Transient Analysis of LBLOCA with Subsequent Loss of Offsite Power and Loss of All AC Power in VVER-1200 Reactor

The generation III+ VVER-1200 (AES-2006) reactor is equipped with improved safety features with component redundancy compared to the previous models. This paper is concerned with the performance of active and passive safety features during the transient of a major beyond design basis accident (BDBA) scenario of the reactor. The study is done considering large break loss of coolant accident (LB LOCA) with loss of offsite power (LOOP) and LB LOCA with loss of all AC power (LOACP) sources namely station blackout (SBO) using VVER-1200 NPP simulator. It was found that the reactor is capable of mitigating such BDBA scenarios by the effectiveness of the active and passive safety systems. Moreover, LB LOCA with LOACP is observed to be more severe compared to LBLOCA with LOOP. This comparative analysis is compatible with IAEA safety guidelines.


Introduction
A safe, dependable, and effective nuclear reactor has long been a human desire.As a result, diverse reactor types are developed depending on various specifications.A system that starts, controls, and maintains a fission chain reaction is known as a nuclear reactor.As a result, nuclear reactors can be categorized according to various reference parameters.The Pressurized Water Reactor (PWR) and Boiling Water Reactor are the two most well-known reactor types (BWR) [1].The International Atomic Energy Agency (IAEA) reports that 451 nuclear reactors around the world are now in operation, including 298 PWR reactors and 73 BWR reactors [1].Any nuclear reactor is built to the high standards of safety outlined in the IAEA safety guidelines.The world has seen some of the worst nuclear disasters despite the deployment of safety measures.The Three Mile Island Accident (1979), the Chernobyl Disaster (1986), and the Fukushima Daiichi Nuclear Disaster stand out among them as the most notable ones (2011).These mishaps can occur for a variety of causes, including human mistakes, poor design, and natural calamities.Nevertheless, the results were dangerous.This led to the death of many people living near the site by getting the radiation dose [2].Three groups of transients, including anticipated operating occurrences (AOOs), design basis accidents 1305 (2024) 012012 IOP Publishing doi:10.1088/1757-899X/1305/1/012012 2 (DBAs), and BDBAs, are taken into consideration and examined to defend the nuclear power plant (NPP) safety from the standpoint of reactor engineering and radiation repercussions [3].It has been seen that most of the major accidents were caused by Loss of Coolant Accident (LOCA), loss of offsite power (LOOP), and loss of AC power (LOACP).Among them, LOCA is studied extensively [8], [10].In the Chinese CPR100 reactor, [6] evaluated vessel rupture under various station blackout (SBO) scenarios together with the failure of the steam generator (SG) safety relief valve.Additionally, [7] indicates how the break size spectrum of LOCAs affects the time of vessel rupture as well as the quantity of generated hydrogen, corium, and fission products released for VVER-1000/V320.The lower head integrity of the RPV for the VVER-1200 (V-491) reactor is assessed in a study conducted by [18] using MELCOR 1.8.6 considering LB LOCAs and SBO occur at the same time.Moreover, [9] analyzed the thermal hydraulic parameters of the VVER-1200 reactor during LBLOCA concurrent with LOOP using PCTRAN.LOCA is the loss of coolant from the reactor due to a rupture in the main coolant pipeline.There are two kinds of LOCA, depending on the size of the leak or rupture [4].A large break LOCA (LB LOCA) is a double-ended rupture in the primary coolant pipeline with a leak greater than 0.1m 2 [10], [11].The scenario of loss of offsite power represents the failure of active part components which need electricity from the grid and failure of the active part as well as backup diesel generators in the case of loss of AC power.Most of the severe accidents occur due to the failure of the cooling system.It is very difficult to predict a severe accident scenario.Before an accident, there are some initiating events like loss of coolant accident (LOCA), small break loss of coolant accident (SBLOCA) [13][14], total loss of feed water (TLOFW), station blackout (SBO), and steam generator tube rupture (SGTR).The comparative analysis of LBLOCA with LOOP and LBLOCA with LOACP using the VVER-1200 NPP simulator is an essential task as these events may arise at any moment.Proper knowledge and management may help to overcome such kinds of disastrous situations.The present work represents the investigation of the safety system failure of the VVER-1200 reactor in the worst accident case and its consequences.This simulation research will be helpful to understand the new coming VVER-1200 Power reactor operating manner and also help to develop skilled manpower.

