Station Blackout Analysis for a 3-Loop Westinghouse PWR Reactor Using TRACE

Transcription

1 The Egyptian Arab Journal of Nuclear Sciences and Applications Society of Nuclear Vol 50, 3, ( ) 2017 Sciences and Applications ISSN Web site: esnsa-eg.com (ESNSA) Station Blackout Analysis for a 3-Loop Westinghouse PWR Reactor Using TRACE N. M. El-Sahlamy Nuclear and Radiological Regulatory Authority, NRRA, Cairo, Egypt Received: 2/1/2017 Accepted: 21/3/2017 ABSTRACT One of the main concerns in the area of severe accidents in nuclear reactors is that of station blackout (SBO).The loss of offsite electrical power concurrent with the unavailability of the onsite emergency alternating current (AC) power system can result in loss of decay heat removal capability, leading to a potential core damage which may lead to undesirable consequences to the public and the environment. To cope with an SBO, nuclear reactors are provided with protection systems that automatically shut down the reactor, and with safety systems to remove the core residual heat. This paper provides a best estimate assessment of the SBO scenario in a 3-loop Westinghouse PWR reactor. The evaluation is performed using TRACE, a best estimate computer code for thermal-hydraulic calculations. Two sets of scenarios for SBO analyses are discussed in the current work. The first scenario is the short term SBO where it is assumed that in addition to the loss of AC power, there is no DC power; i.e., no batteries are available. In the second scenario, a long term SBO is considered. For this scenario, DC batteries are available for four hours. The aim of the current SBO analyses for the 3-loop pressurized water reactor presented in this paper is to focus on the effect of the availability of a DC power source to delay the time to core uncovery and heatup. Keywords: SBO/ PWR/ TRACE/ Station Blackout INTRODUCTION After the Fukushima accident, a rising attention is posed on the strategy to cope with a station blackout event. Station blackout is the complete loss of AC electrical power, from both off-site and on-site sources, to the essential and nonessential switchgear buses in a nuclear power plant. Because many safety systems required for reactor core cooling and containment heat removal depend on AC power, the consequences could be sever. In 1975 the Reactor Safety Study )1(, showed that station blackout could be an important contributor to the total risk from nuclear power plant accidents. Operating plant data and several plant specific probabilistic studies yielded the following important findings regarding station blackout )2, 3). 1. The variability of estimated station blackout likelihood is potentially large, ranging from approximately 10-5 to 10-3 per reactor-year. A "typical" estimated frequency is on the order of 10-4 per reactor-year. 2. The capability to restore offsite power in a timely manner (less than 8 hours) can have a significant effect on accident consequences. 229

2 3. The redundancy of onsite AC power systems and the reliability of individual power supplies have a strong influence on the likelihood of station blackout events. 4. The capability of the decay heat removal system to cope with long duration blackouts (greater than 2 hours) can be a dominant factor influencing the likelihood of core damage or core melt for the accident sequence. 5. The estimated frequency of station blackout events that result in core damage or core melt can range from approximately 10-6 to greater than 10-4 per reactor-year. A "typical" core damage frequency estimate is on the order of 10-5 per reactor-year. The station blackout rule 10 CFR )4(, which became effective on July 21, 1988, was promulgated to reduce the risk of severe accidents resulting from station blackout by: (a) maintaining highly reliable AC electric power systems; and (b) as additional defense in depth, assuring that plants can cope with a station blackout for a specified duration selected on a plant-specific basis )5(. The station blackout rule identifies the reliability of onsite emergency AC power sources as being one of the main factors contributing to the risk of core melt resulting from station blackout. Diesel generator units have been widely used as the power source for the onsite electric power systems )3(. Station Blackout sequences are initiated by a loss of offsite power and the associated reactor scram, followed by failure of the station diesels (or gas turbines, if applicable) to start and load. Station blackout sequences are further discriminated into long-term and short-term station blackouts )3( For the current US PWRs, safety injection systems are lost in a station blackout because most systems rely on AC power (accumulators would still inject if the RCS pressure falls low enough). However, core cooling is initially available in a long-term station blackout sequence through the use of turbine-driven auxiliary feedwater and natural circulation in the primary loops. Turbine-driven auxiliary feedwater can operate until the batteries deplete, which normally leads to a loss of control. If AC power is not recovered soon after loss of control, core damage will follow. Some plants might be able to manually control feedwater after battery depletion, but a continuous source of feedwater is still needed to prevent core damage. For a short-term station blackout sequence in a PWR, the turbine-driven auxiliary feedwater system fails at the beginning of the accident. The most frequent cause is failure to start and run for the required time period. The early loss of heat rejection causes the inventory of the reactor coolant systems to boil off, leading to early core damage. Station blackout results in the loss of reactor coolant pump (RCP) seal cooling which leads to a small loss-of-coolant at the RCPs. This introduces the potential for seal failure from high temperatures, particularly for plants using the old seal material in Westinghouse pumps. The associated leakage from the reactor coolant system can accelerate core damage. This concern is most important for long-term sequences because there is an extended period without seal cooling before core damage occurs. For short-term sequences, the time to core damage is much shorter, so seal failures are more likely to occur after core damage )3(. Because of the general importance of station blackout, a more detailed examination of this particular sequence is provided in this study. Two sets of scenarios for SBO analyses are discussed. The first scenario is the short term SBO with no DC power available; i.e., no batteries are applied. In the second one, long term SBO scenario is considered. For this scenario, DC batteries are available for four hours and reactor coolant pumps seal leak is also assumed. SBO Scenarios Description and Assumptions In the current work, two different SBO scenarios are applied to a generic Westinghouse 3-loop PWR. For both of the two scenarios, loss of off-site and on-site AC power coincident with a reactor trip is assumed. The loss of off-site power assumption leads to the tripping of primary reactor coolant 230

