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Analysis of loss of residual heat removal system during mid-loop operation of a 300 MWe PWR
Progress in Nuclear Energy 112 (2019) 1–6
Contents lists available at ScienceDirect
Progress in Nuclear Energy journal homepage: www.elsevier.com/locate/pnucene
Analysis of loss of residual heat removal system during mid-loop operation of a 300 MWe PWR
T
Mahmood Shakila, Muhammad Ilyasa,∗, Farhan Zahida, Masroor Ahmada, Fatih Aydoganb a b
Pakistan Institute of Engineering and Applied Sciences, Islamabad, Pakistan Jacksonville University, Florida, USA
A R T I C LE I N FO
A B S T R A C T
Keywords: Residual heat removal system Mid-loop operation Decay heat RELAP5 PWR Reflux condensation
Thermal-hydraulic response of a two-loop pressurized water reactor is studied on loss of residual heat removal system (RHRS) during mid-loop operation. Decay heat during mid-loop operation is removed by RHRS. On loss of RHRS, no cooling system is available to cool the core. Transient thermal-hydraulic analysis of the plant is performed using RELAP/SCDAP on loss of RHRS. For this purpose, first steady-state is achieved at mid-loop conditions. System obtains steady-state at primary coolant inventory of 58000 kg, water level of 7.8m in the reactor pressure vessel (RPV) and upper plenum pressure of 104.41 kPa. On loss of RHRS, the core uncovery time, maximum reactor system pressure and cladding temperature are determined under four accidental scenarios in absence of any mitigation measure. The results show that the core uncovery time for the open case of reactor pressure vessel top vent valve (RPV top vent valve), pressurizer relief valve (PZR valve), steam generator man way and pressurizer man way are 423.1, 214.4, 50.1, 67.2 min respectively. The effect of reflux condensation for the above accidental configurations has also been studied. It is found that reflux condensation is a prominent phenomenon to increases core uncovery time for open case of RPV top vent valve and PZR valve.
1. Introduction A nuclear power plant is operated in mid-loop conditions for testing of plant systems and their maintenance after the reactor shut down. In this mode of operation, the coolant level of the reactor is reduced such that it is partially filled with water (Lee et al., 1996). Decay heat of the plant is removed by residual heat removals system (RHRS). The water level in the primary system is maintained at mid-line of the hot leg (Seo and Park, 2000; Sui et al., 2017). On loss of RHRS, no other system is immediately available for decay heat removal from the core due to maintenance. If no cooling conditions persist for longer time, the coolant temperature may increase up to saturation temperature where coolant starts boiling. This could result in core uncovery with ultimate consequences of fuel cladding failure and core damage. Loss of RHRS accident occurred repeatedly in the past history of nuclear reactor operation (Nuclear Regulatory Commission, 1987; Nuclear Regulatory Commission, 1990). It can occur due to many reasons including loss of suction supply to RHRS pump, loss of electric power, accidental closures of valves, inaccuracy in reactor vessel instrumentation and lack of knowledge between water level and RHRS pump speed (Queral et al., 2006). There are some mechanisms/systems available to increase time for core uncovery and maximum reactor
∗
coolant system (RCS) pressurization after loss of RHRS. These are reflux condensation, gravity feed from refueling water storage tank and safety injection system. The effectiveness of these mechanisms/systems depends upon primary system accidental configuration and decay power of the plant. The RHRS plays an important role in mid-loop operation for safe refueling/maintenance of the plant. Loss of RHRS can lead toward severe accident. Research on transient analyses of loss of RHRS has been performed widely. A brief summary of the relevant research on this topic is presented here. Gonzalez et al. (González et al., 2007) carried out research on loss of RHRS for three-loop PWR using TRACE code for two accidental configurations – pressurizer relief valves open (PORVs) and closed primary system. The results show that RPV is highly pressurized however the core remains uncovered in case of closed primary system. While in case of open pressurizer relief valves, primary system is less pressurized but core is uncovered due to steam loss from relief valve. Queral et al.,2008 studied further accidental configurations – RPV top vent valve open, pressurizer man way open – on loss of RHRS for this reactor under both dry layup and wet layup conditions. They found that main parameters of primary and secondary system are highly dependent on vent configurations and number of available steam generators for reflux condensation. It is found that if vessel vent is
Corresponding author. E-mail address: [email protected] (M. Ilyas).
https://doi.org/10.1016/j.pnucene.2018.11.010 Received 23 April 2018; Received in revised form 26 October 2018; Accepted 19 November 2018 0149-1970/ © 2018 Elsevier Ltd. All rights reserved.
