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Thermal-hydraulic analysis of NSSS and containment response during extended station blackout for Maanshan PWR plant
Nuclear Engineering and Design 288 (2015) 1–18
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Nuclear Engineering and Design journal homepage: www.elsevier.com/locate/nucengdes
Thermal-hydraulic analysis of NSSS and containment response during extended station blackout for Maanshan PWR plant Yng-Ruey Yuann ∗ , Keng-Hsien Hsu, Chin-Tsu Lin Institute of Nuclear Energy Research, Atomic Energy Council, R.O.C., 1000, Wenhua Road, Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan
h i g h l i g h t s • • • • •
Calculate NSSS and containment transient response during extended SBO of 24 h. RELAP5-3D and GOTHIC models are developed for Maanshan PWR plant. Reactor coolant pump seal leakage is specifically modeled for each loop. Analyses are performed with and without secondary-side depressurization, respectively. Considering different total available time for turbine driven auxiliary feedwater system.
a r t i c l e
i n f o
Article history: Received 21 October 2014 Received in revised form 2 March 2015 Accepted 16 March 2015
a b s t r a c t A thermal-hydraulic analysis has been performed with respect to the response of the nuclear steam supply system (NSSS) and the containment during an extended station blackout (SBO) duration of 24 h in Maanshan PWR plant. Maanshan plant is a Westinghouse three-loop PWR design with rated core thermal power of 2822 MWt. The analyses in the NSSS and the containment are based on the RELAP53D and GOTHIC models, respectively. Important design features of the plant in response to SBO are considered in the respective models, e.g., the steam generator PORVs, turbine driven auxiliary feedwater system (TDAFWS), accumulators, reactor coolant pump (RCP) seal design, various heat structures in the containment, etc. In the analysis it is assumed that the shaft seal in each RCP failed due to loss of seal cooling and the RCS fluid flows to the containment directly. Some parameters calculated from the RELPA5-3D model are input to the containment GOTHIC model, including the RCS average temperature and the RCP seal leakage flow and enthalpy. The RCS average temperature is used to drive the sensible heat transfer to the containment. It is found that the severity of the event depends mainly on whether the secondary side is depressurized or not. If the secondary side is depressurized in time (within 1 h after SBO) and the TDAFWS is available greater than 19 h, then the reactor core will be covered with water throughout the SBO duration, which ensures the integrity of the reactor core. On the contrary, if the secondary side is not depressurized, then the RCS pressures will remain high in conjunction with the higher RCP seal leakage flow. The accumulators will not be available due to high RCS pressure and the reactor core will eventually become uncovered since there is no any water make-up. In the aspect of the containment response, the high-energy RCP seal leakage fluid continues flowing into the containment and heats up the containment. The containment pressure and temperature will increase to high values, respectively. There exists no clear relationship between the available TDAFWS time and the maximum containment pressure and temperature. The response of the containment temperature is much worse than that of the containment pressure. The most severe containment temperature response occurs for the case with no secondary depressurization and the calculated maximum containment temperature is 336.8 ◦ F, which exceeds the design temperature of 300 ◦ F but is still below the inside-containment safety-related equipment environmental qualification temperature of 450 ◦ F. © 2015 Elsevier B.V. All rights reserved.
∗ Corresponding author. Tel.: +886 3 4711400x6083; fax: +886 3 4711404. E-mail addresses: [email protected] (Y.-R. Yuann), [email protected] (K.-H. Hsu), [email protected] (C.-T. Lin). http://dx.doi.org/10.1016/j.nucengdes.2015.03.011 0029-5493/© 2015 Elsevier B.V. All rights reserved.
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1. Introduction Currently, the four nuclear power plants in Taiwan, including Chinshan BWR-4, Kuosheng BWR-6, Lungmen ABWR, and Maanshan PWR, are all designed to have capabilities to cope with Station Blackout (SBO) event for a duration of 8 h. After the Fukushima accident happening on March 11, 2011, the nuclear regulatory authority in Taiwan asked Taiwan Power Company, the owner of the four nuclear power plants, to perform a detailed evaluation for each plant to see whether the plant has the capability to cope with the SBO for more than 8 h, with the target of 24 h. The evaluation was based on the installed design, either existing or scheduled to be proceeded with, and the scope covered SBO Scenario, Nuclear Steam Supply System (NSSS), Condensate Storage Tank Inventory, DC Capacity, Compressed Air, Instrumentation and Control, Steam Driven Water Make-up System, Loss of Ventilation, and Containment Integrity. All items needed to be evaluated based on an appropriate analysis methodology specific to the plant. The NSSS and Containment integrity analyses were two major tasks to be performed in the SBO evaluation. The study of this paper deals with the NSSS and SBO containment response analysis specifically for Maanshan PWR plant. For the 3-loop Maanshan PWR plant (Taiwan Power Company, 2014), each reactor coolant pump (RCP) is equipped with three shaft seals arranged in series and provided with cooling to maintain
their integrity, as shown in Fig. 1. Normally, the high pressure seal injection flow is provided by the charging pumps with suction from the volume control tank (VCT). The flow enters each RCP through a connection on the thermal barrier flange. Here the flow splits into two paths. The major portion flows down the shaft and through the thermal barrier heat exchanger to cool the bearing and the shaft seal system and then enters the reactor coolant system (RCS). The remainder flows up the shaft through the seals. After passing through the #l seal it returns to the VCT through the #1 leak-off line. Minor flow passes through the #2 seal to its leak-off line and returns to the reactor coolant drain tank (RCDT). A back flush injection from a head tank flows into the #3 seal. At this point the flow divides with half flushing through one side of the seal and out the #2 seal leak-off while the remaining half flushes through the other side and out the #3 seal leak-off to the containment sump. This arrangement ensures essentially zero leakage of reactor coolant or trapped gases to the containment. In addition, the component cooling water (CCW) is provided to the thermal barrier heat exchanger. If a loss of seal injection flow should occur, the thermal barrier heat exchanger cools the reactor coolant to an acceptable level before it enters the bearing and seal area. With seal integrity maintained the total seal leakage rate can be maintained at a designated low value of approximately 3 gpm during normal operation. However, due to loss of AC power to the high pressure seal injection system and the CCW system during the SBO, there will be no cooling to
Fig. 1. Reactor coolant pump seal flow path schematics.
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the RCP seal, which will cause the seal materials to be exposed to the RCS coolant at high temperature. Under this condition, a failure mode of popping or O-ring extrusion may exist in the three seal stages (Westinghouse Electric Company, 2002). The popping is an opening of the seal faces due to hydraulic instability caused by fluid flashing. The O-ring extrusion is an overheating of the secondary sealing elastomers. Both of them will allow excessive leakage. The excessive leakage may become a concern for the NSSS since there is no high pressure make-up water for reactor pressure vessel and the RCS coolant inventory will be lost through the leakages and the core will become uncovered and begin to heat up, eventually leading to a core damage. The seal leakage is also a concern for the containment since the leakage flow at high temperature will flow to the containment and may heat up the containment to a high pressure and temperature to challenge the integrity of the containment. In view of these concerns a thermal-hydraulic analysis is performed in this paper to investigate the effects of RCP seal leakage on the NSSS and Containment during extended SBO of 24 h. Although the RCP seal may be maintained for a certain time after SBO, which causes the event to be less severe, the current analysis considered the worse condition that the RCP seal fails at the beginning of the SBO event. The NSSS analysis for the reactor core and RCS response is performed using the RELAP-3D code (INEEL, 1998) and Maanshan SBO analysis model is employed. The analysis simulates the SBO transient for 24 h with the secondary side power operated relief valve (PORV) manually opened to cool down the RCS at a rate of less than 100 ◦ F/h allowed by the plant Technical Specification, assuming various available time for the turbine driven auxiliary feedwater system (TDAFWS). The shaft seals of each RCP are assumed to fail and the RCP seal leakage is modeled with an initial leakage rate of 25 gpm for each RCP based on the value proposed in the SBO guideline NUMARC 87-00 (NUMARC, 1991). In addition, analysis is performed assuming no manual depressurization of secondary side. This case is expected to be much worse than the case with secondary depressurization since the RCS is not cooled down and remains at pressure above the accumulators initiation pressure. The accumulators will not be able to inject water to the reactor and the inventory of the reactor will be lost through each RCP seal leakage. Eventually, the reactor core will become uncovered with water and its integrity is challenged. The SBO containment response analysis for a PWR is quite different from that for a BWR. For a BWR, the containment suppression pool will receive the blowdown flow and energy from the safety relief valves which open intermittently or continuously. The severity of the pressure and temperature response of the drywell, wetwell air space and suppression pool depends on the core decay power and the heat capacity of the suppression pool which serves as a major heat sink. For a PWR, on the contrary, there is no direct blowdown flow and energy into the containment, and the only heat source is the sensible heat transfer from RCS and the reactor coolant leakage flow. The sensible heat transfer depends on the reactor coolant temperature prior to and after SBO. The reactor coolant leakage is mainly contributed by the RCP seal leakage due to loss of seal cooling during SBO and the flow rate is a function of pressure difference between the RCS pressure and the containment pressure. There is no suppression pool water heat sink, and the only heat sink is through the containment passive heat structures. In view of this, a detailed heat structure modeling is considered in this analysis. The GOTHIC code (NAI, 2012) has the capability to model the detailed heat structure with different geometries in conjunction with various heat transfer option, and therefore is used in the analysis. The sensible heat source and RCP seal leakage flow are obtained from the NSSS calculation of the RELAP5-3D and then input to the GOTHIC model.
