfers heat to WT under SBO accident simulation conditions. Then, the condensed water flows into the secondary side of SG due to gravity. The heat transferred to WT finally dissipates to the atmosphere. The mechanical loop system of ASPTF mainly includes the primary loop system, secondary loop system, feed water system, circulating cooling water system. The primary loop system is composed of core simulator, main pump, SG, pressurizer, safety valve and connecting pipes, which are scaled from the RCS. The secondary system includes SG, flow meter, VDA,

throttles, HX, isolation valves, WT, regulating valves, safety valve, and connecting pipes, which are scaled from the ASP and related systems. The feed water system provides water for the secondary side of SG. The circulating cooling water system can meet the requirements for main pump cooling, condenser cooling and WT injecting. Although the EWT was not designed in ASPTF, the function of EWT was realized by circulating cooling water system. A throttle device is installed at the steam pipeline and backwater pipeline respectively. Size of throttles is customized to simulate the prototypical resistance characteristics of ASP. Several pneumatic valves were installed at inlet of condenser, inlet of the secondary side of SG, inlet and outlet of HX to reflect the function of prototype valves. The maximum heating power of the core simulator is 1.0 MW. The operation pressure and temperature of ASPTF are identical to that of prototype. The main design parameters of ASPTF are listed in Table 3.

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L. Li et al. / Annals of Nuclear Energy 135 (2020) 106998

Table 1 PIRT analysis of ASP. Component phenomenon SG two phase natural circulation heat transfer secondary conditions (including liquid level) primary conditions Steam and backwater pipeline flow pattern radial heat dissipation Isolation valve open/close characteristics VDA flow rate system pressure HX heat transfer secondary conditions two phase natural circulation flow pattern WT evaporation surface water inventory

Operation of VDA

starting up of ASP

long term cooling

L L H

H H H

H H L

L

M

L

L L

H L

L L

M

H

L

M H

M M

L L

L L L L

H M H M

H M H L

L L

M M

H M

Table 2 Scaling ratios of ASPTF. Parameters

Scaling ratio

Pressure Temperature Height/length Diameter Flow area Velocity Time Flow rate/power Volume Drag coefficient

1:1 1:1 1:4 1:7.2 1:52 1:2 1:2 1:104 1:208 1:1

The ASPTF has a total of over 200 instrumentations which were calibrated according to the quality assurance program and inspected by the third party. The key measuring parameters are temperature, pressure, pressure difference (DP), liquid level, flow rate, electric current and voltage. According to the range of temperature, suitable type of thermocouple is used for measuring temperature of the atmosphere and the fluid at various locations. T type thermocouples were employed in measuring atmosphere temperature, water temperature of WT and circulating cooling water system. N type thermocouples were used in high temperature fluid measuring. The pressure of components and systems has been measured by pressure transducers of suitable range. The pressure drop across flow path and the liquid level of components have been measured by differential pressure transducers. The liquid flow rates in pipes have been measured using venturi meters combined with differential pressure transducers. The heating power of core simulator was calculated by electric current and voltage. The electric current of core simulator has been measured by current transducer. The voltage of core simulator has been measured by voltage transducer. The detailed description of main instrumentations and their uncertainties are listed in Table 4. According to the calibration report, the maximum absolute error of the temperature is 0.8. The data acquisition system is established based on national instruments acquisition system and LabVIEW program, which is programmed to monitor the operation of ASPTF and save experi-