Progress in Nuclear Energy 112 (2019) 1–6
M. Shakil et al.
allowed then less mass of steam is released in containment and system is less pressurized. Birchley et al. (2008a) determined probability of component failure which can lead toward core damage during mid-loop operation for twoloop Westinghouse PWR. They performed analyses for closed upper head, open upper head and detention bolts of upper head of RPV in mid-loop conditions using MELCOR. The results show that system is highly pressurized in case of closed upper head and it takes longer time for core to uncover, loss of geometry and core damage. While in case of removed upper head of vessel, system is less pressurized and core damage time is also short. Tong et al. (2009) analyzed loss of residual heat removal system for 300 MWe pressurized water reactor in mid-loop conditions using RELAP5 code for closed primary system, pressurizer relief valve open, RPV top valve open, reactor coolant pump seal open and steam generator man way open cases. He also studied effect of different mitigation measure for closed system configuration under mid-loop conditions. Gou et al. (2017) performed safety assessment on loss of residual heat removal system during mid-loop operation and high level cases for CPR1000 plant using RELAP 5/MOD 3.3. The authors studied 12 different plant configurations and determined time for core uncover, coolant boiling, water inventory and coolant temperature and pressure. Results show that pressure is maximum in case of intact system. It is found that start of coolant boiling time depends upon inventory of coolant in primary system and behavior of water inventory in RPV on the loss of residual heat removal system while time for core to uncover depends upon plant system configurations. Many other researchers (Birchley et al., 2008b; Carlos et al., 2008; Dumont et al., 1994; Shen and Luo, 2000; Zhang et al., 2012) have analyzes loss of RHRS under different accidental configurations. However, owing to the severity of the accident, thermal-hydraulic analysis is required to determine variations in temperature, pressure, and water level in the reactor core for different possible accidental configurations during loss of RHRS. It is also required to estimate time for core to uncover so that operator can take mitigation measures before core uncovers. The objective of this work is to analyze loss of RHRS for different accidental configurations and study the effect of reflux condensation for these accidental configurations using RELAP/SCDAP code. What follows is section 2 which deals with materials and methods. A brief description of the reference nuclear power plant is presented in section 2.1. The simulation tool used and the modeling philosophy is given in section 2.2. Results are presented in section 3. Section 3.1 is about achieving steady-state operating conditions. The transient analysis results are presented in section 3.2. Four accidental configurations are considered under dry and wet layup conditions of secondary side. Important conclusions are drawn in section 4.
Table 1 Plant nominal operating parameters (Shen and Luo, 2000). Parameter
Value
Reactor power (MWth) Reactor core inlet temperature (K) Reactor core outlet temperature (K) Pressurizer pressure (MPa) Reactor pressure vessel level (m) Steam generator pressure (MPa)
998.6 561.65 588 15.3 12.5 5.54
2.2. RELAP/SCDAP model RELAP5 is a widely used system code for thermal-hydraulic analysis of light water reactors (Commission, 1998). In the present work, RELAP/SCDAP Mod 3.2 developed by Innovative System Software (Allison and Hohorst, 2010) has been used for thermal-hydraulic analysis of the reference reactor in mid-loop operation following loss of RHRS. The nodalization diagram of the reactor and associated systems is shown in Fig. 1. The reactor core is divided in two channels – average and bypass channels. The fuel assemblies are modeled as heat structure attached to the average channel. Each channel is divided into 8 axial sub-volumes. The steam generator is modeled using single volumes, pipe components, control variables, junctions and heat structures. Pressurizer is modeled as a pipe component having seven sub-volumes while pressurizer dome is modeled as a branch component. A surge line is modeled as a pipe component connected to hot leg of one of the two loops. Model of the reactor coolant pump is available as a separate component in RELAP/SCDAP. The RHRS is modeled with two time dependent volumes (660, 860), a time dependent junction 659 and a single junction 859. Water is supplied from time dependent volume 660 and collected in time dependent volume 860. Flow rate is maintained with the help of time dependent junction 659 at a rate of 142 kg/s to remove decay heat from the reactor core. Loss of RHRS is modeled by a trip in RELAP/SCDAP. In the transient analysis, four vent configurations are considered as given in Table 2. 3. Results and discussions 3.1. Achieving steady-state conditions Before transient analysis on loss of RHRS, steady-state conditions are achieved under mid-loop operation. For this purpose, the reactor coolant system is modeled with water up to mid-line level of the hot leg. Above the water level, primary system is filled with air at pressure and temperature of 101.325 kPa and 323 K respectively. Reactor coolant pumps are modelled as switched off because they are not operating in mid-loop operation. Decay heat generation after reactor shut down is calculated from equation given below:
P (t ) = 0.066P0 [ts−0.2 − (ts+τ)−0.2]
2. Materials and methods
(1)
where P0 is the reactor nominal power, τs is reactor operation time and ts is time after shut down (Todreas and Kazimi, 1990). Decay heat generation rate after 2 days of reactor shutdown with infinite reactor operation time is determined to be 5.9 MWth.
2.1. Description of the reference nuclear power plant A two-loop 300 MWe PWR is selected as the reference nuclear power plant (Shen and Luo, 2000) to performed thermal-hydraulic analysis under loss of RHRS during mid-loop operation. The nominal operating parameters of the selected nuclear power plant are given in Table 1. Each loop contains a steam generator, a reactor coolant pump, a hot leg and a cold leg. The RHRS consists of two parallel trains connected with one of the hot legs of RCS. Each train contains a pump, heat exchanger, instrumentations/valves for operational control and associated piping. One of these trains is redundant as each individual train has capacity to remove all the decay heat. In this analysis, only one train is considered in operation to remove all the decay heat.