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2. SBO scenarios with respect to NSSS and containment Maanshan plant is a twin-unit Westinghouse 3-loop PWR currently operated at rated core thermal power of 2822 MWt (the original licensed core thermal power is 2775 MWt and had been uprated to 2822 MWt through Measurement Uncertainty Recapture Power Uprating, or so called MUR PU) for each unit since 2009. Loss of offsite power will result in loss of all power to the plant auxiliaries, i.e., the RCPs, condensate pumps, circulation water pumps, etc. The main turbine will trip due to loss of condenser vacuum caused by loss of circulation water pumps. The reactor will trip due to turbine trip or loss of power to control rod drive mechanism. Although the three reactor feedwater pumps are turbine-driven it will trip due to feedwater pump low suction pressure resulting from loss of condensate pumps. A loss of condenser vacuum will preclude the use of steam dump to the condenser. The decrease in heat removal by the secondary system is accompanied by a RCS flow coastdown which further reduces the capacity of the primary coolant to remove heat from the core. As the steam system pressure rises following the turbine trip, the steam generator (SG) PORVs may automatically open to relieve steam to the atmosphere. There are two PORVs per SG. If PORV is not available, the steam generator safety valves may lift to dissipate the sensible heat of the fuel and coolant plus the core decay heat. However, the safety vales are not considered in the present analysis. As the no-load temperature is approached, the SG PORVs are used to dissipate the core decay heat. The water level in the three steam generators will decrease continuously due to loss of normal feedwater flow. The TDAFWS will start on low-low level signal in any two steam generators. It will also be stared on low voltage in 4.16 kV class 1E bus. The pump takes suction from the condensate storage tank and is designed to supply rated flow within 1 min of the initiating signal. The water level in each SG will be recovered by the water injection of the TDAFWS. The plant emergency operating procedure (EOP) requires the operator manually depressurize the steam generators and perform the RCS cool-down at maximum rate but not exceeding 100 ◦ F/h. The RCS pressure and temperature will decrease continuously and the accumulators will begin to inject water into the RCS when the RCS pressure decreases to the accumulator initiation setpoint of 600 psia. If no manual depressurization of the secondary side is taken, the RCS is not cooled down and will remain at a relatively high pressure, which will cause the RCP seal leakage rate to be much higher. The RCS pressure will not decrease to the setpoint of accumulators initiation and the accumulator is not available for this case. The core will eventually become uncovered due to continuous RCP seal leakage. The containment is a dry type with design pressure and temperature of 74.7 psia and 300 ◦ F, respectively. There are two containment sumps, separated by 90◦ , inside the containment, each is located 6.67 ft below the plant grade. During normal operation, three of the four fan cooler units operate continuously to remove heat from the containment atmosphere to maintain the average temperature below 120 ◦ F and the relative humidity below 70%. During SBO all fan cooler units will be tripped due to loss of AC power. The containment will be heated up by the reactor coolant sensible heat and the energy of the incoming RCP seal leakage flow. The only heat sink to slow down the heating-up process is through the passive containment heat structures. For the case with manual SG depressurization using PORV the RCS leakage flow will decrease to a relatively low value since the RCS pressure is decreasing continuously. The sensible heat will also decrease to a relatively low value since the RCS average temperature is decreasing and the containment temperature is increasing. For the case without secondary depressurization, the situation may differ much from the case with secondary depressurization. The RCS has a very limited cool-down through the intermittent open and close of SG PORVs. This causes
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the RCS temperature and pressure to remain at high value. The RCP leakage flow rate will be higher than that corresponding to the case with secondary depressurization. Higher RCP leakage flow rate results in an adverse effect on the containment temperature and pressure response and may challenge the design values. 3. RELAP5-3D NSSS model RELAP5-3D is used to simulate the SBO NSSS transient and the nodalization schematics is shown in Fig. 2. The report (Yuann, 2014) can be referenced for the detailed description of the nodalization schemes. In summary, there are three loops and only loop 1 and loop 2 are shown in the diagram. Loop 2 is a loop with pressurizer and loop 3 is identical to loop 1. Major components included in the model are reactor pressure vessel (RPV), core and bypass region, fuel bundles, RCPs, pressurizer, accumulators, SG U tubes and shell sides, main steam lines, pressurizer PORV and safety valves, SG PORVs and safety valves, main steam isolation valves, turbine stop valves, main feedwater, TDAFWS, and RCP seal leakage path. The components are modeled as volume, junction or heat structure with identification number shown in Table 1. The system and components not available during SBO are not modeled. The initial conditions and major assumptions associated with the model is described below. (1) Initial conditions reactor core power: 2830.5 MWt pressurizer pressure: 2250 psia RCS average temperature: 587.8 ◦ F RCS flow: 92,600 gpm/loop SG outlet pressure: 942 psia total steam mass flow rate: 12.58 Mlbm/h.
Table 1 RELAP5-3D model volume and junction identification number. Component name
RELAP5-3D model identification number
RPV upper part, downcomer, and lower part Core
Volumes 170, 172, 174, 190, 192, 32, 40, 42, 62, 90, 92, 30, 60, 2, 12, 22 Volumes 122, 124, 126, 128, 130, 132, 134, 136 Volume 100–116 Volume 320 Loop 1: Volume 290, 292, 294, 298 Loop 2: Volume 390, 392, 394, 398 Loop 3: Volume 490, 492, 494, 498 Loop 1: Volume 280, Loop 2: 380, Loop 3: 480 SG1: Volume 250, SG2: 350, SG3: 450 SG1: Volume 520, SG2: 620, SG3: 720 SG1: Volume 532, SG2: 632, SG3: 732 SG1: Volume 540, 542, 544, SG2: Volume 640, 642, 644, SG3: Volume 740, 742, 744 Volume 760 Junction 981, 982, 983 Junction 984 SG1: Junction 556, 558, SG2: 656, 658, SG3: 756, 758 SG1: Junction 546, 548, 550, 552, 554 SG2: Junction 646, 648, 650, 652, 654 SG3: Junction 746, 748, 750, 752, 754 SG1: Junction 543, SG2: 643, SG3: 743 Junction 775 SG1: Junction 505, SG2: 605, SG3: 705 SG1: Junction 503, SG2: 603, SG3: 703
Core bypass regions Pressurizer RCS cold legs
RCPs SG U tubes SG shell sides SG dome Main steam lines
Main steam header Pressurizer PORVs Pressurizer safety valve SG PORVs SG safety valves
MSIVs Turbine stop valve Main feedwater flow Turbine driven auxiliary feedwater flow RCP seal leakage paths
RCP1: Junction 941, RCP2: 942, RCP3: 943
(2) TDAFWS mass flow rate provided to each SG is assumed as 33.82 lbm/s (Taiwan Power Company, 2014). (3) Shaft seals in each RCP are assumed to fail and the initial RCP seal leakage flow is assumed as 25 gpm based on based on NUMARC-87-00 (NUMARC, 1991). The RCP seal leakage junction flow area is adjusted appropriately to have this flow rate at conditions of t = 0 s. The back pressure (containment pressure) for the RCP seal leakage flow modeling is conservatively assumed as 14.7 psia. (4) Three accumulators start to inject water into RCS when RCS pressure drops below 600 psia. (5) The turbine trip will cause the SG pressure to increase to the PORV opening setpoint and PORV is expected to open and close intermittently. The analysis assumes one PORV of each SG is manually opened to depressurize the SG such that the cooling down rate of the primary system in the first hour is approximately at the maximum value of 100 ◦ F/h allowed by the plant Technical Specification, and then the pressure is controlled at ∼200 psia.
4. GOTHIC containment model GOTHIC (NAI, 2012) is a general purpose thermal-hydraulics computer code which can be used to analyze a simple separate heat transfer problem to a complicated containment configurations involved in traditional or advanced light water nuclear power plants. The GOTHIC containment model used to calculate the pressure and temperature response during SBO of duration 24 h for Maanshan PWR plant is shown in Fig. 3. The model consists of three control volumes, which are volume 1: containment, volume 2: containment sump, and volume 3: environment. There are three flow paths. Flow path 1 simulates the RCP seal leakage resulting from seal damage due to loss of seal cooling during the SBO. Flow path 2 simulates the RCS normal leakage flow, including unidentified and identified leakages. Flow path 3 simulates the flow from containment floor entering the containment sump. As described in Sections 1 and 2 there exists a sensible heat transferring from the RCS to the containment air space. This sensible heat is simulated as a heat source in control volume 1. Also the passive heat structures of the containment are classified into 18 categories, as shown in Table 2. Of these 18 heat sinks, three are modeled as external conductors to transfer heat between the containment and the environment through the containment wall. The three heat sinks are heat sink #1 containment dome and part of wall, #2 part of containment with carbon steel lining of 0.6-in. thickness, and #3 part of containment with carbon lining of 1.6-in. thickness. Four heat sinks are assumed to be covered with the containment sump water on one side and thermally insulated on the outside surface where exposed to earth. The four heat sinks are #4 part of containment floor with stainless steel, #5 part of containment floor with epoxy, #6 part of containment floor with inorganic zinc, and #7 part of containment floor with bare concrete. These four heat sinks are also modeled as internal conductors. The remaining 11 heat sinks are entirely within the containment and are modeled as internal conductors to transfer heat between the heat structures and the containment air space. The initial conditions and major assumptions associated with the model is described below.
(1) Initial conditions a. Volume 1 (containment) pressure = 14.7 psia, temperature = 120 ◦ F, relative humidity = 100% b. Volume 2 (containment sump)
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Fig. 2. Maanshan PWR RELAP-3D SBO model nodalization schematics.
Table 2 Containment passive heat structures and data (Taiwan Power Company, 2014). Heat sink #
Designation
Material and thickness (ft)
1
Containment dome and part of wall
2
Part of containment (carbon steel lining = 0.6 in.)
3
Part of containment (carbon steel lining = 1.5 in.)