3 pumps at reactor trip. Also, failure of the station diesels to start and load is assumed, i.e., no onsite AC power supply is available for either case. In the first case, a short-term station blackout sequence in a 3-loop PWR is discussed. The turbine-driven auxiliary feedwater system is assumed to fail at the beginning of the accident due to the loss of a DC power source. For this case, early core damage is expected as there is no available system for core cooling. In the second case, a long-term station blackout scenario is considered. Turbine-driven auxiliary feedwater is available until the batteries deplete, where a 4-hour battery life is assumed. In this scenario, reactor coolant pump seal leakage of 21 gpm/pump ( m 3 /s/pump) is considered, since no seal cooling is available. A steam generator secondary side leak of 0.5 in 2 /SG (3.226 cm 2 /SG) is also assumed as it is generally thought that it would not be possible to maintain pressure in an isolated, dry steam generator. Table (1) gives the available systems and assumptions used for both SBO scenarios. Table (1): Current Investigation System Availability and Assumptions Parameter Short-Term SBO Long-Term SBO Reactor/turbine trip At t = 0 At t = 0 RCP trip At t = 0 At t = 0 HPSI Not available Not available LPSI Not available Not available Primary system charging/makeup pumps Not available Not available Pressurizer spray Not available Not available Main feedwater system Not available Not available Auxiliary feedwater system Not available Available during battery life (4 hours) SG secondary side steam dump Not available Not available DC batteries Not available pplied during the first 4 hours RCP seal leak Not modeled gpm/pump starting at t = 0 SG secondary side leak Not modeled 5 in 2 /SG starting at t = 0 The plant analyzed in the current study uses a rated power of 2,346 MWth and a thermal design flow rate of m 3 /s. The core is composed of 157 fuel assemblies of the 15x15 lattice design and has an active fuel height of 3.66 m. TRACE V5.0 Patch 4 was used for these analyses. Table (2) shows the initial conditions for the SBO TRACE analysis. Table (2): Current Investigation Initial Conditions Parameter 231 Initial Value Core Power (MWth) 102% of 2,346 Fuel Type 15x15 Thermal Design Flow Rate (m 3 /s) Initial RCS Pressure (MPa) 15.8 Vessel Inlet Temperature (K) Vessel Outlet Temperature (K) SG Secondary Pressure (MPa) 5.82 SG Feed Water Temperature (K) 490 Steam Flow per SG (Kg/s) 421 Steam Generator Liquid Level (m) 0.86 Total Pressurizer height (m) Pressurizer Liquid Level (m) 4.92