3.2. Transient analysis After achieving steady-state in mid-loop operation, transient analysis is performed on loss of RHRS without any mitigation measure. Four accidental configurations are studied under dry and wet layup conditions of secondary side. 3.2.1. Transients under dry layup conditions In the base case analysis, steam generators are kept under dry layup with air at pressure and temperature conditions of 101.325 kPa and 2
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Fig. 1. Complete nodalization diagram for thermal-hydraulics analysis of the plant. Table 2 Vent configurations and their area of flow. Vent Configuration
Flow area (m2)
Pressurizer relief valve (PZR valve) Reactor pressurizer top vent valve (RPV top vent valve) Pressurizer man way (PZR man way) Steam generator man way (SG man way)
1.32 × 10−3 4.91 × 10−4 0.1661 0.13202
Fig. 3. Variation in the coolant temperature in core for different cases.
different cases studied. It can be seen from the graph that coolant temperature rises with time. The coolant temperature almost becomes constant for the open cases of steam generator man way and pressurizer man way as boiling occurs at saturated conditions in these cases. After incipience of coolant boiling, liquid level in RPV starts to decrease. RPV liquid level has been plotted in Fig. 4 as a function of time. The results show that core is uncovered quickly as steam leaves the system in open case of steam generator man way and pressurizer man way because their flow area is large. The time for core to uncover and maximum reactor coolant system pressurization is determined. The core uncovery time for PZR man way, PZR valve, RPV top vent valve and SG man way is 67.283 min, 297.33 min, 423.166 min and 50.133 min respectively. As no cooling media is available, an increase in the fuel rod and cladding temperatures is observed. There is an abrupt rise in the cladding temperature after a certain time depending upon the case studied. The sequence of rise in cladding temperature, as shown in Fig. 5, is same as that of core to uncover. Fig. 6 shows variation in steam flow rate from vents for all above cases. Steam flow rate is the highest in open case of the SG man way because of its larger flow area. The summary of results for dry layup cases is given in Table 3.
Fig. 2. Variation in the pressure of upper plenum for different cases.
298 K respectively. Fig. 2 shows variation of upper plenum pressure in reactor pressure vessel. It can be seen from the figure that pressure increases with passage of time. This increase in pressure is dominant in case of open reactor pressure vessel top vent valve (RPV top vent valve) and pressurizer relief valve (PZR valve) because of their smaller area. In both cases, the steam production rate is greater than venting rate. However, when the core liquid level decreases below bottom of the core, no more steam is produced however, the accumulated steam keeps on venting. This results in further pressure reduction. Maximum reactor coolant pressure for pressurizer man way, PZR valve, RPV top vent valve and SG man way open case is 1.88 × 105Pa, 8.02 × 105Pa, 2.47 × 106 Pa and 1.12 × 105 Pa respectively. Fig. 3 shows response of coolant temperature in reactor core for the 3
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Fig. 4. Variation in the coolant level in rector pressure vessel. Fig. 7. Upper plenum pressure for PZR valve open.
Fig. 5. Variation in the cladding temperature for different cases. Fig. 8. Upper plenum pressure for RPV top vent valve open.
and RPV top vent valve. In both the cases, the pressure first increases, passes through a peak and then decreases due to the reason explained in dry layup case. It is seen from the figures that higher pressure is achieved for the RPV top vent valve case. The RCS peak pressure for opening of both PZR valve and RPV top vent valve has been shifted toward later time in case of wet layup case as compared to dry layup case because of reflux condensation. Fig. 9 and Fig. 10 show a comparison of coolant temperature for dry and wet layup conditions for PZR valve and RPV top vent valve open cases. The trend of curve for both dry and wet layup cases is same, however, the slop of pressure rise is higher for the former case. Fig. 11 and Fig. 12 show liquid level inside reactor pressure vessel for dry and wet layup cases. It is seen from the graphs that time for core to uncover is considerably larger in opened cases of both RPV top vent valve and PZR valve for wet layup conditions as compared to dry lay conditions due to reflux condensation provided by steam generator. The variation of fuel rod cladding temperature with time for PZR valve and RPV top vent valve open configurations under wet layup condition is shown in Fig. 13 and Fig. 14. Difference in time for abrupt rise in cladding temperature for dry and wet layup condition revels the fact that reflux condensation plays a significant role to keep the reactor core covered for longer time. The steam flow rate in containment for dry and wet layup cases for PZR valve and RPV top vent valve open configurations are compared in Fig. 15 and Fig. 16. It can be seen from
Fig. 6. Steam flow rate into the containment for different cases.
3.2.2. Transients under wet layup conditions In the SG wet layup case, secondary side of steam generator is filled with water up to the top of U-tubes. Water level in secondary side is maintained at 8.1 m against the pressure and temperature conditions of 101.325 kPa and 398 K respectively. Water in secondary side of SG causes condensation of the steam produced in the core. The condensate flows back to the core. This reflux condensation increases the time for the core to uncover. The results of the SG dry and wet layups are compared in Fig. 7 and Fig. 8 for two vent configurations – PZR valve Table 3 Summary of dry layup cases. Configuration
Time to start core uncover (min)
Maximum RCS pressurization (Pa)
Time for cladding to reach 1000 K (min)
Steam generator man way (SG man way) Pressurizer man way (PZR man way) Pressurizer relief valve (PZR valve) Reactor pressure vessel top vent valve (RPV top vent valve)
50.1 67.2 214.4 423.1
1.12 × 105 1.88 × 105 8.02 × 105 2.47 × 106
144.2 161.4 328.1 509.1
4
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Fig. 13. Cladding temperature for PZR valve open configuration.