4
Part of containment floor with stainless steel
5 6
Part of containment floor with epoxy paint Part of containment floor with inorganic zinc
Inorganic zinc (0.00038), carbon steel (0.02084), air (0.00035), concrete (3.8427) Inorganic zinc (0.00038), carbon steel (0.05), air (0.00035), concrete (4.0) Inorganic zinc (0.00038), carbon steel (0.125), air (0.00035), concrete (4.0) Stainless steel (9.92083), air (0.000348), concrete (8.25) Epoxy paint (0.02417), concrete (1.94042) Epoxy paint (0.00038), inorganic zinc (0.00038), carbon steel (0.02083), air (0.00035), concrete (8.25) Concrete (8.25) Epoxy paint (0.00033), concreter (1.76417) Epoxy paint (0.000242), concreter (1.82358) Stainless steel (0.02083), air (0.00035), concrete (2.11167) Zinc galvanizing (0.00017), carbon steel (0.005), air (0.00035), concrete (2.34) Stainless steel (0.022) Inorganic zinc (0.00375), carbon steel (0.06558) Inorganic zinc (0.00038), carbon steel (0.02225) Inorganic zinc (0.00038), carbon steel (0.17018) Zinc galvanizing (0.00025), carbon steel (0.01042) Epoxy paint (0.001), carbon steel (0.0345) Stainless steel (0.082)
7 8 9 10
Part of containment floor (bare concrete) Concrete interior wall Concrete floor slabs and walls Stainless steel lined concrete
11
Galvanized steel lined concrete
12 13
Miscellaneous stainless steel less than 0.5 in. Miscellaneous carbon steel less than 1 in. with inorganic coating
14
Miscellaneous carbon steel less than 0.5 in. with inorganic coating
15
Miscellaneous carbon steel greater than 1 in. with inorganic coating
16
Miscellaneous carbon steel with galvanizing
17 18
Miscellaneous carbon steel with epoxy coating Miscellaneous stainless steel greater than 0.5 in.
Exposed surface area (ft2 ) 71,946 5978 2891 490 119,559 4312
10,735 59,755 11,400 6935 7350 9755 44,042 108,271 13,913 52,031 5240 5380
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Fig. 3. Maanshan PWR GOTHIC containment model for SBO analysis.
pressure = 14.7 psia, temperature = 120 ◦ F, relative humidity = 100% c. Volume 3 (environment) pressure = 14.7 psia, temperature = 98 ◦ F, relative humidity = 80% (2) Flow boundary conditions a. Flow boundary condition 1 (RCP seal leakage) The leakage flow rate and enthalpy is obtained from the RELPA5-3D calculation. b. Flow boundary condition 2 (normal RCS leakage) The flow rate is assumed to be constant value of 11 gpm per plant technical specification. The enthalpy is assumed to be corresponding to the RCS average enthalpy during SBO and is obtained from RELAP5-3D calculation. (3) Heat source The sensible heat transfer rate from the RCS to the containment is determined by Eq. (1) qfull power ×
Tavg − TCTMT (Tavg − TCTMT )full power
(1)
where Tavg is the RCS average temperature and TCTMT is the containment airspace temperature, qfull power is the fullpower sensible heat transfer rate. At any time step, Tavg is obtained from the RELAP5-3D calculation and TCTMT is from the GOTHIC calculation at the previous time step. (4) The thickness and exposed surface area for each of the 18 passive heat sinks is shown in Table 2 (Taiwan Power Company, 2014). The properties of the different materials
Table 3 Properties of materials associated with the structural passive heat sink (Taiwan Power Company, 2014). Material
Density (lbm/ft3 )
Specific heat (Btu/lbm-◦ F)
Thermal conductivity (Btu/h-ft-◦ F)
Carbon steel Stainless steel Zinc galvanizing Concrete Epoxy paint Inorganic zinc Air
490.0 490.0 445.0 143.0 115.0 170.0 0.06
0.11 0.11 0.09 0.21 0.27 0.12 0.171
25.0 10.0 65.0 40.8 0.5 1.0 0.0174
associated with the heat structures, including density, specific heat, and thermal conductivity, are shown in Table 3 (Taiwan Power Company, 2014). (5) For external conductors #1, 2, and 3, a conservative heat transfer coefficient of 2.0 Btu/h − ft2 − ◦ F, taken form reference (Taiwan Power Company, 2014), is assumed to exist on outside surfaces (exposed to ambient, volume 3), and a combination of condensation heat transfer with any convection and radiation heat transfer is specified for the internal surface (exposed to the containment airspace, volume 1). UCHIDA condensation correlation (Uchida et al., 1965) is used. (6) For internal conductors #4, 5, 6, and 7, a conservative convective heat transfer coefficient of 0.4 Btu/h − ft2 − ◦ F, taken form reference (Taiwan Power Company, 2014), is assumed
Y.-R. Yuann et al. / Nuclear Engineering and Design 288 (2015) 1–18 Table 4 Analysis cases for the NSSS and the containment. Case ID
Description
A1
SBO with secondary depressurization and TDAFWS available for 0 h SBO with secondary depressurization and TDAFWS available for 4 h SBO with secondary depressurization and TDAFWS available for 8 h SBO with secondary depressurization and TDAFWS available for 12 h SBO with secondary depressurization and TDAFWS available for 16 h SBO with no secondary depressurization and TDAFWS available for 24 h
A2 A3 A4 A5 B
to apply between the sump water and the submerged heat sink surface (exposed to containment sump, volume 2), and insulation is assumed on the surface exposed to earth. For internal conductors #8–#18, a combination of condensation heat transfer with any convection and radiation heat transfer is specified for one surface (exposed to containment airspace, volume 1), and insulation is assumed for another surface. 5. NSSS analysis and results Two categories, categories A and B, of the SBO scenario are considered in this study. Category A simulates the transient with secondary depressurization, and different available hours, from 0 to 24 h, for the TDAFWS are assumed. Category B simulates the transient without secondary depressurization, and the TDAFWS is assumed to be available for 24 h. Totally, six cases are analyzed, including five for category A, and one for category B, as shown in Table 4. The purpose of the analysis for the cases of category
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B is to show the importance of adopting secondary depressurization during SBO, therefore, only one case with TDAFWS available for the whole duration of SBO (24 h) is considered. The category A cases assume that operator uses one PORV of each SG to manually depressurize the SGs and perform cooling down of the RCS at maximum rate of 100 ◦ F/h during the first 1 h after SBO. After the first hour, the SG pressure is controlled at a pressure near 200 psia. The transient response of the key system parameters, including SG pressure, SG water level, RCS pressure, RCS average temperature, RCP seal leakage flow rate and quality, and reactor water level, for the selected Cases A2, A3, and A4 are shown together in Figs. 4–11, respectively. The event sequence for the Case A4 (SBO with secondary depressurization and TDAFWS available for 12 h) is used as an illustration example to explain the transient behavior during the SBO. Fig. 4 shows the #1 SG pressure response. In the first hour, the SG PORV is manually opened to achieve maximum cooldown of the RCS, which causes the SG pressure to decrease sharply. Then the SG pressure is controlled at approximately 200 psia. Since TDAFWS is assumed to trip at t = 12 h (43,200 s), the SG water level begins to decrease after this time and drop to the bottom of the narrow the narrow range level span, as shown in Fig. 5. Due to termination of cold water injection from TDAFWS, the SG pressure stops decreasing and starts to increase. However, the increase of the SG pressure is limited by the continuous water inventory decrease. Between t = 60,000 s and 75,000 s, there exists a pressure oscillation. The pressure response for the other two SGs will behave in the same manner. The SG level will continuously decrease and eventually drop to the bottom of the U-tube region, as can be seen from the wide rang level response in Fig. 5. Starting from t = 0 s, the RCS pressure and temperature will decrease continuously since it is cooled down through the depressurization of the secondary side. At t ∼ 68,100 s, there is a sharp increase of the RCS pressure and temperature, as shown in Figs. 6 and 7, since at this time the SG level has dropped to near the bottom of the SG and the heat removal rate
Fig. 4. #1SG pressure response during SBO for Cases A2, A3, and A4.
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Fig. 5. #1SG water level response during SBO for Cases A2, A3, and A4.
Fig. 6. RCS pressure response during SBO for Cases A2, A3, and A4.
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Fig. 7. RCS average temperature response during SBO for Cases A2, A3, and A4.
Fig. 8. RCP seal leakage liquid mass flow rate per pump during SBO for Cases A2, A3, and A4.
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Fig. 9. Equilibrium quality of the upstream volume of RCP seal leakage junction during SBO for Cases A2, A3, and A4.
Fig. 10. RCP seal leakage vapor mass flow rate per pump during SBO for Cases A2, A3, and A4.
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Fig. 11. RPV water level response during SBO for Cases A2, A3, and A4.
Fig. 12. RCS pressure response during SBO for Case B.
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Fig. 13. RPV water level response during SBO for Case B.