mental data. The instrument control system is employed to operate the valves and regulate the heating power automatically or manually. The heating power of core simulator can be adjusted continuously and the scaled residual power can be simulated conveniently. The data have been collected at the frequency of 3 Hz in the experiment. 3. Experimental methods 3.1. Resistance matching The resistance along the flow path is one of the most important influencing factors of natural circulation and the resistance matching is necessary before the experiment. The resistance matching experiment is carried out on SG, steam pipeline, HX and backwater pipeline. This experiment achieved by regulating the opening size of the throttle device and regulating valves installed at ASPTF. The procedure of resistance matching is listed as follows: a) Calculate the resistance coefficient characteristics of all components in prototype and ASPTF, and obtain their resistance differences in all components; b) Design the opening size of throttles based on the above step and install them on ASPTF; c) Disconnect the tee of backwater pipeline and establish forced circulation conditions at room temperature. Measure resistance of the components showed in Fig. 3 individually. The resistance of backwater is calculated by two parts added as shown in Fig. 3. The drag coefficient characteristics varying with Reynolds number were obtained by adjusting the flow rate. The size of throttles and valve opening were adjusted until the relative deviation of difference between measured ones and prototype is within 10% individually. The size of throttle #1 and valve opening of backwater pipeline were obtained in this step. The drag coefficient characteristic of steam pipeline varying with higher Reynolds number would be carried out under natural circulation conditions. d) Connect the tee of backwater pipeline and establish natural circulation conditions with the adjusted reactor simulator heating power. Measure resistance of the components showed in Fig. 4 individually. The drag coefficient characteristics varying with Reynolds number were obtained by continuously increasing the heating power to 1.2% scaled full power (FP). The size of throttle #2 and regulating valve opening of steam pipeline was adjusted until the relative deviation of difference between measured one and prototype is within 10%. The result of the prototypical resistance matching is listed in Table 5. Considering the scaling criteria, the drag coefficient of manufactured steam generator and heat exchanger with reduced flow area were larger than designed value of prototype, which were tested without any throttle in the experiment. 3.2. Transient experiment The heating power of core simulator is scaled from the decay power of prototype and the heat loss of ASPTF is also considered in the transient experiment. Considering the temperature limitation of main pump and that the main purpose of this experiment is to observe whether the decay heat can be efficiently removed, the main pump is in operation in experiment. The initial secondary side liquid level of SG was 100% wide range (WR). The WT and HX tube is full of water at initial, which was at room temperature. The heating power of core simulator was gradually increased to 1 MW. The secondary side pressure of SG was gradually reached 7.85 MPa. After stable initial boundaries were established, the automatic control program is executed and the

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L. Li et al. / Annals of Nuclear Energy 135 (2020) 106998

Pressure reducing valve Isolation valve of condenser

To condenser Water tank Safety valve Vapor discharge to atomosphere valve Isolation valve of steam pipeline Regulating valve of steam pipeline

Heat exchanger

Throttle #1 Safety valve Flow meter Pressurizer

Regulating valve of backwater pipeline Throttle #2 Isolation valve of backwater pipeline From feedwater system

Isolation valve of water tank inletline Isolation valve of water tank outletline

Isolation valve of feedwater system Steam generator

Main pump Reactor simulator Fig. 2. Schematic flow diagram of ASPTF.

Table 3 Design parameters of ASPTF. ASPTF design

Parameters

Design pressure of the primary loop (MPa) Design temperature of the primary loop ( ) Maximum operation pressure of the primary loop (MPa) Maximum operation temperature of the primary loop ( ) Maximum heating power (MW) Design pressure of the secondary loop (MPa) Design temperature of the secondary loop ( ) Maximum operation pressure of the secondary loop (MPa) Maximum operation temperature of the secondary loop ( )

17.23 360 15.5 320 1.0 10 311 8.6 303

isolation valves of condenser and feed water system are closed. When the system pressure reached 7.85 MPa, the VDA located on the pipeline of ASP would open automatically. When the secondary side liquid level of SG decreased to 45% WR, the isolation valve of steam pipeline and backwater pipeline automatically opened in turn, then ASP was in operation. The operation characteristics of ASP were obtained not only during 3 h, which was represented 6 h of prototype considering the scaling ratio. The prolonged performance of ASP was also got with the water injected into WT at 4 h. The total operation time of ASP was 5 h in this experiment.