Fig. 9. Coolant temperature at core inlet for open PZR valve.
Fig. 14. Cladding temperature for RPV top vent valve open configuration.
Fig. 10. Coolant temperature for RPV top vent valve open.
Fig. 11. RPV level for PZR valve open configuration.
Fig. 15. Steam flow rate in containment for PZR valve open configuration.
Fig. 12. RPV level for RPV top vent valve open configuration. Fig. 16. Steam flow rate in containment for RPV top vent valve open configuration.
5
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Table 4 Summary of dry and wet layup cases. Configuration PZR valve RPV top vent valve
Time to start core uncover (min) Dry layup Wet layup Dry layup Wet layup
Maximum RCS pressurization (Pa) 5
214.7 464.1 423.1 700
English Alphabets M Mega P Reactor Power t Time Greek Symbol
4. Conclusions Decay heat of a nuclear power plant depends upon reactor nominal thermal power, operation time of reactor and time after reactor shutdown. The RHRS plays an important role to remove decay heat of the plant during mid-loop operation for safe refueling/maintenance of the plant. Loss of the RHRS results in substantial rise in temperature/ pressure and can lead toward severe accident. The loss of RHRS can occur due to many reasons including loss of suction supply to pump, loss of electric power, accidental closures of valves, and inaccuracy in reactor vessel instrumentation and lack of knowledge between water level and RHRS pump speed. The incipience of coolant boiling on accident depends upon reactor power and coolant inventory in RPV. While time for start of core to uncover depends upon the vent configuration. The transient analysis for dry layup configurations show that time for core to uncover is larger for smaller opening area but system pressure is higher in this case. Time for core to uncover for PZR man way, PZR valve, RPV top vent valve and SG man way are 67.283 min, 297.33 min, 423.166 min and 50.133 min respectively. The maximum RCS pressure for all above configurations is 1.88 × 105Pa, 8.02 × 105Pa, 2.47 × 106 Pa and 1.12 × 105 Pa respectively. A sharp rise in cladding temperature is observed at the time of core uncovery. This rapid increase in clad temperature can lead to clad failure. The transient analysis for wet layup configurations shows that reflux condensation can increase time for core to uncover for PZR valve and RPV top vent valve but level of core is not maintained at mid-line for longer time. No reflux condensation is observed for SG man way and PZR man way because steam produced quickly leaves the system and level of RPV decreases and core is uncovered.
τ Reactor operation time Subscripts 0 s th e Units
Initial after shutdown Thermal Electric
K m W Pa s min
Kelvin meter Watt Pascal seconds minutes
References Allison, C.M., Hohorst, J.K., 2010. Role of RELAP/SCDAPSIM in nuclear safety. Sc. Technol. Nucl. Install. 17 2010. Birchley, J., Haste, T., Richner, M., 2008a. Accident management following loss of residual heat removal during mid-loop operation in a Westinghouse two-loop PWR. Nucl. Eng. Des. 238 (9), 2173–2181. Birchley, J., Haste, T.J., Richner, M., 2008b. Accident management following loss of residual heat removal during mid-loop operation in a Westinghouse two-loop PWR. Nucl. Eng. Des. 238 (9), 2173–2181. Carlos, S., et al., 2008. Analysis of a loss of residual heat removal system during mid-loop conditions at PKL facility using RELAP5/Mod3.3. Nucl. Eng. Des. 238 (10), 2561–2567. Commission, U.S.N.R., 1998. Scdap/relap5/mod 3.2 Code Manual. User's Guide and Input Manual, Nureg/cr-6150, Vol. 3, Rev. 1. U.s. Nuclear Regulatory Commission. Dumont, D., et al., 1994. Loss of residual heat removal during mid-loop operation: BETHSY experiments. Nucl. Eng. Des. 149 (1), 365–374. González, I., Queral, C., Expósito, A., 2007. Phenomenology during the loss of residual heat removal system at midloop conditions with pressurizer PORVs open: associated boron dilution. Ann. Nucl. Energy 34 (3), 166–176. Gou, J., et al., 2017. Analysis of loss of residual heat removal system for CPR1000 under cold shutdown operation. Ann. Nucl. Energy 105, 25–35. Lee, C.H., et al., 1996. Investigation of mid-loop operation with loss of RHR at INER integral system test (IIST) facility. Nucl. Eng. Des. 163 (3), 349–358. Nuclear Regulatory Commission, W.C.C.A.R.V., 1987. Loss of Residual Heat Removal System: Diablo Canyon, Unit 2, April 10, 1987. United States. p. 105. Nuclear Regulatory Commission, W.D.C., 1990. Loss of Vital Ac Power and the Residual Heat Removal System during Mid-loop Operations at Vogtle Unit 1 on March 20, 1990. pp. 516 United States. Queral, C., González, I., Expósito, A., 2006. Analysis of the cooling capability of steam generators during the loss of residual heat removal system at midloop operation with closed primary system. Ann. Nucl. Energy 33 (13), 1102–1115. Queral, C., González, I., Expósito, A., 2008. Analysis of abnormal operation procedures in sequences of loss of the RHRS at midloop operation. Ann. Nucl. Energy 35 (7), 1321–1334. Seo, J.K., Park, G.C., 2000. Return momentum effect on water level distribution during midloop operations. Nucl. Eng. Des. 202 (1), 97–108. Shen, Z., Luo, S., 2000. Design for Chashma nuclear power plant of Pakistan. Nucl. Power Eng. 21 (1), 44–47. Sui, D., et al., 2017. Research on capability of secondary passive residual decay heat removal system after Main Feedwater Line Break (MFLB) accident. Nucl. Eng. Des. 325, 156–163. Todreas, N.E., Kazimi, M.S., 1990. Nuclear Systems I : Thermal Hydraulic Fundamentals. Hemisphere, New York; London. Tong, L.L., et al., 2009. Thermal hydraulic behaviors during loss of RHR system at mid-loop operation of Chinese 300MWe PWR NPP. Nucl. Eng. Des. 239 (12), 3027–3033. Zhang, Y.P., et al., 2012. Design and transient analyses of emergency passive residual heat removal system of CPR1000. Nucl. Eng. Des. 242, 247–256.