from the primary side to the secondary side is significantly reduced. The RCP liquid leakage mass flow rate in each loop decreases with decrease of the RCS pressure until t ∼ 60,000 s and increases with the increase of the RCS pressure after t ∼ 60,000 s, as shown in Fig. 8. The upstream volume of the RCP seal leakage junction becomes a two-phase mixture at t ∼ 78,920 s, as shown in the equilibrium quality plot (Fig. 9). After this time, due to low density of the twophase mixture flow, the RCP seal leakage liquid mass flow rate drops sharply, as shown in Fig. 8. It is also observed that a vapor flow is appearing after t = 78,920 s, as shown in Fig. 10. To be noted is that there is a momentary leakage mass flow rate increasing at the early transient (the secondary peak amidst the flow decrease) at t = 1676 s. This is explained as follows: the leakage flow path is opened at t = 0 s, and the mass flow rate is determined by critical flow model. The pressure decreasing has a negative contribution to the critical mass flow rate, while the temperature decreasing has a positive contribution to the critical mass flow rate. Fig. 6 shows that the pressure decreases sharply initially, then there is a significant slope change, i.e., the pressure decreasing rate slows down This causes the mass flow decreasing rate to become slower. Meanwhile, Fig. 7 shows that the temperature decreases continuously, which increase the mass flow rate. At the time when the pressure decreasing rate turns to slower, the positive contribution exceeds the negative contribution, which causes the leakage mass flow rate to increase momentarily. Fig. 11 shows the reactor water level response. The reactor water level will be maintained by the accumulator flow before ∼44,040 s. After this time the accumulator becomes unavailable since accumulator pressure has dropped to below the RCS pressure. The reactor water level then starts to decrease due to existence of continuous RCP seal leakage flow. The first time the reactor water level drops to top of active fuel (TAF) is at t ∼ 60,840 s (16.9 h). The time it takes for the reactor water level to drop to below TAF for the six cases are shown in Table 5. It can be extrapolated from this table that, for the case with secondary
Table 5 Calculated time to top of active fuel (TAF) for different cases. Case #
A1 (with secondary depressurization) A2 (with secondary depressurization) A3 (with secondary depressurization) A4 (with secondary depressurization) A5 (with secondary depressurization) B (no secondary depressurization)
Total available time for TDAFWS (h) (referenced to beginning of SBO)
Calculated time to TAF (h) (referenced to beginning of SBO)
0
1.52
4
9.21
8
13.42
12
16.90
16
21.15
24
6.50
depressurization, to prevent the water level from dropping below TAF during the SBO period of 24 h, the TDAFWS must be available for at least 19 h. For the case with no secondary depressurization (Case B), the reactor water level will drop to TAF in 7 h even if the TDAFWS is available for 24 h. This is because the RCS is not cooled down and remains at a relatively high pressure, as shown in Fig. 12. The RCS pressure will not decrease to the setpoint of accumulators initiation (600 psia) and accumulator is not available. The core eventually becomes uncovered due to existence of continuous RCP seal leakage, as shown in Fig. 13. The analyses in this section are based on the assumptions that the shaft seals in each RCP fail initially due to loss of cooling and the seal leakage flow rate is assumed to be 25 gpm initially per RCP. However, the RCP seal may not fail during the SBO. To investigate the effect of the RCP seal leakage, analyses have been performed by deleting the leakage flow path to simulate the case without RCP
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Fig. 14. RCS pressure response during SBO for Cases A4 and A4.1.
Fig. 15. RPV water level response during SBO for Cases A4 and A4.1.
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Fig. 16. RCS pressure response during SBO for Cases B and B.1.
Fig. 17. RPV water level response during SBO for Cases B and B.1.
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seal leakages. Two cases, Cases A4.1 and B.1, of no RCP seal leakages are considered, with the former modified from case A4 and the latter modified from case B, respectively. Transient response for the major parameters, including RCS pressure and reactor water level, are compared, between Cases A4 and A4.1, and between Case B and B.1, as shown in Figs. 14–17. The results show that for the case with secondary depressurization and no RCP seal leakage (Case A4.1), even the RCS pressure is higher than that of Case A4, the RPV water level will not drop to TAF during the SBO of 24 h. For the case with no secondary depressurization and no RCP seal leakage (Case B.1), the RCS pressure will cycle at the pressurizer safety valve opening setpoint (2500 psia), and the RPV level will not drop to TAF. However, for the case B, even the RCS pressure is lower than that of Case B.1, the RPV water level will drop to below TAF due to continuous RCS inventory loss through the RCP seal leakages. 6. Containment analysis and results A containment analysis is performed for the SBO of 24 h based on the GOTHIC model described in Section 4. The containment pressure and temperature response for the six SBO cases discussed in Section 5 are analyzed. And the case A.4 (SBO with secondary depressurization and TDAFWS available for 12 h) is used as an illustration example. The calculated pressure and temperature response for the containment air space are shown in Figs. 18 and 19, respectively. After SBO occurs, heat is added to the containment through the sensible heat transfer, the RCS normal leakage flow and the RCP seal leakage flow. The heat addition is larger than the heat transferring out through various heat structures and the pressure and temperature of the containment air space increase continuously until t = 11,400 s. At this time both the pressure and the temperature reach the first peak values and start to decrease. The drop of the pressure and the temperature is explained below.
15
As shown in Fig. 20, there is a significant phase change to generate more steam around 11,400 s, which leads to more condensation on the colder surfaces. This in turn enhances the heat transfer to the heat structures. It can also be observed in Figs. 21 and 22 that there is a significant increase in heat transfer coefficient and heat transfer rate from the containment to the #1 heat structure surface around 11,400 s (likewise for other heat structures but only #1 heat structure result is shown). This explains why the containment pressure and temperature cease to increase at 11,400 s and have a drop of values at that time. At t ∼ 68,100 s the RCS pressure rises again significantly (as shown in Fig. 6) and causes the RCP seal leakage liquid flow to have a large increase (as shown in Fig. 8). This in turn adds more energy into the containment and causes the containment pressure and temperature to increase sharply until they reach the second peak values, respectively. After this time the containment pressure and temperature start to decrease again since the upstream volumes of the RCP seal leakage junction have been full of low-density vapor and only a very small leakage vapor flow rate exists, which is not large enough to heat up the containment. The containment maximum pressure (the second peak value) calculated is 21.8 psia, which is far below the design pressure of 74.7 psia. The containment maximum temperature (the second peak value) calculated is 329.0 ◦ F, which is higher than the design temperature of 300.0 ◦ F. The maximum containment pressure and temperature calculated for the six cases are summarized in Table 6. It shows that, for the case with secondary depressurization, there exists no clear relationship between the available TDAFWS time and the calculated maximum containment pressure and temperature. This is because the containment pressure and temperature responses are determined by the complicated and combined effects of the RCP seal leakage flow plus energy and the sensible heat transfer due to the higher RCS temperature. The RCP leakage flow rate and fluid energy depend on the reactor coolant state. The reactor coolant
Fig. 18. Containment pressure response during SBO for Case A4.
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Fig. 19. Containment temperature response during SBO for Case A4.
Fig. 20. Phase change rate in containment air space during SBO for Case A4.
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Fig. 21. Heat transfer coefficient on the containment side of heat conductor 1 for Case A4.
Fig. 22. Heat transfer rate on the containment side of heat conductor 1 for Case A4.
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Table 6 Calculated maximum containment pressure and temperature for different cases. Case #
Total available time for TDAFWS (h) (referenced to beginning of SBO)
Calculated maximum containment pressure (psia)
Calculated maximum containment temperature (◦ F)
A1 (with secondary depressurization) A2 (with secondary depressurization) A3 (with secondary depressurization) A4 (with secondary depressurization) A5 (with secondary depressurization) B (no secondary depressurization)
0 4 8 12 16 24
26.3 22.0 21.7 21.8 20.1 22.1
284.1 321.1 327.8 329.0 296.7 336.8
may become two-phase mixture earlier or later depending on the available time for TDAFWS assumed in the NSSS analysis. The maximum containment pressure and temperature for the case with no secondary depressurization, i.e., case B, are 22.1 psia and 336.8 ◦ F, respectively. 7. Conclusion The extended SBO of 24 h has been analyzed with respect to the response of the NSSS and containment for the Maanshan PWR plant. The RELAP5-3D model is used to calculate the NSSS response, and the GOTHIC model is used to calculate the containment response. The shaft seals of the three RCPs are assumed to fail due to loss of seal cooling. The containment model is linked to the RELAP53D model through the RCS sensible heat transfer and the RCP seal leakage flow. Analysis results show that if the RCS is cooled down in time (within an hour) during the SBO through the manual depressurization of the SGs, then there will be no challenge to the reactor core integrity if the TDAFWS is available for at least 19 h. On the contrary, if there is no any SG depressurization action taken during the SBO, then the core will become uncovered and the integrity of the core is challenged, even if the TDAFWS is available for the entire SBO period of 24 h. This is because the RCS is not cooled down and remains at a relatively high pressure and temperature, which causes the RCP seal leakage rate to be much higher. The RCS pressure will not decrease to the setpoint of accumulators initiation and the accumulator is not available. The core eventually becomes uncovered due to existence of continuous RCP seal leakage. In the aspect of containment response, the high-energy RCP seal leakage fluid continues flowing into the containment and heats up the containment. The containment pressure and temperature increase to a high values. There exists no clear relationship between the available TDAFWS time and the calculated maximum
containment pressure and temperature. The containment pressure response is mild with the maximum value of 20–26 psia, which is far below the design value of 74.7 psia and is not a concern. However, the response of the containment temperature is much worse than that of the containment pressure. The most severe containment temperature response occurs for the case with no secondary depressurization and the calculated maximum containment temperature for this case is 336.8 ◦ F, which exceeds the design temperature of 300 ◦ F but is still below the inside-containment safety-related equipment environmental qualification temperature of 450 ◦ F. The study identifies some requisites needed to cope with the extended SBO of 24 h, such as (1) acknowledging the importance of conducting in-time action to depressurize the secondary side, (2) assuring the operability of existing TDAFWS can be maintained for at least 19 h, and (3) enhancing the RCP seal design to minimize the potential seal leakage flow during loss of seal cooling. References Idaho National Engineering and Environmental Laboratory (INEEL), 1998. RELAP53D Code Manual, INEEL-EXT-98-00834. Nuclear Management and Resources Council, Inc. (NUMARC), 1991, August. Guideline and Technical Bases for NUMARC Initiatives Addressing Station Blackout at Light Water Reactors, NUMARC 87-00. Numerical Applications Inc., 2012, January. GOTHIC Thermal Hydraulic Analysis Package Technical Manual, Version 8.0(QA), NAI 8907-02 Rev 20. Taiwan Power Company, 2014. Final Safety Analysis Report for Maanshan Nuclear Power Station Units 1 and 2 (FSAR). Uchida, H., et al., 1965. Evaluation of Post-Incident Cooling Systems of Light Water Power Reactors. In: Third International Conference on the Peaceful Uses of Atomic Energy, New York, A/Conf/28/P/436, l965. 14. Westinghouse Electric Company, 2002, May. WOG 2000 Reactor Coolant Pump Seal Leakage Model for Westinghouse PWRs, WCAP-15603. Yuann, Y.R., 2014. Maanshan PWR RELAP3-D Nodalization and Model Description, and Steady State Initiation, NED-THA-03A16806-REP-009-01. Institute of Nuclear Energy Research, Atomic Energy Council, R.O.C.