4 Description of 3-Loop Westinghouse TRACE Model To perform the analyses, a generic TRACE input model for a 3-loop PWR reactor has been used. The model simulates a 3-loop PWR of Westinghouse design, as shown in Figure (1). It is a generic model that is not intended to be specific to any particular plant. The model consists of 116 hydraulic components, 348 control blocks, 36 heat structures and a power component. The main structure of this model includes the reactor pressure vessel, pressurizer, hot and cold legs, reactor coolant pumps, steam generators, steam piping at the secondary side, steam dump system, accumulators, and high and low pressure safety injection. The pressure vessel is divided into 16 levels in the axial direction, and two rings in the radial direction, where the inner ring represents the core region (from axial levels 3 to 10) and the outer ring represents the down comer region, and three equal azimuthal sectors in the θ direction. The fuel region nodalization uses eight axial layers, between the third and tenth axial levels of the vessel. Heat structures were added to simulate the reactor core and other vessel internals. Fig. (1): TRACE Model for a 3-Loop Westinghouse PWR The details of the TRACE model used are described in the next section. 1. Pressurizer and Related Control System The pressurizer plays an important role in the PWR operation, since it maintains the pressure of the RCS. The important pressurizer internals included in the model are the sprays, the power operated relief valves and the safety relief valves. The pressurizer heater is not modeled, however, this is acceptable since it will not be available during SBO simulations. The pressurizer is modelled with one PIPE component, PIPE 324, with a total of 13axial cells. A detailed modelling of the pressurizer component and its related verification is given in a previous study )6(. 2. Steam Generator and Feedwater System The steam generator of the 3-loop Westinghouse power plant modeled is a vertical U-tube heat exchanger. The U-Tube in the primary side is modeled by a PIPE component with 18 cells. The 232

5 secondary side is modeled with three components; the boiler, downcomer and the steam dome and separator component. The region of the boiler is separated into seven cells, the region of the downcomer is separated into 12cells, and the region of the steam dome and the separator are separated into 12cells. A feedwater control system is applied to control the feedwater flow. The main function of the feedwater control system is to maintain a fixed water level in the steam generator secondary side. 3. Emergency Core Cooling System (ECCS) The ECCS is designed to automatically deliver cooling water to the reactor core in the event of a LOCA. This limits the fuel cladding temperature increase and ensures that the core remains intact and amenable to cooling. The ECCS includes the accumulators, the low pressure safety injection, and the high pressure safety injection. Figure (2) gives a schematic diagram for a PWR emergency core cooling system. The accumulators were simulated using a PIPE component divided to five volumes. The total volume for each accumulator is 34 m 3 containing 24.8 m 3 of liquid. The accumulator flow is controlled through a check valve. When the pressure of the RCS is reduced to a value lower than the accumulator pressure, the check valve opens and the accumulator injection begins. The LPSI and HPSI are modeled by FILL components in each loop. Each FILL component uses a table of RCS pressure versus flow rate. Fig. (2): A Schematic Diagram for a PWR Emergency Core Cooling System Nodalization and Steady State Qualification The nodalization used in the current investigation is shown in Figure (1). In the first step of this study, steady state calculations are carried out. After that transient analyses are performed where loss of off-site and on-site AC power coincident with a reactor trip is assumed. The loss of off-site power assumption leads to the tripping of primary reactor coolant pumps at reactor trip. Also, failure of the station diesels to start and load is assumed, i.e., no onsite AC power supply is available, as stated before. For nodalization qualification, steady state runs are performed for comparison with the nominal values of the reference plant parameters. The reference plant considered for comparison is H.B. 233

6 Robinson nuclear power plant. Data for this plant is taken a previous study )7(. The parameters compared are core thermal power, inlet vessel temperature, primary pressure, steam generator parameters and reactor coolant pump flow rate. Steady state results obtained are compared with the reference plant parameters as outlined in Table (3). Table (3): Steady state conditions Quantity Reference value Calculated value Error (%) Key reactor parameters Number of loops 3 3 Core nominal power at full power (MWth) 2,300 2,346 2 Start up power for transient - 2,392 (102% of 2,346) Nominal inlet temperature (K) Primary pressure nominal (bar) Active core height (m) Steam generator parameters Steam generator inlet temperature (primary) (K) Steam generator outlet temperature (primary) (K) Steam generator flow rate (primary), (kg/s) Reactor coolant pump and coolant piping parameters Reactor coolant pump flow (L/s), each Flow per loop (kg/s) Velocity hot leg (m/s) Velocity cold leg (m/s) Reactor outlet (hot leg), ID, (m) Reactor inlet (cold leg), ID, (m) As shown in Table (3), good agreements are found between the steady state results and the related values of the reference plant parameters. Therefore, the current nodalization can be used with confidence to perform the analyses in the study. SBO RESULTS AND DISCUSSION In this section, the investigation results for both long-term and short-term SBO scenarios are given. The SBO is assumed to occur at t = 0.0 s, reactor trip is initiated and from this moment forward, any active system of the plant becomes unavailable. For the short-term SBO, the auxiliary feedwater pumps are unavailable. The scenario of the accidents for both of the long-term and the short-term SBO are discussed in the next sections. Long-Term SBO Scenario At t =0.0 s, the SBO occurs and the reactor trip is initiated. The core thermal power falls rapidly to the decay heat level with the effect of shutdown rods, as shown in Figure (3).Primary system pressure falls from its nominal value until reaching a minimum of7.78 MPa at t = 11,082 s, as a consequence of the scram and it stabilizes at about 8.13 MPa at t = 12,534 s, Figure (4). As the system pressure decreases, coolant saturation temperature also decreases, as shown in Figure (5). Coolant temperature slightly increases, with some fluctuations, until t = 12,534 s ( Figure 5). 234