Acknowledgment The authors are grateful to their respective Departments for extending support to complete this work. Appendix A. Supplementary data Supplementary data to this article can be found online at https:// doi.org/10.1016/j.pnucene.2018.11.010. Nomenclature Abbreviations RHRS RPV PZR RCS PWR
328.1 567.9 509.1 790.1
8.02 × 10 8.23 × 105 2.47 × 106 2.35 × 106
these figures that the peak value of steam flow rate is reduced in both cases due to the effect of reflux condensation in SG U tubes. The summary of results for all wet layup cases is shown in Table 4.
Time for cladding to reach 1000 K (min)
Residual Heat Removal Reactor Pressure Vessel Pressurizer Reactor Coolant System Pressurized Power Reactor 6
Contents lists available at ScienceDirect
Progress in Nuclear Energy journal homepage: www.elsevier.com/locate/pnucene
Analysis of loss of residual heat removal system during mid-loop operation of a 300 MWe PWR
T
Mahmood Shakila, Muhammad Ilyasa,∗, Farhan Zahida, Masroor Ahmada, Fatih Aydoganb a b
Pakistan Institute of Engineering and Applied Sciences, Islamabad, Pakistan Jacksonville University, Florida, USA
A R T I C LE I N FO
A B S T R A C T
Keywords: Residual heat removal system Mid-loop operation Decay heat RELAP5 PWR Reflux condensation
Thermal-hydraulic response of a two-loop pressurized water reactor is studied on loss of residual heat removal system (RHRS) during mid-loop operation. Decay heat during mid-loop operation is removed by RHRS. On loss of RHRS, no cooling system is available to cool the core. Transient thermal-hydraulic analysis of the plant is performed using RELAP/SCDAP on loss of RHRS. For this purpose, first steady-state is achieved at mid-loop conditions. System obtains steady-state at primary coolant inventory of 58000 kg, water level of 7.8m in the reactor pressure vessel (RPV) and upper plenum pressure of 104.41 kPa. On loss of RHRS, the core uncovery time, maximum reactor system pressure and cladding temperature are determined under four accidental scenarios in absence of any mitigation measure. The results show that the core uncovery time for the open case of reactor pressure vessel top vent valve (RPV top vent valve), pressurizer relief valve (PZR valve), steam generator man way and pressurizer man way are 423.1, 214.4, 50.1, 67.2 min respectively. The effect of reflux condensation for the above accidental configurations has also been studied. It is found that reflux condensation is a prominent phenomenon to increases core uncovery time for open case of RPV top vent valve and PZR valve.
1. Introduction A nuclear power plant is operated in mid-loop conditions for testing of plant systems and their maintenance after the reactor shut down. In this mode of operation, the coolant level of the reactor is reduced such that it is partially filled with water (Lee et al., 1996). Decay heat of the plant is removed by residual heat removals system (RHRS). The water level in the primary system is maintained at mid-line of the hot leg (Seo and Park, 2000; Sui et al., 2017). On loss of RHRS, no other system is immediately available for decay heat removal from the core due to maintenance. If no cooling conditions persist for longer time, the coolant temperature may increase up to saturation temperature where coolant starts boiling. This could result in core uncovery with ultimate consequences of fuel cladding failure and core damage. Loss of RHRS accident occurred repeatedly in the past history of nuclear reactor operation (Nuclear Regulatory Commission, 1987; Nuclear Regulatory Commission, 1990). It can occur due to many reasons including loss of suction supply to RHRS pump, loss of electric power, accidental closures of valves, inaccuracy in reactor vessel instrumentation and lack of knowledge between water level and RHRS pump speed (Queral et al., 2006). There are some mechanisms/systems available to increase time for core uncovery and maximum reactor
∗
coolant system (RCS) pressurization after loss of RHRS. These are reflux condensation, gravity feed from refueling water storage tank and safety injection system. The effectiveness of these mechanisms/systems depends upon primary system accidental configuration and decay power of the plant. The RHRS plays an important role in mid-loop operation for safe refueling/maintenance of the plant. Loss of RHRS can lead toward severe accident. Research on transient analyses of loss of RHRS has been performed widely. A brief summary of the relevant research on this topic is presented here. Gonzalez et al. (González et al., 2007) carried out research on loss of RHRS for three-loop PWR using TRACE code for two accidental configurations – pressurizer relief valves open (PORVs) and closed primary system. The results show that RPV is highly pressurized however the core remains uncovered in case of closed primary system. While in case of open pressurizer relief valves, primary system is less pressurized but core is uncovered due to steam loss from relief valve. Queral et al.,2008 studied further accidental configurations – RPV top vent valve open, pressurizer man way open – on loss of RHRS for this reactor under both dry layup and wet layup conditions. They found that main parameters of primary and secondary system are highly dependent on vent configurations and number of available steam generators for reflux condensation. It is found that if vessel vent is
Corresponding author. E-mail address: [email protected] (M. Ilyas).