Contents lists available at ScienceDirect
Nuclear Engineering and Design journal homepage: www.elsevier.com/locate/nucengdes
Thermal-hydraulic analysis of NSSS and containment response during extended station blackout for Maanshan PWR plant Yng-Ruey Yuann ∗ , Keng-Hsien Hsu, Chin-Tsu Lin Institute of Nuclear Energy Research, Atomic Energy Council, R.O.C., 1000, Wenhua Road, Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan
h i g h l i g h t s • • • • •
Calculate NSSS and containment transient response during extended SBO of 24 h. RELAP5-3D and GOTHIC models are developed for Maanshan PWR plant. Reactor coolant pump seal leakage is specifically modeled for each loop. Analyses are performed with and without secondary-side depressurization, respectively. Considering different total available time for turbine driven auxiliary feedwater system.
a r t i c l e
i n f o
Article history: Received 21 October 2014 Received in revised form 2 March 2015 Accepted 16 March 2015
a b s t r a c t A thermal-hydraulic analysis has been performed with respect to the response of the nuclear steam supply system (NSSS) and the containment during an extended station blackout (SBO) duration of 24 h in Maanshan PWR plant. Maanshan plant is a Westinghouse three-loop PWR design with rated core thermal power of 2822 MWt. The analyses in the NSSS and the containment are based on the RELAP53D and GOTHIC models, respectively. Important design features of the plant in response to SBO are considered in the respective models, e.g., the steam generator PORVs, turbine driven auxiliary feedwater system (TDAFWS), accumulators, reactor coolant pump (RCP) seal design, various heat structures in the containment, etc. In the analysis it is assumed that the shaft seal in each RCP failed due to loss of seal cooling and the RCS fluid flows to the containment directly. Some parameters calculated from the RELPA5-3D model are input to the containment GOTHIC model, including the RCS average temperature and the RCP seal leakage flow and enthalpy. The RCS average temperature is used to drive the sensible heat transfer to the containment. It is found that the severity of the event depends mainly on whether the secondary side is depressurized or not. If the secondary side is depressurized in time (within 1 h after SBO) and the TDAFWS is available greater than 19 h, then the reactor core will be covered with water throughout the SBO duration, which ensures the integrity of the reactor core. On the contrary, if the secondary side is not depressurized, then the RCS pressures will remain high in conjunction with the higher RCP seal leakage flow. The accumulators will not be available due to high RCS pressure and the reactor core will eventually become uncovered since there is no any water make-up. In the aspect of the containment response, the high-energy RCP seal leakage fluid continues flowing into the containment and heats up the containment. The containment pressure and temperature will increase to high values, respectively. There exists no clear relationship between the available TDAFWS time and the maximum containment pressure and temperature. The response of the containment temperature is much worse than that of the containment pressure. The most severe containment temperature response occurs for the case with no secondary depressurization and the calculated maximum containment temperature is 336.8 ◦ F, which exceeds the design temperature of 300 ◦ F but is still below the inside-containment safety-related equipment environmental qualification temperature of 450 ◦ F. © 2015 Elsevier B.V. All rights reserved.
∗ Corresponding author. Tel.: +886 3 4711400x6083; fax: +886 3 4711404. E-mail addresses: [email protected] (Y.-R. Yuann), [email protected] (K.-H. Hsu), [email protected] (C.-T. Lin). http://dx.doi.org/10.1016/j.nucengdes.2015.03.011 0029-5493/© 2015 Elsevier B.V. All rights reserved.
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1. Introduction Currently, the four nuclear power plants in Taiwan, including Chinshan BWR-4, Kuosheng BWR-6, Lungmen ABWR, and Maanshan PWR, are all designed to have capabilities to cope with Station Blackout (SBO) event for a duration of 8 h. After the Fukushima accident happening on March 11, 2011, the nuclear regulatory authority in Taiwan asked Taiwan Power Company, the owner of the four nuclear power plants, to perform a detailed evaluation for each plant to see whether the plant has the capability to cope with the SBO for more than 8 h, with the target of 24 h. The evaluation was based on the installed design, either existing or scheduled to be proceeded with, and the scope covered SBO Scenario, Nuclear Steam Supply System (NSSS), Condensate Storage Tank Inventory, DC Capacity, Compressed Air, Instrumentation and Control, Steam Driven Water Make-up System, Loss of Ventilation, and Containment Integrity. All items needed to be evaluated based on an appropriate analysis methodology specific to the plant. The NSSS and Containment integrity analyses were two major tasks to be performed in the SBO evaluation. The study of this paper deals with the NSSS and SBO containment response analysis specifically for Maanshan PWR plant. For the 3-loop Maanshan PWR plant (Taiwan Power Company, 2014), each reactor coolant pump (RCP) is equipped with three shaft seals arranged in series and provided with cooling to maintain
their integrity, as shown in Fig. 1. Normally, the high pressure seal injection flow is provided by the charging pumps with suction from the volume control tank (VCT). The flow enters each RCP through a connection on the thermal barrier flange. Here the flow splits into two paths. The major portion flows down the shaft and through the thermal barrier heat exchanger to cool the bearing and the shaft seal system and then enters the reactor coolant system (RCS). The remainder flows up the shaft through the seals. After passing through the #l seal it returns to the VCT through the #1 leak-off line. Minor flow passes through the #2 seal to its leak-off line and returns to the reactor coolant drain tank (RCDT). A back flush injection from a head tank flows into the #3 seal. At this point the flow divides with half flushing through one side of the seal and out the #2 seal leak-off while the remaining half flushes through the other side and out the #3 seal leak-off to the containment sump. This arrangement ensures essentially zero leakage of reactor coolant or trapped gases to the containment. In addition, the component cooling water (CCW) is provided to the thermal barrier heat exchanger. If a loss of seal injection flow should occur, the thermal barrier heat exchanger cools the reactor coolant to an acceptable level before it enters the bearing and seal area. With seal integrity maintained the total seal leakage rate can be maintained at a designated low value of approximately 3 gpm during normal operation. However, due to loss of AC power to the high pressure seal injection system and the CCW system during the SBO, there will be no cooling to
Fig. 1. Reactor coolant pump seal flow path schematics.
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the RCP seal, which will cause the seal materials to be exposed to the RCS coolant at high temperature. Under this condition, a failure mode of popping or O-ring extrusion may exist in the three seal stages (Westinghouse Electric Company, 2002). The popping is an opening of the seal faces due to hydraulic instability caused by fluid flashing. The O-ring extrusion is an overheating of the secondary sealing elastomers. Both of them will allow excessive leakage. The excessive leakage may become a concern for the NSSS since there is no high pressure make-up water for reactor pressure vessel and the RCS coolant inventory will be lost through the leakages and the core will become uncovered and begin to heat up, eventually leading to a core damage. The seal leakage is also a concern for the containment since the leakage flow at high temperature will flow to the containment and may heat up the containment to a high pressure and temperature to challenge the integrity of the containment. In view of these concerns a thermal-hydraulic analysis is performed in this paper to investigate the effects of RCP seal leakage on the NSSS and Containment during extended SBO of 24 h. Although the RCP seal may be maintained for a certain time after SBO, which causes the event to be less severe, the current analysis considered the worse condition that the RCP seal fails at the beginning of the SBO event. The NSSS analysis for the reactor core and RCS response is performed using the RELAP-3D code (INEEL, 1998) and Maanshan SBO analysis model is employed. The analysis simulates the SBO transient for 24 h with the secondary side power operated relief valve (PORV) manually opened to cool down the RCS at a rate of less than 100 ◦ F/h allowed by the plant Technical Specification, assuming various available time for the turbine driven auxiliary feedwater system (TDAFWS). The shaft seals of each RCP are assumed to fail and the RCP seal leakage is modeled with an initial leakage rate of 25 gpm for each RCP based on the value proposed in the SBO guideline NUMARC 87-00 (NUMARC, 1991). In addition, analysis is performed assuming no manual depressurization of secondary side. This case is expected to be much worse than the case with secondary depressurization since the RCS is not cooled down and remains at pressure above the accumulators initiation pressure. The accumulators will not be able to inject water to the reactor and the inventory of the reactor will be lost through each RCP seal leakage. Eventually, the reactor core will become uncovered with water and its integrity is challenged. The SBO containment response analysis for a PWR is quite different from that for a BWR. For a BWR, the containment suppression pool will receive the blowdown flow and energy from the safety relief valves which open intermittently or continuously. The severity of the pressure and temperature response of the drywell, wetwell air space and suppression pool depends on the core decay power and the heat capacity of the suppression pool which serves as a major heat sink. For a PWR, on the contrary, there is no direct blowdown flow and energy into the containment, and the only heat source is the sensible heat transfer from RCS and the reactor coolant leakage flow. The sensible heat transfer depends on the reactor coolant temperature prior to and after SBO. The reactor coolant leakage is mainly contributed by the RCP seal leakage due to loss of seal cooling during SBO and the flow rate is a function of pressure difference between the RCS pressure and the containment pressure. There is no suppression pool water heat sink, and the only heat sink is through the containment passive heat structures. In view of this, a detailed heat structure modeling is considered in this analysis. The GOTHIC code (NAI, 2012) has the capability to model the detailed heat structure with different geometries in conjunction with various heat transfer option, and therefore is used in the analysis. The sensible heat source and RCP seal leakage flow are obtained from the NSSS calculation of the RELAP5-3D and then input to the GOTHIC model.