Table 4 Main instrumentations and their uncertainties. Categories

Instrumentation type

Instrumentation location

Range

Maximum permission error

Flow rate Pressure Pressure Pressure Pressure Liquid level Liquid level Temperature Temperature Temperature Temperature Temperature Pressure difference Pressure difference Pressure difference Pressure difference Electric current Voltage

Venturi meter Pressure transducer Pressure transducer Pressure transducer Pressure transducer DP transducer DP transducer N type thermocouple N type thermocouple N type thermocouple N type thermocouple T type thermocouple DP transducer DP transducer DP transducer DP transducer current transducer Voltage transducer

HX outlet SG outlet HX inlet HX outlet Venturi meter inlet SG WT HX inlet HX outlet SG outlet Venturi meter inlet WT SG HX Steam pipeline backwater pipeline DC power DC power

0.05–1.50 kg/s 0–12 MPa 0–12 MPa 0–12 MPa 0–12 MPa 0–100 kPa 0–100 kPa 0–450 0–450 0–450 0–450 0–150 0–100 kPa 0–100 kPa 0–100 kPa 0–100 kPa 0–24000 A 0–250 V

0.5% 0.1% 0.1% 0.1% 0.1% 0.1% 0.1% 0.4% 0.4% 0.4% 0.4% 0.4% 0.1% 0.1% 0.1% 0.1% 0.03% 0.1%

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L. Li et al. / Annals of Nuclear Energy 135 (2020) 106998

DP2 DP3 DP1 DP4 DP5

To waterways Regulating valve of feedwater system From feedwater system Isolation valve of feedwater system Fig. 3. Schematic of ASP under forced circulation conditions.

DP2 DP3

DP1 DP4

From feedwater system

were divided by an arbitrary value and plotted on the nondimensional axis. Fig. 5 shows the residual heating power curve of core simulator. The decay power of prototype PWR decreased dramatically at the initial. According to the scaling ratio, the initial full power decreased to 10% FP within tens of seconds. In most test facilities, reduced-scaled core power, rather than full scaled core power, is adopted at steady state condition and at initial short period of the transient due to both economic concern and scaling constraints, for example, in the APEX (Reyes and Hochreiter, 1998) and in the ATLAS (Park, et al., 2007). Therefore, The ASPTF was held at 10% scaled FP constant to 85 s to obtain correct integral energy. The pressure of secondary side of SG is shown in Fig. 6. The VDA was actuated when the system pressure reached the set value, then the system pressure was maintained around the set value. The heat transferred to SG was mainly removed by the steam released through VDA in this period. This is because the released ability of VDA is large enough, which leads to its reciprocating opening. The liquid level of secondary side of SG was decreased with the releasing of steam. The ASP commissioning signal was triggered when the liquid level of secondary side of SG reached 45% WR. The isolation valve of steam pipeline opens, then, the pressure of HX inlet increased dramatically at 1045 s. After few seconds, the isolation valve of backwater pipeline opens. The cooling ability of HX is adequate to remove the energy stored in the solid structures and transferred to SG. Therefore, the pressure of SG outlet and HX inlet were decreased gradually until 4 h and the system pressure was decreased again after water injected into WT as shown in Figs. 6 and 7, in which the pressure was decreased from rapid to slow. The natural circulation flow rate of ASP was mapped in Fig. 8. The initial water in HX tube flows into the secondary side of SG by gravity, leading to the peak of flow rate. A stable natural circulation was established and the flow rate was slightly decreased in 3 h with the gradually decreasing heating power. The stable flow rate was about 21% of the peak value. After 3 h, the natural circulation began to be unstable until 4 h. The fluctuation of flow rate at this stage was attributed to the bareness of HX tube resulting from the continuously decreasing of liquid level of WT. Fig. 9 gives the temperature change of HX inlet and outlet. The temperature of HX inlet was increased dramatically from room temperature to saturation temperature after the isolation valve of steam pipeline opened. Then, the temperature of HX outlet was increased dramatically after the isolation valve of backwater pipeline opened. The temperature of HX inlet and outlet were both

Fig. 4. Schematic of ASP under natural circulation conditions.

1.2

Table 5 Resistance matching result. ASPTF

Prototype

Absolute deviation

Drag coefficient of SG Drag coefficient of steam pipeline Drag coefficient of HX Drag coefficient of backwater pipeline Total drag coefficient

0.82 76.82 15.93 72.49

0.29 70.39 10.35 72.90

0.53 6.43 5.58 0.41

166.1

153.9

12.2

1.0

1.0

Non-dimensional heating power(-)

Parameters

Non-dimensional heating power(-)

1.2

0.8

0.6

0.8

0.6

0.4

0.2

0.4

0.0 0

0.2

20

40

60

80

100

Time (s)

4. Results and discussion 4.1. Transient experimental results According to the project requirement, all the data should be confidential. Thus, all of the experimental results in this paper

0.0 0

2000

4000

6000

8000

10000 12000 14000 16000 18000

Time(s) Fig. 5. Residual heating power curve of core simulator.