https://doi.org/10.1016/j.pnucene.2018.11.010 Received 23 April 2018; Received in revised form 26 October 2018; Accepted 19 November 2018 0149-1970/ © 2018 Elsevier Ltd. All rights reserved.
Progress in Nuclear Energy 112 (2019) 1–6
M. Shakil et al.
allowed then less mass of steam is released in containment and system is less pressurized. Birchley et al. (2008a) determined probability of component failure which can lead toward core damage during mid-loop operation for twoloop Westinghouse PWR. They performed analyses for closed upper head, open upper head and detention bolts of upper head of RPV in mid-loop conditions using MELCOR. The results show that system is highly pressurized in case of closed upper head and it takes longer time for core to uncover, loss of geometry and core damage. While in case of removed upper head of vessel, system is less pressurized and core damage time is also short. Tong et al. (2009) analyzed loss of residual heat removal system for 300 MWe pressurized water reactor in mid-loop conditions using RELAP5 code for closed primary system, pressurizer relief valve open, RPV top valve open, reactor coolant pump seal open and steam generator man way open cases. He also studied effect of different mitigation measure for closed system configuration under mid-loop conditions. Gou et al. (2017) performed safety assessment on loss of residual heat removal system during mid-loop operation and high level cases for CPR1000 plant using RELAP 5/MOD 3.3. The authors studied 12 different plant configurations and determined time for core uncover, coolant boiling, water inventory and coolant temperature and pressure. Results show that pressure is maximum in case of intact system. It is found that start of coolant boiling time depends upon inventory of coolant in primary system and behavior of water inventory in RPV on the loss of residual heat removal system while time for core to uncover depends upon plant system configurations. Many other researchers (Birchley et al., 2008b; Carlos et al., 2008; Dumont et al., 1994; Shen and Luo, 2000; Zhang et al., 2012) have analyzes loss of RHRS under different accidental configurations. However, owing to the severity of the accident, thermal-hydraulic analysis is required to determine variations in temperature, pressure, and water level in the reactor core for different possible accidental configurations during loss of RHRS. It is also required to estimate time for core to uncover so that operator can take mitigation measures before core uncovers. The objective of this work is to analyze loss of RHRS for different accidental configurations and study the effect of reflux condensation for these accidental configurations using RELAP/SCDAP code. What follows is section 2 which deals with materials and methods. A brief description of the reference nuclear power plant is presented in section 2.1. The simulation tool used and the modeling philosophy is given in section 2.2. Results are presented in section 3. Section 3.1 is about achieving steady-state operating conditions. The transient analysis results are presented in section 3.2. Four accidental configurations are considered under dry and wet layup conditions of secondary side. Important conclusions are drawn in section 4.
Table 1 Plant nominal operating parameters (Shen and Luo, 2000). Parameter
Value
Reactor power (MWth) Reactor core inlet temperature (K) Reactor core outlet temperature (K) Pressurizer pressure (MPa) Reactor pressure vessel level (m) Steam generator pressure (MPa)
998.6 561.65 588 15.3 12.5 5.54
2.2. RELAP/SCDAP model RELAP5 is a widely used system code for thermal-hydraulic analysis of light water reactors (Commission, 1998). In the present work, RELAP/SCDAP Mod 3.2 developed by Innovative System Software (Allison and Hohorst, 2010) has been used for thermal-hydraulic analysis of the reference reactor in mid-loop operation following loss of RHRS. The nodalization diagram of the reactor and associated systems is shown in Fig. 1. The reactor core is divided in two channels – average and bypass channels. The fuel assemblies are modeled as heat structure attached to the average channel. Each channel is divided into 8 axial sub-volumes. The steam generator is modeled using single volumes, pipe components, control variables, junctions and heat structures. Pressurizer is modeled as a pipe component having seven sub-volumes while pressurizer dome is modeled as a branch component. A surge line is modeled as a pipe component connected to hot leg of one of the two loops. Model of the reactor coolant pump is available as a separate component in RELAP/SCDAP. The RHRS is modeled with two time dependent volumes (660, 860), a time dependent junction 659 and a single junction 859. Water is supplied from time dependent volume 660 and collected in time dependent volume 860. Flow rate is maintained with the help of time dependent junction 659 at a rate of 142 kg/s to remove decay heat from the reactor core. Loss of RHRS is modeled by a trip in RELAP/SCDAP. In the transient analysis, four vent configurations are considered as given in Table 2. 3. Results and discussions 3.1. Achieving steady-state conditions Before transient analysis on loss of RHRS, steady-state conditions are achieved under mid-loop operation. For this purpose, the reactor coolant system is modeled with water up to mid-line level of the hot leg. Above the water level, primary system is filled with air at pressure and temperature of 101.325 kPa and 323 K respectively. Reactor coolant pumps are modelled as switched off because they are not operating in mid-loop operation. Decay heat generation after reactor shut down is calculated from equation given below:
P (t ) = 0.066P0 [ts−0.2 − (ts+τ)−0.2]
2. Materials and methods
(1)
where P0 is the reactor nominal power, τs is reactor operation time and ts is time after shut down (Todreas and Kazimi, 1990). Decay heat generation rate after 2 days of reactor shutdown with infinite reactor operation time is determined to be 5.9 MWth.