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2. SBO scenarios with respect to NSSS and containment Maanshan plant is a twin-unit Westinghouse 3-loop PWR currently operated at rated core thermal power of 2822 MWt (the original licensed core thermal power is 2775 MWt and had been uprated to 2822 MWt through Measurement Uncertainty Recapture Power Uprating, or so called MUR PU) for each unit since 2009. Loss of offsite power will result in loss of all power to the plant auxiliaries, i.e., the RCPs, condensate pumps, circulation water pumps, etc. The main turbine will trip due to loss of condenser vacuum caused by loss of circulation water pumps. The reactor will trip due to turbine trip or loss of power to control rod drive mechanism. Although the three reactor feedwater pumps are turbine-driven it will trip due to feedwater pump low suction pressure resulting from loss of condensate pumps. A loss of condenser vacuum will preclude the use of steam dump to the condenser. The decrease in heat removal by the secondary system is accompanied by a RCS flow coastdown which further reduces the capacity of the primary coolant to remove heat from the core. As the steam system pressure rises following the turbine trip, the steam generator (SG) PORVs may automatically open to relieve steam to the atmosphere. There are two PORVs per SG. If PORV is not available, the steam generator safety valves may lift to dissipate the sensible heat of the fuel and coolant plus the core decay heat. However, the safety vales are not considered in the present analysis. As the no-load temperature is approached, the SG PORVs are used to dissipate the core decay heat. The water level in the three steam generators will decrease continuously due to loss of normal feedwater flow. The TDAFWS will start on low-low level signal in any two steam generators. It will also be stared on low voltage in 4.16 kV class 1E bus. The pump takes suction from the condensate storage tank and is designed to supply rated flow within 1 min of the initiating signal. The water level in each SG will be recovered by the water injection of the TDAFWS. The plant emergency operating procedure (EOP) requires the operator manually depressurize the steam generators and perform the RCS cool-down at maximum rate but not exceeding 100 ◦ F/h. The RCS pressure and temperature will decrease continuously and the accumulators will begin to inject water into the RCS when the RCS pressure decreases to the accumulator initiation setpoint of 600 psia. If no manual depressurization of the secondary side is taken, the RCS is not cooled down and will remain at a relatively high pressure, which will cause the RCP seal leakage rate to be much higher. The RCS pressure will not decrease to the setpoint of accumulators initiation and the accumulator is not available for this case. The core will eventually become uncovered due to continuous RCP seal leakage. The containment is a dry type with design pressure and temperature of 74.7 psia and 300 ◦ F, respectively. There are two containment sumps, separated by 90◦ , inside the containment, each is located 6.67 ft below the plant grade. During normal operation, three of the four fan cooler units operate continuously to remove heat from the containment atmosphere to maintain the average temperature below 120 ◦ F and the relative humidity below 70%. During SBO all fan cooler units will be tripped due to loss of AC power. The containment will be heated up by the reactor coolant sensible heat and the energy of the incoming RCP seal leakage flow. The only heat sink to slow down the heating-up process is through the passive containment heat structures. For the case with manual SG depressurization using PORV the RCS leakage flow will decrease to a relatively low value since the RCS pressure is decreasing continuously. The sensible heat will also decrease to a relatively low value since the RCS average temperature is decreasing and the containment temperature is increasing. For the case without secondary depressurization, the situation may differ much from the case with secondary depressurization. The RCS has a very limited cool-down through the intermittent open and close of SG PORVs. This causes
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the RCS temperature and pressure to remain at high value. The RCP leakage flow rate will be higher than that corresponding to the case with secondary depressurization. Higher RCP leakage flow rate results in an adverse effect on the containment temperature and pressure response and may challenge the design values. 3. RELAP5-3D NSSS model RELAP5-3D is used to simulate the SBO NSSS transient and the nodalization schematics is shown in Fig. 2. The report (Yuann, 2014) can be referenced for the detailed description of the nodalization schemes. In summary, there are three loops and only loop 1 and loop 2 are shown in the diagram. Loop 2 is a loop with pressurizer and loop 3 is identical to loop 1. Major components included in the model are reactor pressure vessel (RPV), core and bypass region, fuel bundles, RCPs, pressurizer, accumulators, SG U tubes and shell sides, main steam lines, pressurizer PORV and safety valves, SG PORVs and safety valves, main steam isolation valves, turbine stop valves, main feedwater, TDAFWS, and RCP seal leakage path. The components are modeled as volume, junction or heat structure with identification number shown in Table 1. The system and components not available during SBO are not modeled. The initial conditions and major assumptions associated with the model is described below. (1) Initial conditions reactor core power: 2830.5 MWt pressurizer pressure: 2250 psia RCS average temperature: 587.8 ◦ F RCS flow: 92,600 gpm/loop SG outlet pressure: 942 psia total steam mass flow rate: 12.58 Mlbm/h.
Table 1 RELAP5-3D model volume and junction identification number. Component name
RELAP5-3D model identification number
RPV upper part, downcomer, and lower part Core
Volumes 170, 172, 174, 190, 192, 32, 40, 42, 62, 90, 92, 30, 60, 2, 12, 22 Volumes 122, 124, 126, 128, 130, 132, 134, 136 Volume 100–116 Volume 320 Loop 1: Volume 290, 292, 294, 298 Loop 2: Volume 390, 392, 394, 398 Loop 3: Volume 490, 492, 494, 498 Loop 1: Volume 280, Loop 2: 380, Loop 3: 480 SG1: Volume 250, SG2: 350, SG3: 450 SG1: Volume 520, SG2: 620, SG3: 720 SG1: Volume 532, SG2: 632, SG3: 732 SG1: Volume 540, 542, 544, SG2: Volume 640, 642, 644, SG3: Volume 740, 742, 744 Volume 760 Junction 981, 982, 983 Junction 984 SG1: Junction 556, 558, SG2: 656, 658, SG3: 756, 758 SG1: Junction 546, 548, 550, 552, 554 SG2: Junction 646, 648, 650, 652, 654 SG3: Junction 746, 748, 750, 752, 754 SG1: Junction 543, SG2: 643, SG3: 743 Junction 775 SG1: Junction 505, SG2: 605, SG3: 705 SG1: Junction 503, SG2: 603, SG3: 703
Core bypass regions Pressurizer RCS cold legs
RCPs SG U tubes SG shell sides SG dome Main steam lines
Main steam header Pressurizer PORVs Pressurizer safety valve SG PORVs SG safety valves
MSIVs Turbine stop valve Main feedwater flow Turbine driven auxiliary feedwater flow RCP seal leakage paths
RCP1: Junction 941, RCP2: 942, RCP3: 943
(2) TDAFWS mass flow rate provided to each SG is assumed as 33.82 lbm/s (Taiwan Power Company, 2014). (3) Shaft seals in each RCP are assumed to fail and the initial RCP seal leakage flow is assumed as 25 gpm based on based on NUMARC-87-00 (NUMARC, 1991). The RCP seal leakage junction flow area is adjusted appropriately to have this flow rate at conditions of t = 0 s. The back pressure (containment pressure) for the RCP seal leakage flow modeling is conservatively assumed as 14.7 psia. (4) Three accumulators start to inject water into RCS when RCS pressure drops below 600 psia. (5) The turbine trip will cause the SG pressure to increase to the PORV opening setpoint and PORV is expected to open and close intermittently. The analysis assumes one PORV of each SG is manually opened to depressurize the SG such that the cooling down rate of the primary system in the first hour is approximately at the maximum value of 100 ◦ F/h allowed by the plant Technical Specification, and then the pressure is controlled at ∼200 psia.
4. GOTHIC containment model GOTHIC (NAI, 2012) is a general purpose thermal-hydraulics computer code which can be used to analyze a simple separate heat transfer problem to a complicated containment configurations involved in traditional or advanced light water nuclear power plants. The GOTHIC containment model used to calculate the pressure and temperature response during SBO of duration 24 h for Maanshan PWR plant is shown in Fig. 3. The model consists of three control volumes, which are volume 1: containment, volume 2: containment sump, and volume 3: environment. There are three flow paths. Flow path 1 simulates the RCP seal leakage resulting from seal damage due to loss of seal cooling during the SBO. Flow path 2 simulates the RCS normal leakage flow, including unidentified and identified leakages. Flow path 3 simulates the flow from containment floor entering the containment sump. As described in Sections 1 and 2 there exists a sensible heat transferring from the RCS to the containment air space. This sensible heat is simulated as a heat source in control volume 1. Also the passive heat structures of the containment are classified into 18 categories, as shown in Table 2. Of these 18 heat sinks, three are modeled as external conductors to transfer heat between the containment and the environment through the containment wall. The three heat sinks are heat sink #1 containment dome and part of wall, #2 part of containment with carbon steel lining of 0.6-in. thickness, and #3 part of containment with carbon lining of 1.6-in. thickness. Four heat sinks are assumed to be covered with the containment sump water on one side and thermally insulated on the outside surface where exposed to earth. The four heat sinks are #4 part of containment floor with stainless steel, #5 part of containment floor with epoxy, #6 part of containment floor with inorganic zinc, and #7 part of containment floor with bare concrete. These four heat sinks are also modeled as internal conductors. The remaining 11 heat sinks are entirely within the containment and are modeled as internal conductors to transfer heat between the heat structures and the containment air space. The initial conditions and major assumptions associated with the model is described below.
(1) Initial conditions a. Volume 1 (containment) pressure = 14.7 psia, temperature = 120 ◦ F, relative humidity = 100% b. Volume 2 (containment sump)
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Fig. 2. Maanshan PWR RELAP-3D SBO model nodalization schematics.
Table 2 Containment passive heat structures and data (Taiwan Power Company, 2014). Heat sink #
Designation
Material and thickness (ft)
1
Containment dome and part of wall
2
Part of containment (carbon steel lining = 0.6 in.)
3
Part of containment (carbon steel lining = 1.5 in.)