7 On the secondary side, as the SBO starts, the turbine stop valves (TSV) close. Steam driven AFW pump is available for the first four hours and supplies feedwater to the steam generator, as shown in Figure (6). As a consequence, steam generator level fluctuates near its nominal value as observed in Figure (7). Also, as the TSVs close the SG pressure starts to increase from 6.5 MPa until reaching the steam generator safety relief valve (SRV) set point( Figure 8). During this period, core decay heat is transferred from the primary side to the secondary side through the SG tubes. This heat transfer is reduced and eventually stops after the SG secondary boils dry after the AFW pump stops supplying feedwater to the steam generator when the batteries are depleted at four hours. At this point, coolant temperature reaches saturation temperature given the reduced heat transfer to the SGs and boiling begins in the core. Also, primary pressure starts to fluctuate near a fixed value of 8.13 MPa, as shown in Figure (4). Until this point, the steam generator pressure is still near the relief valve set point pressure, as shown in Figure (8). At t = 22,866 s, steam generator level reaches zero, and the steam generator is totally void of liquid( Figure 7). At this point, primary system coolant temperature increases (Figure 5). As boiling present in the core, the primary pressure increases until reaching the pressurizer SRV set point. The pressurizer SRVs cycle, keeps the primary pressure between 16.2 and 15.7 MPa until the end of the simulation, as shown in Figure (4). Also, the pressurizer level reaches a maximum during this time as liquid is drawn up into the pressurizer as the relief valves open, as shown in Figure (9). Due to core boiling and loss of mass through the pressurizer SRVs and RCP seal leaks, the pressurizer empties by t = 27,160 s and remains empty until the transient is over. At t = 27,663 s, the core has boiled dry and has a collapsed liquid level of zero, as shown in Figures (10 and 11). The downcomer level reaches zero at t = 29,036 s, as seen in Figure (10). As the core liquid level reaches zero and the core is totally voided, core heatup occurs, and temperature excursions are noticed as seen on Figure (12). The fuel rod temperature increases until it reaches the fuel rod clad melting point at t = 32,715 s at which point the TRACE simulation is ends. It should be noted that even though an RCP seal leak is assumed, the system depressurization results from this leakage is negligible. The primary system pressure remains high from the start to the end of the transient, as seen in Figure (4). The resulting high primary pressure prevents the only safety injection available, the accumulators, from injecting into the cold leg and supplying cold water to the core. Fig. (3): Total Reactor Power (W) Fig. (4): Long-Term SBO-Primary System Pressure (Pa) 235

8 Fig. (5): Long-Term SBO-Primary Coolant and Saturation Temperatures (K) Fig. (6): Long-Term SBO-AFW Pump Mass Flow (kg/s) Fig. (7): Long-Term SBO-SG Level (m) Fig. (8): Long-Term SBO-SG Pressure (Pa) Fig. (9): Long-Term SBO-Pressurizer Level (m) Fig. (10): Long-Term SBO-Downcomer and Core Level (m) 236