2.1. Description of the reference nuclear power plant A two-loop 300 MWe PWR is selected as the reference nuclear power plant (Shen and Luo, 2000) to performed thermal-hydraulic analysis under loss of RHRS during mid-loop operation. The nominal operating parameters of the selected nuclear power plant are given in Table 1. Each loop contains a steam generator, a reactor coolant pump, a hot leg and a cold leg. The RHRS consists of two parallel trains connected with one of the hot legs of RCS. Each train contains a pump, heat exchanger, instrumentations/valves for operational control and associated piping. One of these trains is redundant as each individual train has capacity to remove all the decay heat. In this analysis, only one train is considered in operation to remove all the decay heat.
3.2. Transient analysis After achieving steady-state in mid-loop operation, transient analysis is performed on loss of RHRS without any mitigation measure. Four accidental configurations are studied under dry and wet layup conditions of secondary side. 3.2.1. Transients under dry layup conditions In the base case analysis, steam generators are kept under dry layup with air at pressure and temperature conditions of 101.325 kPa and 2
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Fig. 1. Complete nodalization diagram for thermal-hydraulics analysis of the plant. Table 2 Vent configurations and their area of flow. Vent Configuration
Flow area (m2)
Pressurizer relief valve (PZR valve) Reactor pressurizer top vent valve (RPV top vent valve) Pressurizer man way (PZR man way) Steam generator man way (SG man way)
1.32 × 10−3 4.91 × 10−4 0.1661 0.13202
Fig. 3. Variation in the coolant temperature in core for different cases.
different cases studied. It can be seen from the graph that coolant temperature rises with time. The coolant temperature almost becomes constant for the open cases of steam generator man way and pressurizer man way as boiling occurs at saturated conditions in these cases. After incipience of coolant boiling, liquid level in RPV starts to decrease. RPV liquid level has been plotted in Fig. 4 as a function of time. The results show that core is uncovered quickly as steam leaves the system in open case of steam generator man way and pressurizer man way because their flow area is large. The time for core to uncover and maximum reactor coolant system pressurization is determined. The core uncovery time for PZR man way, PZR valve, RPV top vent valve and SG man way is 67.283 min, 297.33 min, 423.166 min and 50.133 min respectively. As no cooling media is available, an increase in the fuel rod and cladding temperatures is observed. There is an abrupt rise in the cladding temperature after a certain time depending upon the case studied. The sequence of rise in cladding temperature, as shown in Fig. 5, is same as that of core to uncover. Fig. 6 shows variation in steam flow rate from vents for all above cases. Steam flow rate is the highest in open case of the SG man way because of its larger flow area. The summary of results for dry layup cases is given in Table 3.
Fig. 2. Variation in the pressure of upper plenum for different cases.
298 K respectively. Fig. 2 shows variation of upper plenum pressure in reactor pressure vessel. It can be seen from the figure that pressure increases with passage of time. This increase in pressure is dominant in case of open reactor pressure vessel top vent valve (RPV top vent valve) and pressurizer relief valve (PZR valve) because of their smaller area. In both cases, the steam production rate is greater than venting rate. However, when the core liquid level decreases below bottom of the core, no more steam is produced however, the accumulated steam keeps on venting. This results in further pressure reduction. Maximum reactor coolant pressure for pressurizer man way, PZR valve, RPV top vent valve and SG man way open case is 1.88 × 105Pa, 8.02 × 105Pa, 2.47 × 106 Pa and 1.12 × 105 Pa respectively. Fig. 3 shows response of coolant temperature in reactor core for the 3
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Fig. 4. Variation in the coolant level in rector pressure vessel. Fig. 7. Upper plenum pressure for PZR valve open.