4
Part of containment floor with stainless steel
5 6
Part of containment floor with epoxy paint Part of containment floor with inorganic zinc
Inorganic zinc (0.00038), carbon steel (0.02084), air (0.00035), concrete (3.8427) Inorganic zinc (0.00038), carbon steel (0.05), air (0.00035), concrete (4.0) Inorganic zinc (0.00038), carbon steel (0.125), air (0.00035), concrete (4.0) Stainless steel (9.92083), air (0.000348), concrete (8.25) Epoxy paint (0.02417), concrete (1.94042) Epoxy paint (0.00038), inorganic zinc (0.00038), carbon steel (0.02083), air (0.00035), concrete (8.25) Concrete (8.25) Epoxy paint (0.00033), concreter (1.76417) Epoxy paint (0.000242), concreter (1.82358) Stainless steel (0.02083), air (0.00035), concrete (2.11167) Zinc galvanizing (0.00017), carbon steel (0.005), air (0.00035), concrete (2.34) Stainless steel (0.022) Inorganic zinc (0.00375), carbon steel (0.06558) Inorganic zinc (0.00038), carbon steel (0.02225) Inorganic zinc (0.00038), carbon steel (0.17018) Zinc galvanizing (0.00025), carbon steel (0.01042) Epoxy paint (0.001), carbon steel (0.0345) Stainless steel (0.082)
7 8 9 10
Part of containment floor (bare concrete) Concrete interior wall Concrete floor slabs and walls Stainless steel lined concrete
11
Galvanized steel lined concrete
12 13
Miscellaneous stainless steel less than 0.5 in. Miscellaneous carbon steel less than 1 in. with inorganic coating
14
Miscellaneous carbon steel less than 0.5 in. with inorganic coating
15
Miscellaneous carbon steel greater than 1 in. with inorganic coating
16
Miscellaneous carbon steel with galvanizing
17 18
Miscellaneous carbon steel with epoxy coating Miscellaneous stainless steel greater than 0.5 in.
Exposed surface area (ft2 ) 71,946 5978 2891 490 119,559 4312
10,735 59,755 11,400 6935 7350 9755 44,042 108,271 13,913 52,031 5240 5380
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Fig. 3. Maanshan PWR GOTHIC containment model for SBO analysis.
pressure = 14.7 psia, temperature = 120 ◦ F, relative humidity = 100% c. Volume 3 (environment) pressure = 14.7 psia, temperature = 98 ◦ F, relative humidity = 80% (2) Flow boundary conditions a. Flow boundary condition 1 (RCP seal leakage) The leakage flow rate and enthalpy is obtained from the RELPA5-3D calculation. b. Flow boundary condition 2 (normal RCS leakage) The flow rate is assumed to be constant value of 11 gpm per plant technical specification. The enthalpy is assumed to be corresponding to the RCS average enthalpy during SBO and is obtained from RELAP5-3D calculation. (3) Heat source The sensible heat transfer rate from the RCS to the containment is determined by Eq. (1) qfull power ×
Tavg − TCTMT (Tavg − TCTMT )full power
(1)
where Tavg is the RCS average temperature and TCTMT is the containment airspace temperature, qfull power is the fullpower sensible heat transfer rate. At any time step, Tavg is obtained from the RELAP5-3D calculation and TCTMT is from the GOTHIC calculation at the previous time step. (4) The thickness and exposed surface area for each of the 18 passive heat sinks is shown in Table 2 (Taiwan Power Company, 2014). The properties of the different materials
Table 3 Properties of materials associated with the structural passive heat sink (Taiwan Power Company, 2014). Material
Density (lbm/ft3 )
Specific heat (Btu/lbm-◦ F)
Thermal conductivity (Btu/h-ft-◦ F)
Carbon steel Stainless steel Zinc galvanizing Concrete Epoxy paint Inorganic zinc Air
490.0 490.0 445.0 143.0 115.0 170.0 0.06
0.11 0.11 0.09 0.21 0.27 0.12 0.171
25.0 10.0 65.0 40.8 0.5 1.0 0.0174
associated with the heat structures, including density, specific heat, and thermal conductivity, are shown in Table 3 (Taiwan Power Company, 2014). (5) For external conductors #1, 2, and 3, a conservative heat transfer coefficient of 2.0 Btu/h − ft2 − ◦ F, taken form reference (Taiwan Power Company, 2014), is assumed to exist on outside surfaces (exposed to ambient, volume 3), and a combination of condensation heat transfer with any convection and radiation heat transfer is specified for the internal surface (exposed to the containment airspace, volume 1). UCHIDA condensation correlation (Uchida et al., 1965) is used. (6) For internal conductors #4, 5, 6, and 7, a conservative convective heat transfer coefficient of 0.4 Btu/h − ft2 − ◦ F, taken form reference (Taiwan Power Company, 2014), is assumed
Y.-R. Yuann et al. / Nuclear Engineering and Design 288 (2015) 1–18 Table 4 Analysis cases for the NSSS and the containment. Case ID
Description
A1
SBO with secondary depressurization and TDAFWS available for 0 h SBO with secondary depressurization and TDAFWS available for 4 h SBO with secondary depressurization and TDAFWS available for 8 h SBO with secondary depressurization and TDAFWS available for 12 h SBO with secondary depressurization and TDAFWS available for 16 h SBO with no secondary depressurization and TDAFWS available for 24 h
A2 A3 A4 A5 B
to apply between the sump water and the submerged heat sink surface (exposed to containment sump, volume 2), and insulation is assumed on the surface exposed to earth. For internal conductors #8–#18, a combination of condensation heat transfer with any convection and radiation heat transfer is specified for one surface (exposed to containment airspace, volume 1), and insulation is assumed for another surface. 5. NSSS analysis and results Two categories, categories A and B, of the SBO scenario are considered in this study. Category A simulates the transient with secondary depressurization, and different available hours, from 0 to 24 h, for the TDAFWS are assumed. Category B simulates the transient without secondary depressurization, and the TDAFWS is assumed to be available for 24 h. Totally, six cases are analyzed, including five for category A, and one for category B, as shown in Table 4. The purpose of the analysis for the cases of category
7
B is to show the importance of adopting secondary depressurization during SBO, therefore, only one case with TDAFWS available for the whole duration of SBO (24 h) is considered. The category A cases assume that operator uses one PORV of each SG to manually depressurize the SGs and perform cooling down of the RCS at maximum rate of 100 ◦ F/h during the first 1 h after SBO. After the first hour, the SG pressure is controlled at a pressure near 200 psia. The transient response of the key system parameters, including SG pressure, SG water level, RCS pressure, RCS average temperature, RCP seal leakage flow rate and quality, and reactor water level, for the selected Cases A2, A3, and A4 are shown together in Figs. 4–11, respectively. The event sequence for the Case A4 (SBO with secondary depressurization and TDAFWS available for 12 h) is used as an illustration example to explain the transient behavior during the SBO. Fig. 4 shows the #1 SG pressure response. In the first hour, the SG PORV is manually opened to achieve maximum cooldown of the RCS, which causes the SG pressure to decrease sharply. Then the SG pressure is controlled at approximately 200 psia. Since TDAFWS is assumed to trip at t = 12 h (43,200 s), the SG water level begins to decrease after this time and drop to the bottom of the narrow the narrow range level span, as shown in Fig. 5. Due to termination of cold water injection from TDAFWS, the SG pressure stops decreasing and starts to increase. However, the increase of the SG pressure is limited by the continuous water inventory decrease. Between t = 60,000 s and 75,000 s, there exists a pressure oscillation. The pressure response for the other two SGs will behave in the same manner. The SG level will continuously decrease and eventually drop to the bottom of the U-tube region, as can be seen from the wide rang level response in Fig. 5. Starting from t = 0 s, the RCS pressure and temperature will decrease continuously since it is cooled down through the depressurization of the secondary side. At t ∼ 68,100 s, there is a sharp increase of the RCS pressure and temperature, as shown in Figs. 6 and 7, since at this time the SG level has dropped to near the bottom of the SG and the heat removal rate
Fig. 4. #1SG pressure response during SBO for Cases A2, A3, and A4.
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Fig. 5. #1SG water level response during SBO for Cases A2, A3, and A4.
Fig. 6. RCS pressure response during SBO for Cases A2, A3, and A4.
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Fig. 7. RCS average temperature response during SBO for Cases A2, A3, and A4.
Fig. 8. RCP seal leakage liquid mass flow rate per pump during SBO for Cases A2, A3, and A4.
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Fig. 9. Equilibrium quality of the upstream volume of RCP seal leakage junction during SBO for Cases A2, A3, and A4.
Fig. 10. RCP seal leakage vapor mass flow rate per pump during SBO for Cases A2, A3, and A4.
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Fig. 11. RPV water level response during SBO for Cases A2, A3, and A4.
Fig. 12. RCS pressure response during SBO for Case B.
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Fig. 13. RPV water level response during SBO for Case B.