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L. Li et al. / Annals of Nuclear Energy 135 (2020) 106998 1.2

1.0

0.8

1.2

HE inlet HE outlet

1.0

0.8

0.6

0.4 0

400

800

0.6

1200

1600

2000

Time (s)

0.4

Non-dimensional temperature (-)

Reciprocating opening of VDA

Non-dimensional pressure of SG outlet (-)

Non-dimensional pressure of SG outlet (-)

1.2

0.2 0

2000

4000

6000

8000

1.0

0.8

Injection of water

0.6

0.4

0.2

0.0

10000 12000 14000 16000 18000

0

Time (s)

2000

Non-dimensional pressure of HE inlet (-)

Operation of ASP

0.8

0.6

0.4

0.2

0.0 2000

4000

6000

8000

8000

10000 12000 14000 16000 18000

Fig. 9. Temperature of HX inlet and outlet.

1.2

0

6000

Time (s)

Fig. 6. Pressure of secondary side of SG.

1.0

4000

10000 12000 14000 16000 18000

Time (s)

the changing of heating power. The temperature of HX inlet and outlet reached a stable level at 4 h. After 4 h, the temperature and temperature difference were both decreased again with water at room temperature injected into WT. In addition, Fig. 10 shows the liquid level of WT began to decrease with the evaporation of water at 1 h and reached the initial value again after water was recharged at 4 h. As shown in Fig. 10, the temperature of WT was increased with the operation of ASP at 1045 s. The temperature of WT reached the saturation temperature at 1 h by the heat transfer from core simulator to water in WT. Then, the temperature of WT was held at saturation point until the water was recharged at 4 h. The temperature of WT reached the saturation temperature again after 4.5 h. Owing to the recharging of water in WT, the HX tube was covered by water again. The pressure of HX inlet and secondary side of SG were decreased again at 4 h as shown in Fig. 6 and Fig. 7. The stable natural circulation was also recovered at 4 h as shown in Fig. 8.

Fig. 7. Pressure of HX inlet.

4.2. Sensitivity analysis

1.0

0.8

0.6

0.8

0.6

0.2

0.0 0

500

1000

1500

2000

Time (s)

0.2

Bareness of HE tube 0.0 2000

4000

Liquid level Temperature

Evaporation of water

0.4

0

0.5

1.2 0.4

6000

8000 10000 12000 14000 16000 18000 20000

Time (s) Fig. 8. Natural circulation flow rate of ASP.

1.0

0.4

0.8

Injection of water

0.3

0.6 0.2 0.4

Injection stopped 0.1

0.2

Operation of ASP 0.0

decreased gradually in 4 h. The temperature decreasing rate of HX outlet is relatively small compared to that of HX inlet. The temperature difference between HX inlet and HX outlet decreased with

0

2000

4000

6000

8000

10000

12000

14000

16000

Time(s) Fig. 10. Liquid level and temperature of WT.

0.0 18000

Non-dimensional temperature of WT(-)

Non-dimensional flow rate (-)

1.0

Non-dimensional flow rate (-)

The experimental results show that the heat generated in the primary loop can be removed by ASP under the design conditions. The SBO accident also can be mitigated by an appropriate accident management measure. However, the natural circulation dominant heat removal process may be affected by the drag coefficient of

1.2

Non-dimensional liquid level of WT(-)