9 Fig. (11): Long-Term SBO-Core Liquid Volume Short-Term SBO Fraction Scenario The major difference between short-term and long-term SBO scenarios is the absence of DC power at the beginning of the short-term SBO transient. As a consequence, the turbine-driven auxiliary feedwater system fails to start, and the early loss of heat rejection causes the inventory of the reactor coolant to boil off sooner, leading to earlier core damage. The two scenarios are nearly similar, but the expected events during the transient happen earlier for the short-term SBO. A comparison between events sequence for both of the two accident scenarios is given in Table (4), and Figures (13) through (17) present comparisons of key parameters for the two station blackout transients. Table (4): Events Sequence for SBO Event Time (seconds) Short-Term SBO Long-Term SBO Transient begins Steam generators empty 3,294 22,866 Pressurizer empty 6,843 27,160 Core totally voided 7,314 27,663 Downcomer totally voided 8,325 29,036 Core melt starts 10,974 32,715 Fig. (12): Long-Term SBO-Maximum Peak Clad Surface Temperature (K) Core damage is expected to occur about three hours after the start of the transient in the shortterm SBO. In the long-term SBO with four hours of battery available, core damage is expected after nine hours. It should be noted that while the auxiliary feedwater runs for four hours in the long-term SBO, the core melt begins about 6 hours later than in the short-term SBO. This is because core decay heat is lower after the four hour battery life, so it takes additional time for the core to boil off and dryout. During this period, if AC power is recovered and both safety injection, HPSI and auxiliary feedwater is made available, then core damage can be mitigated. 237

10 Fig. (13): Primary System Pressure (Pa) Fig. (14): Pressurizer Level (m) Fig. (15): SG Level (m) Fig. (16): Core Level (m) Fig. (17): Maximum Peak Clad Surface Temperature (K) 238

11 CONCLUSIONS In the current study, the TRACE best estimate code was used to perform Station blackout accident analysis for a 3-loop Westinghouse reactor. Two different SBO scenarios are considered. In the first scenario, a short-term station blackout in a 3-loop PWR is discussed. The turbine-driven auxiliary feedwater system is assumed to fail at the beginning of the accident due to the loss of a DC power source. In the second scenario, a long-term station blackout is considered. Turbine-driven auxiliary feedwater is available until the batteries deplete, where a 4-hour battery life is assumed. In this scenario, a reactor coolant pump seal leakage of 21 gpm/pump ( m3/s/pump) is considered, since no seal cooling is available. A steam generator secondary side leak of 0.5 in2/sg (3.226 cm 2 /SG) is also assumed. For both of the two scenarios, loss of off-site and on-site AC power coincident with a reactor trip is assumed. The loss of off-site power assumption leads to the tripping of primary reactor coolant pumps at reactor trip. Also, failure of the station diesels to start and load is assumed, i.e., no onsite AC power supply is available for either case. From the analyses, it is found that core damage is expected to occur about three hours after the start of the transient in the short-term SBO. In the long-term SBO with four hours of battery available, core damage is expected after nine hours. So, in the long-term SBO, the core melt begins about 6 hours later than in the short-term SBO. This is because core decay heat is lower after the four hour battery life so it takes additional time for the core to boil off and dryout. During this period, if AC power is recovered and both safety injections, HPSI and auxiliary feedwater is made available, then core damage can be mitigated. ACKNOWLEDGEMENTS The author gratefully acknowledges the International Atomic Energy Agency (IAEA), Argonne National Laboratories, and the United States - NRC since this research study is performed during an IAEA assignment at the United States - NRC. REFERENCES (1) "Reactor Safety Study: An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants," WASH-1400, October (2) U. S. Nuclear Regulatory Commission, "Evaluation of Station Blackout Accidents at Nuclear Power Plants, Technical Findings Related to Unresolved Safety Issue A-44", NUREG-1032, June 1988, p (3) NUREG/CR-6042, Rev. 2 SAND , Perspectives on Reactor Safety,ERI Consulting Sandia National Laboratories, Oak Ridge National Laboratory, U.S. Nuclear Regulatory Commission, Office of Human Resources, Washington, DC (4) 10 CFR 50.63, "Loss of All Alternating Current Power," July 21, (5) 53 FR 23203, "Statement of Considerations for Final Station Blackout Rule," June 21, (6) Cheng, Y.H., Shih, C., Wang, J.R., Lin, H.T., Benchmark Calculations of Pressurizer Model for Maanshan Nuclear Power Plant using TRACE Code. In: Proceedings of the 16 th International Conference on Nuclear Engineering (ICONE16), Orlando, Florida, USA, ICONE , May 11 15, (7) NUREG/CR-4183, Pressurized Thermal Shock Evaluation of the H. B. Robinson Unit 2 Nuclear Power Plant, Oak Ridge National Laboratory, Volume 1, ORNL/TM-9567/V1. 239