Fig. 5. Variation in the cladding temperature for different cases. Fig. 8. Upper plenum pressure for RPV top vent valve open.
and RPV top vent valve. In both the cases, the pressure first increases, passes through a peak and then decreases due to the reason explained in dry layup case. It is seen from the figures that higher pressure is achieved for the RPV top vent valve case. The RCS peak pressure for opening of both PZR valve and RPV top vent valve has been shifted toward later time in case of wet layup case as compared to dry layup case because of reflux condensation. Fig. 9 and Fig. 10 show a comparison of coolant temperature for dry and wet layup conditions for PZR valve and RPV top vent valve open cases. The trend of curve for both dry and wet layup cases is same, however, the slop of pressure rise is higher for the former case. Fig. 11 and Fig. 12 show liquid level inside reactor pressure vessel for dry and wet layup cases. It is seen from the graphs that time for core to uncover is considerably larger in opened cases of both RPV top vent valve and PZR valve for wet layup conditions as compared to dry lay conditions due to reflux condensation provided by steam generator. The variation of fuel rod cladding temperature with time for PZR valve and RPV top vent valve open configurations under wet layup condition is shown in Fig. 13 and Fig. 14. Difference in time for abrupt rise in cladding temperature for dry and wet layup condition revels the fact that reflux condensation plays a significant role to keep the reactor core covered for longer time. The steam flow rate in containment for dry and wet layup cases for PZR valve and RPV top vent valve open configurations are compared in Fig. 15 and Fig. 16. It can be seen from
Fig. 6. Steam flow rate into the containment for different cases.
3.2.2. Transients under wet layup conditions In the SG wet layup case, secondary side of steam generator is filled with water up to the top of U-tubes. Water level in secondary side is maintained at 8.1 m against the pressure and temperature conditions of 101.325 kPa and 398 K respectively. Water in secondary side of SG causes condensation of the steam produced in the core. The condensate flows back to the core. This reflux condensation increases the time for the core to uncover. The results of the SG dry and wet layups are compared in Fig. 7 and Fig. 8 for two vent configurations – PZR valve Table 3 Summary of dry layup cases. Configuration
Time to start core uncover (min)
Maximum RCS pressurization (Pa)
Time for cladding to reach 1000 K (min)
Steam generator man way (SG man way) Pressurizer man way (PZR man way) Pressurizer relief valve (PZR valve) Reactor pressure vessel top vent valve (RPV top vent valve)
50.1 67.2 214.4 423.1
1.12 × 105 1.88 × 105 8.02 × 105 2.47 × 106
144.2 161.4 328.1 509.1
4
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Fig. 13. Cladding temperature for PZR valve open configuration.
Fig. 9. Coolant temperature at core inlet for open PZR valve.
Fig. 14. Cladding temperature for RPV top vent valve open configuration.
Fig. 10. Coolant temperature for RPV top vent valve open.
Fig. 11. RPV level for PZR valve open configuration.
Fig. 15. Steam flow rate in containment for PZR valve open configuration.
Fig. 12. RPV level for RPV top vent valve open configuration. Fig. 16. Steam flow rate in containment for RPV top vent valve open configuration.
5
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Table 4 Summary of dry and wet layup cases. Configuration PZR valve RPV top vent valve
Time to start core uncover (min) Dry layup Wet layup Dry layup Wet layup
Maximum RCS pressurization (Pa) 5
214.7 464.1 423.1 700
English Alphabets M Mega P Reactor Power t Time Greek Symbol
4. Conclusions Decay heat of a nuclear power plant depends upon reactor nominal thermal power, operation time of reactor and time after reactor shutdown. The RHRS plays an important role to remove decay heat of the plant during mid-loop operation for safe refueling/maintenance of the plant. Loss of the RHRS results in substantial rise in temperature/ pressure and can lead toward severe accident. The loss of RHRS can occur due to many reasons including loss of suction supply to pump, loss of electric power, accidental closures of valves, and inaccuracy in reactor vessel instrumentation and lack of knowledge between water level and RHRS pump speed. The incipience of coolant boiling on accident depends upon reactor power and coolant inventory in RPV. While time for start of core to uncover depends upon the vent configuration. The transient analysis for dry layup configurations show that time for core to uncover is larger for smaller opening area but system pressure is higher in this case. Time for core to uncover for PZR man way, PZR valve, RPV top vent valve and SG man way are 67.283 min, 297.33 min, 423.166 min and 50.133 min respectively. The maximum RCS pressure for all above configurations is 1.88 × 105Pa, 8.02 × 105Pa, 2.47 × 106 Pa and 1.12 × 105 Pa respectively. A sharp rise in cladding temperature is observed at the time of core uncovery. This rapid increase in clad temperature can lead to clad failure. The transient analysis for wet layup configurations shows that reflux condensation can increase time for core to uncover for PZR valve and RPV top vent valve but level of core is not maintained at mid-line for longer time. No reflux condensation is observed for SG man way and PZR man way because steam produced quickly leaves the system and level of RPV decreases and core is uncovered.
τ Reactor operation time Subscripts 0 s th e Units
Initial after shutdown Thermal Electric
K m W Pa s min
Kelvin meter Watt Pascal seconds minutes
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Acknowledgment The authors are grateful to their respective Departments for extending support to complete this work. Appendix A. Supplementary data Supplementary data to this article can be found online at https:// doi.org/10.1016/j.pnucene.2018.11.010. Nomenclature Abbreviations RHRS RPV PZR RCS PWR
328.1 567.9 509.1 790.1
8.02 × 10 8.23 × 105 2.47 × 106 2.35 × 106
these figures that the peak value of steam flow rate is reduced in both cases due to the effect of reflux condensation in SG U tubes. The summary of results for all wet layup cases is shown in Table 4.
Time for cladding to reach 1000 K (min)
Residual Heat Removal Reactor Pressure Vessel Pressurizer Reactor Coolant System Pressurized Power Reactor 6