1.2

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L. Li et al. / Annals of Nuclear Energy 135 (2020) 106998

steam pipeline, the drag coefficient of backwater pipeline, the start-up conditions of ASP, so that the sensitivity analysis on the effects were performed. A thermal hydraulic system analysis code, RELAP5, was utilized to assess the system response of ASPTF at various conditions of the sensitive parameters. Fig. 11 shows the nodalization of ASPTF, with most of the hydraulic components composed of pipes, branches, single volumes, and junctions. A sensitivity analysis was carried out to optimize the suitable nodalization before the nodes were adopted, but it is not presented here due to space limitations. The selected conditions to be simulated in this paper are listed in Table 6. Figs. 12–15 compares the pressure of secondary side of SG, system pressure of ASP and natural circulation flow rate of ASP in the simulation and experimental results. The natural circulation flow rate of ASP in simulation result was a little higher than experimental result during the initial 1 h, which mainly caused by the calculation deviation of condescension in HX tube and natural circulation in WT. Yu et al. (2008) and Zhou et al. (2013) suggest that condensation model in RELAP5 needs to be improved for better predictions of different type passive heat removal systems. After 1 h, the heat transferred to secondary side of HX tube was removed by the evaporation of water in WT and this mechanism could be better simulated by RELAP5, which led to better simulation of natural circulation flow rate in this period. As shown in Fig. 15, the fluid was supercooled water in the experiment. However, the calculated temperature of HX outlet was higher than

experimental result, which mainly was due to the underestimate of condescension inside HX tube in the simulation. After 3 h, with the bareness of HX tube, the pressure descends slowly. However, the simulated pressure is slightly increased before the water injected into WT at 4 h. The pressure increases because of the bareness of the HX tubes due to water evaporation in the water tank. With water injected into WT, the pressure decreased again in simulation and experimental results. The sensitivity analysis results of operation time of ASP with the same strategy for water injection of WT were shown in Figs. 16-18. The pressure of ASP and SG were fluctuating around the set value of VDA after ASP was operated in Transient-1 and Transient-2, which was caused by the lower capacity of HX compared with the residual heat during this period. The later the operation of ASP, the more relatively enough capacity of HX for residual heat removal, which led to the pressure of ASP and secondary side of SG gradually decreased after ASP was operated in other simulation conditions. Earlier operation of ASP would transfer more heat to WT, which results in higher temperature of water in WT or lower liquid level after the water was boiling at the same time. The lower liquid level of WT may result in more bareness area of HX tube. The higher temperature of water in WT and the more bareness area of HX tube reduced the heat transfer capacity of HX. Therefore, the earlier operation of ASP, the higher pressure of ASP and secondary side of SG will reach. The natural circulation flow rate of ASP was similar, except for the initial fluctuating in Transient-1 and

Fig. 11. Schematic nodalization of ASPTF.

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L. Li et al. / Annals of Nuclear Energy 135 (2020) 106998 Table 6 The selected conditions to be simulated. Simulating condition NO.

Power

Transient-1 Transient-2 Transient-3 Transient-4 Transient-5 Transient-6 Transient-7 Transient-8

Base Base Base Base Base Base Base Base

Resistance

power power power power power power power power

backwater pipeline

Base steam pipeline Base steam pipeline Base steam pipeline Base steam pipeline Base steam pipeline Base steam pipeline Elevated steam pipeline Base steam pipeline

Base backwater pipeline Base backwater pipeline Base backwater pipeline Base backwater pipeline Base backwater pipeline Base backwater pipeline Base backwater pipeline Elevated backwater pipeline

0.6

3h

0.4

4h

1.0

Non-dimensional flow rate (-)

0.8

Experiment Transient-3

1.2

1.0

Operation of ASP

0.8

0.6

0.8

0.6

0.4

0.2

0.4 0.0 0

500

1000

1500

2000

Time (s)

0.2

0.2

0.0 0

2000

4000

6000

8000

10000

12000

14000

16000

18000

0

2000

4000

6000

Time (s)

8000

10000

1.2

14000

16000

18000

Fig. 14. Natural circulation flow rate of ASP in simulation and experimental results.

1.2

HE inlet in experiment HE outlet in experiment HE inlet in Transient-3 HE outlet in Transient-3

Experiment Transient-3

Operation of ASP

Non-dimensional temperature (-)

1.0

0.8

0.6

3h

0.4

12000

Time (s)

Fig. 12. Pressure of secondary side of SG in simulation and experimental results.