Technical specification of the simulator
The simulator is supplied by Western Service Co. (WSC), US with 3KEYMASTER™ modeling tools which include 3KEYMASTER™ instructor Station.The simulator includes a full range of plant operations, including cold shutdown, hot standby, hot zero power, and a full range of power maneuvers, in addition to transient.The simulator is versatile enough to simulate the VVER-1200/V392M reactor technology.It functions as a real-time, comprehensive, and high-fidelity simulator.Students can use the simulator to conduct full startup and shutdown operations, load maneuvers, and simulate both typical and atypical plant transients, such as failure situations.The simulator is a Pressurized Light Water Reactor consisting of four circulation loops, four steam generators, and four reactor coolant transfer pumps.The primary circuit consists of a reactor pressure vessel (RPV) which holds the fuel and control rods, pressurizer (PRZ), steam generator (SG), and main coolant pumps (MCP) as shown in Fig. 1. (a).The turbine, which has a high-pressure turbine and four low-pressure turbine cylinders, receives steam from the steam generators.The condensate pumps draw condensate from three condensers and pump it through low-pressure heaters, a deaerator, and the suction of five electric feedwater pumps.Through high-pressure heaters, the feedwater pumps deliver the required feed flow to the steam generators.Steam is sent to the low-pressure turbines by four moister separator reheaters, which remove moisture from the high-pressure turbine exhaust steam.All of these arrangements of the secondary circuit are shown in Fig. 1. (b).Table I contains the primary equipment and specifications that are the basis for the simulator development.

Reactor core modeling
The following design criteria have been considered for modelling the reactor core: During routine operations and anticipated operational occurrences, the minimal deviation from the nucleate boiling ratio is not less than 1.19.The design overpowers the condition's maximum fuel centerline temperature is below that limit, which could trigger centerline fuel melting.During routine operations and anticipated operational events, the melting point of UO2 is not exceeded.Fuel rod cladding is made to keep its integrity over the course of the fuel's life.Each reactor system is built to effectively dampen any xenon transients.The reactor coolant system is built and planned to remain functional over the anticipated plant life.Power excursions that might be the result of a plausible reactivity addition incident do not damage the pressure vessel by deforming it or rupturing it, nor do they interfere with the operation of the engineered safety systems.The simulator has a straightforward, user-friendly interface design.The simulator can be used in accelerated time mode or real-time mode.Utilizing a real-time simulator has the benefit of allowing users to comprehend system responses immediately and without the constraints of pre-recorded scenarios.

Process of simulation
The initial conditions and process of simulation are described below: 1) To simulate the LB LOCA scenario concurrent with LOOP and LOACP, the following initial condition at the nominal power level has been considered.2) In the simulation, 3 different initial conditions (IC) have been established based on core fuel lifetimes such as the Beginning of core life (BOL), Middle of core life (MOL), End of core life (EOL), and core power level: 3) The following boundary conditions and shutdown setpoints have been implemented for simulating the studied BDBA scenario.The initial conditions and reactor trip setpoints have been chosen keeping the nominal operating condition of the reactor in mind which is prescribed by the status report of VVER-1200 [8].
• Short neutron flux period T <10 second • High neutron flux >107% Nnom • Low pressurizer level LPRZ<4.6m4) The initial condition was set to BOL which indicates 100% core power, core fuel loading is the beginning of core life for the equilibrium cycle and equilibrium Xe and Sm.The reactor was operated in rated power condition for about 780 seconds in the case of LBLOCA with LOOP and for about 838 seconds in the case of LBLOCA with LOACP.Full rupture Dnom = 850mm (100%) of Cold Leg has been applied between the Reactor and RCP.All 10KV normal power supply systems from the power grid have been disconnected for LB LOCA with LOOP scenario and the 10KV emergency power supply has been disconnected for LB LOCA with LOACP scenario.5) The sequence of input commands for the simulator has been described below.At the time of Simulation initiation, LBLOCA with LOACP had to be initiated manually.As a result, the scenario actuation time was not the same i.e., 780 seconds for LBLOCA with LOOP and 838 seconds for LBLOCA with LOACP.So, the time should be considered concerning the accident initiation time.For both cases, the reactor trip occurred followed by the turbine trip within 2 seconds.All the control rods were fully inserted to safely shut down the reactor and to make the reactor subcritical.