from the primary side to the secondary side is significantly reduced. The RCP liquid leakage mass flow rate in each loop decreases with decrease of the RCS pressure until t ∼ 60,000 s and increases with the increase of the RCS pressure after t ∼ 60,000 s, as shown in Fig. 8. The upstream volume of the RCP seal leakage junction becomes a two-phase mixture at t ∼ 78,920 s, as shown in the equilibrium quality plot (Fig. 9). After this time, due to low density of the twophase mixture flow, the RCP seal leakage liquid mass flow rate drops sharply, as shown in Fig. 8. It is also observed that a vapor flow is appearing after t = 78,920 s, as shown in Fig. 10. To be noted is that there is a momentary leakage mass flow rate increasing at the early transient (the secondary peak amidst the flow decrease) at t = 1676 s. This is explained as follows: the leakage flow path is opened at t = 0 s, and the mass flow rate is determined by critical flow model. The pressure decreasing has a negative contribution to the critical mass flow rate, while the temperature decreasing has a positive contribution to the critical mass flow rate. Fig. 6 shows that the pressure decreases sharply initially, then there is a significant slope change, i.e., the pressure decreasing rate slows down This causes the mass flow decreasing rate to become slower. Meanwhile, Fig. 7 shows that the temperature decreases continuously, which increase the mass flow rate. At the time when the pressure decreasing rate turns to slower, the positive contribution exceeds the negative contribution, which causes the leakage mass flow rate to increase momentarily. Fig. 11 shows the reactor water level response. The reactor water level will be maintained by the accumulator flow before ∼44,040 s. After this time the accumulator becomes unavailable since accumulator pressure has dropped to below the RCS pressure. The reactor water level then starts to decrease due to existence of continuous RCP seal leakage flow. The first time the reactor water level drops to top of active fuel (TAF) is at t ∼ 60,840 s (16.9 h). The time it takes for the reactor water level to drop to below TAF for the six cases are shown in Table 5. It can be extrapolated from this table that, for the case with secondary
Table 5 Calculated time to top of active fuel (TAF) for different cases. Case #
A1 (with secondary depressurization) A2 (with secondary depressurization) A3 (with secondary depressurization) A4 (with secondary depressurization) A5 (with secondary depressurization) B (no secondary depressurization)
Total available time for TDAFWS (h) (referenced to beginning of SBO)
Calculated time to TAF (h) (referenced to beginning of SBO)
0
1.52
4
9.21
8
13.42
12
16.90
16
21.15
24
6.50
depressurization, to prevent the water level from dropping below TAF during the SBO period of 24 h, the TDAFWS must be available for at least 19 h. For the case with no secondary depressurization (Case B), the reactor water level will drop to TAF in 7 h even if the TDAFWS is available for 24 h. This is because the RCS is not cooled down and remains at a relatively high pressure, as shown in Fig. 12. The RCS pressure will not decrease to the setpoint of accumulators initiation (600 psia) and accumulator is not available. The core eventually becomes uncovered due to existence of continuous RCP seal leakage, as shown in Fig. 13. The analyses in this section are based on the assumptions that the shaft seals in each RCP fail initially due to loss of cooling and the seal leakage flow rate is assumed to be 25 gpm initially per RCP. However, the RCP seal may not fail during the SBO. To investigate the effect of the RCP seal leakage, analyses have been performed by deleting the leakage flow path to simulate the case without RCP
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Fig. 14. RCS pressure response during SBO for Cases A4 and A4.1.
Fig. 15. RPV water level response during SBO for Cases A4 and A4.1.
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Fig. 16. RCS pressure response during SBO for Cases B and B.1.
Fig. 17. RPV water level response during SBO for Cases B and B.1.
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seal leakages. Two cases, Cases A4.1 and B.1, of no RCP seal leakages are considered, with the former modified from case A4 and the latter modified from case B, respectively. Transient response for the major parameters, including RCS pressure and reactor water level, are compared, between Cases A4 and A4.1, and between Case B and B.1, as shown in Figs. 14–17. The results show that for the case with secondary depressurization and no RCP seal leakage (Case A4.1), even the RCS pressure is higher than that of Case A4, the RPV water level will not drop to TAF during the SBO of 24 h. For the case with no secondary depressurization and no RCP seal leakage (Case B.1), the RCS pressure will cycle at the pressurizer safety valve opening setpoint (2500 psia), and the RPV level will not drop to TAF. However, for the case B, even the RCS pressure is lower than that of Case B.1, the RPV water level will drop to below TAF due to continuous RCS inventory loss through the RCP seal leakages. 6. Containment analysis and results A containment analysis is performed for the SBO of 24 h based on the GOTHIC model described in Section 4. The containment pressure and temperature response for the six SBO cases discussed in Section 5 are analyzed. And the case A.4 (SBO with secondary depressurization and TDAFWS available for 12 h) is used as an illustration example. The calculated pressure and temperature response for the containment air space are shown in Figs. 18 and 19, respectively. After SBO occurs, heat is added to the containment through the sensible heat transfer, the RCS normal leakage flow and the RCP seal leakage flow. The heat addition is larger than the heat transferring out through various heat structures and the pressure and temperature of the containment air space increase continuously until t = 11,400 s. At this time both the pressure and the temperature reach the first peak values and start to decrease. The drop of the pressure and the temperature is explained below.
15
As shown in Fig. 20, there is a significant phase change to generate more steam around 11,400 s, which leads to more condensation on the colder surfaces. This in turn enhances the heat transfer to the heat structures. It can also be observed in Figs. 21 and 22 that there is a significant increase in heat transfer coefficient and heat transfer rate from the containment to the #1 heat structure surface around 11,400 s (likewise for other heat structures but only #1 heat structure result is shown). This explains why the containment pressure and temperature cease to increase at 11,400 s and have a drop of values at that time. At t ∼ 68,100 s the RCS pressure rises again significantly (as shown in Fig. 6) and causes the RCP seal leakage liquid flow to have a large increase (as shown in Fig. 8). This in turn adds more energy into the containment and causes the containment pressure and temperature to increase sharply until they reach the second peak values, respectively. After this time the containment pressure and temperature start to decrease again since the upstream volumes of the RCP seal leakage junction have been full of low-density vapor and only a very small leakage vapor flow rate exists, which is not large enough to heat up the containment. The containment maximum pressure (the second peak value) calculated is 21.8 psia, which is far below the design pressure of 74.7 psia. The containment maximum temperature (the second peak value) calculated is 329.0 ◦ F, which is higher than the design temperature of 300.0 ◦ F. The maximum containment pressure and temperature calculated for the six cases are summarized in Table 6. It shows that, for the case with secondary depressurization, there exists no clear relationship between the available TDAFWS time and the calculated maximum containment pressure and temperature. This is because the containment pressure and temperature responses are determined by the complicated and combined effects of the RCP seal leakage flow plus energy and the sensible heat transfer due to the higher RCS temperature. The RCP leakage flow rate and fluid energy depend on the reactor coolant state. The reactor coolant
Fig. 18. Containment pressure response during SBO for Case A4.
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Fig. 19. Containment temperature response during SBO for Case A4.
Fig. 20. Phase change rate in containment air space during SBO for Case A4.
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Fig. 21. Heat transfer coefficient on the containment side of heat conductor 1 for Case A4.
Fig. 22. Heat transfer rate on the containment side of heat conductor 1 for Case A4.
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Table 6 Calculated maximum containment pressure and temperature for different cases. Case #
Total available time for TDAFWS (h) (referenced to beginning of SBO)
Calculated maximum containment pressure (psia)
Calculated maximum containment temperature (◦ F)
A1 (with secondary depressurization) A2 (with secondary depressurization) A3 (with secondary depressurization) A4 (with secondary depressurization) A5 (with secondary depressurization) B (no secondary depressurization)
0 4 8 12 16 24
26.3 22.0 21.7 21.8 20.1 22.1
284.1 321.1 327.8 329.0 296.7 336.8
may become two-phase mixture earlier or later depending on the available time for TDAFWS assumed in the NSSS analysis. The maximum containment pressure and temperature for the case with no secondary depressurization, i.e., case B, are 22.1 psia and 336.8 ◦ F, respectively. 7. Conclusion The extended SBO of 24 h has been analyzed with respect to the response of the NSSS and containment for the Maanshan PWR plant. The RELAP5-3D model is used to calculate the NSSS response, and the GOTHIC model is used to calculate the containment response. The shaft seals of the three RCPs are assumed to fail due to loss of seal cooling. The containment model is linked to the RELAP53D model through the RCS sensible heat transfer and the RCP seal leakage flow. Analysis results show that if the RCS is cooled down in time (within an hour) during the SBO through the manual depressurization of the SGs, then there will be no challenge to the reactor core integrity if the TDAFWS is available for at least 19 h. On the contrary, if there is no any SG depressurization action taken during the SBO, then the core will become uncovered and the integrity of the core is challenged, even if the TDAFWS is available for the entire SBO period of 24 h. This is because the RCS is not cooled down and remains at a relatively high pressure and temperature, which causes the RCP seal leakage rate to be much higher. The RCS pressure will not decrease to the setpoint of accumulators initiation and the accumulator is not available. The core eventually becomes uncovered due to existence of continuous RCP seal leakage. In the aspect of containment response, the high-energy RCP seal leakage fluid continues flowing into the containment and heats up the containment. The containment pressure and temperature increase to a high values. There exists no clear relationship between the available TDAFWS time and the calculated maximum
containment pressure and temperature. The containment pressure response is mild with the maximum value of 20–26 psia, which is far below the design value of 74.7 psia and is not a concern. However, the response of the containment temperature is much worse than that of the containment pressure. The most severe containment temperature response occurs for the case with no secondary depressurization and the calculated maximum containment temperature for this case is 336.8 ◦ F, which exceeds the design temperature of 300 ◦ F but is still below the inside-containment safety-related equipment environmental qualification temperature of 450 ◦ F. The study identifies some requisites needed to cope with the extended SBO of 24 h, such as (1) acknowledging the importance of conducting in-time action to depressurize the secondary side, (2) assuring the operability of existing TDAFWS can be maintained for at least 19 h, and (3) enhancing the RCP seal design to minimize the potential seal leakage flow during loss of seal cooling. References Idaho National Engineering and Environmental Laboratory (INEEL), 1998. RELAP53D Code Manual, INEEL-EXT-98-00834. Nuclear Management and Resources Council, Inc. (NUMARC), 1991, August. Guideline and Technical Bases for NUMARC Initiatives Addressing Station Blackout at Light Water Reactors, NUMARC 87-00. Numerical Applications Inc., 2012, January. GOTHIC Thermal Hydraulic Analysis Package Technical Manual, Version 8.0(QA), NAI 8907-02 Rev 20. Taiwan Power Company, 2014. Final Safety Analysis Report for Maanshan Nuclear Power Station Units 1 and 2 (FSAR). Uchida, H., et al., 1965. Evaluation of Post-Incident Cooling Systems of Light Water Power Reactors. In: Third International Conference on the Peaceful Uses of Atomic Energy, New York, A/Conf/28/P/436, l965. 14. Westinghouse Electric Company, 2002, May. WOG 2000 Reactor Coolant Pump Seal Leakage Model for Westinghouse PWRs, WCAP-15603. Yuann, Y.R., 2014. Maanshan PWR RELAP3-D Nodalization and Model Description, and Steady State Initiation, NED-THA-03A16806-REP-009-01. Institute of Nuclear Energy Research, Atomic Energy Council, R.O.C.