Non-dimensional pressure of HE inlet (-)

100%WR 70%WR 45%WR 30%WR 20%WR 10%WR 45%WR 45%WR

1.2

Experiment Transient-3

Reciprocating of VDA 1.0

steam pipeline

Non-dimensional flow rate (-)

Non-dimensional pressure of SG outlet (-)

1.2

Trigger signal of ASP

4h

1.0

0.8

0.6

0.4

0.2

0.2

0.0

0.0 0

2000

4000

6000

8000

10000

12000

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Time (s) Fig. 15. Temperature of HX inlet and outlet in simulation and experimental results. Fig. 13. Pressure of HX inlet in simulation and experimental results.

Transient-2. Initial fluctuating of natural circulation flow rates in the mentioned two conditions were caused by the reciprocating opening of VDA. The heat transferred from the primary loop can be effectively removed witin 3 h and the pressure was decreased again after water was injected into WT at 4 h in the mentioned simulated conditions. The earlier operation of ASP can reduce the steam generated from the secondary side of SG released from VDA to atmosphere of containment. The later operation of ASP can provide more time for the operator.

Figs. 19–21 presented the sensitivity analysis results of elevated resistance of steam pipeline and backwater pipeline. The pressure of ASP and secondary side of SG was almost the same in different conditions, which mainly because of the minimal changing of pressure caused by the elevated drag coefficient compared with the system pressure. The natural circulation flow rate of ASP was slightly decreased with the increased drag coefficient of steam pipeline or backwater pipeline. This phenomenon can be directly explained by the mechanism of natural circulation. The peak value

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L. Li et al. / Annals of Nuclear Energy 135 (2020) 106998 1.2

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Time (s) Fig. 16. Pressure of secondary side of SG in sensitivity analysis of operation time.

Fig. 19. Pressure of secondary side of SG in sensitivity analysis of resistance.

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Fig. 20. Pressure of HX inlet in sensitivity analysis of resistance. Fig. 17. Pressure of HX inlet in sensitivity analysis of operation time.

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Time (s) Fig. 21. Natural circulation flow rate of ASP in sensitivity analysis of resistance.

L. Li et al. / Annals of Nuclear Energy 135 (2020) 106998

of natural circulation flow rate was decreased only in Transient-8 with elevated drag coefficient of backwater pipeline. This situation was because of the peak value of natural circulation was formed by the outflow of initial water in HX tube and affected by both the inner liquid level of HX tube and the drag coefficient of back water pipeline. 5. Summary and conclusion ASPTF has been constructed to evaluate the safety performance of ASP of generation II+ 1000 MWe PWR under SBO scenarios. Prototypical resistance matching has been conducted to restore the prototypical resistance characteristics under cold and hot shakedown tests before the experiment, respectively. The transient experimental results showed that the pressure of secondary side of SG was maintained at 7.85 MPa by the reciprocating opening of VDA. With the trigger signal of 45% WR of SG, ASP was operated and the VDA was shut down at the same time. The pressure of secondary side of SG was decreased gradually at 3 h and this downtrend was continued to 4 h even though the tube of HX was absorbed at 3 h. Stable natural circulation was established in ASP during 3 h. Although the natural circulation of ASP begun to oscillate after HX was bareness, it was recovered to the stable state after water was injected into WT. From the experimental results, it was proved that the residual heat can be efficiently removed by ASP. The experimental data can also be used to evaluate the prediction capability of safety analysis codes and identify any code deficiency in predicting an SBO transient. Sensitivity analysis with RELAP5 code indicates that the elevated drag coefficient of steam pipeline and backwater pipeline mainly affected the natural circulation flow rate. The downward trend of system pressure was similar under different resistance conditions. It was found that the trigger signal of ASP mainly affected the operation time of VDA. In addition, the downward trend of system pressure and natural circulation flow rate of ASP were similar under different trigger signal conditions. Residual heat was efficiently removed by ASP under all simulation conditions. Acknowledgments The authors are grateful to the China Nuclear Power Technology Engineering Co. Ltd for its sponsorship for this project. Appendix A. Supplementary data Supplementary data to this article can be found online at https://doi.org/10.1016/j.anucene.2019.106998.

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