Result and discussion
The VVER-1200 simulator has been used to do the numerical analysis for LBLOCA with loss of offsite power (LOOP) and loss of AC power (LOACP).The loss of coolant caused the reactor to trip.As a result, in the case of LBLOCA with LOOP, the power dropped quickly, reaching 1.7% from 100% within 3 seconds, and from 100% to 2.7% in the case of LBLOCA with LOACP.The power did not reach 0 immediately because of the decay heat from the fission fragment.Both scenarios are shown in Fig. 2. The pressure in the pressurizer (PRZ) dropped quickly from 15.85 MPa to around 1 MPa within 18 seconds for LBLOCA with LOOP and within 17 seconds for LBLOCA with LOACP due to coolant loss from the primary circuit and malfunctioning proportional heaters.In both situations, the water level in the pressurizer dropped sharply and stabilized at roughly 3.2 m after 100 seconds.These phenomena are depicted correspondingly in Fig. 3. Following the immediate cessation of steam flow, steam generator pressures should rise due to the trip of the Reactor Coolant Pump (RCP) and feed water pump (FWP).Up until the steam generator atmospheric dump valve's function, the pressure is anticipated to rise.Steam generator pressure falls as steam is released from them, and it will do so even after the reactor trips.When the turbine tripped, the steam generator level first dropped as a result of shrinkage.In the event of LBLOCA with LOOP, the auxiliary feed water was fed to the SG and the water level increased while the backup diesel generators operated.Diesel generators were turned off for LBLOCA with LOACP, though.The temperature of the hot leg and cool leg is reduced as a result of the reactor trip.The temperature for LBLOCA with LOOP indicated in Fig. 5 reached around 104°C in the hot leg and approximately 69 °C in the cold leg after 900 seconds of the scenario activation.At roughly the same period, the temperature for LBLOCA with LOACP was indicated to reach around 135°C in the hot leg and approximately 125°C in the cold leg.Due to the backup diesel generator usage of the emergency cooldown system to provide coolant to the core in the LOACP scenario, the temperature in the hot and cold leg is considerably greater in the case of LBLOCA with LOACP.Hydro accumulators are the passive component of ECCS which provides borated water to the core having a concentration of 16g/L and starts operating after the pressure drops below a prescribed level in the boundary conditions.The Water flow from the Hydro accumulator through the core is shown in Fig. 6.During LOCA, the coolant is lost from the core resulting pressure decrease in the RPV.The HA-1 started operating after the pressure dropped below 5.9MPa and had a maximum flow of about 1835 kg/s in the case of LBLOCA with LOOP and 1830 kg/s in the case of LOCA with LOACP.This supply of coolant kept the temperature and pressure in the RPV down.The HA-1 started within 7 seconds and operated for about 194 seconds and 168 seconds respectively for both cases.The actuation of HA-2 occurred when the pressure dropped below 1.5MPa and had a maximum flow of 32kg/s and 40kg/s respectively for LBLOCA with LOOP and LB LOCA with LOACP.The HA-2 would continue supplying coolant to the core for about 24 hours respectively after the accident scenario.
In normal conditions, the containment was kept at a negative pressure typically of -0.24kPa to balance the pressure during any leakage.The leaked coolant was contained in the containment which caused a rise in containment pressure.The pressure increased and reached a steady value of 40kPa for both cases which is below the design pressure [8].
As hot water entered the containment building, the temperature increased to a maximum of 150ºC (initially 36ºC), then decreased due to the actuation of backup power (diesel generators), the hydroaccumulator (HA), Emergency core cooling system.On the other hand, the temperature decreased slowly after reaching 150ºC due to the absence of diesel generators and ECCS active part, only HA-1, HA-2, and Passive Heat Removal System were operational in LOCA with LOACP scenario.The simulated HA-1 and HA-2 data were in good agreement with MELCORE 1.8.6 code-based data [8].Similarly, Fig. 7 represents the containment temperature after LOCA with LOOP and LOACP conditions.The Overall experimental simulation process is given in chronological sequence below Table III and Table IV respectively.The sequence of events for the case of LBLOCA with LOACP/SBO starting from the initiation of the event to the end of simulation has been compared with MELCORE code generated results as shown in TABLE V.The reactor trip action, actuation of HA-1, actuation of PHRS and operation time of HA-1 has been observed to be having very small deviation from the MELCORE code generated results.Simulated MELCOR analysis, VVER-1200/V491 [18] Break and SBO occur.

Conclusion
The total power dropped quickly, after reaching 1.7% and 2.7% within 3 seconds in the case of LBLOCA with LOOP and LBLOCA with LOACP respectively.The total power decreased more abruptly for LB LOCA with LOOP compared to LB LOCA with LOACP.Hot leg and cold leg temperatures were observed to decrease safely (far below design temperature 350℃), which suggests maximum cladding temperature did not rise above the safety margin (should be <1200 ℃) [8].The maximum recorded containment temperature and pressure was 150℃ and 40kPa which is below the design tolerable range (Design temperature 210℃, pressure 500kPa) [8].Comparatively, the hydro accumulator first stage (HA-1) actuated simultaneously in both situations, but it ran for a longer period in the case of LB LOCA with LOOP (196sec as compared to 171sec for the other case) since the ECCS active component and SG emergency cooldown system were actuated.On the other hand, the hydro accumulator second stage (HA-2) began to run sooner for LB LOCA with LOACP and the hydro accumulators handled all of the core cooling.It was discovered that the hydro accumulator findings in the scenario of LB LOCA with LOACP were in good agreement with those provided by the MELCORE 1.8.6 code [18].
In accordance with International Atomic Energy Agency (IAEA) safety reports, neither the pressurizer pressure nor the steam generator pressure was above the limits of safety limits.When compared to LBLOCA with LOOP, it has been found that LBLOCA with LOACP is more severe.Based on the results, it is evident that the VVER-1200 reactor will be safe during the major design extension scenario of LBLOCA with LOOP or LBLOCA with LOACP.

Table 3 .
Chronological sequence of events for the case of LBLOCA with LOOP.

Table 4 .
The chronological sequence of events for the case of LBLOCA with LOACP

Table 5 .
Comparison between simulated and MELCORE code-based sequenceIn the case of LBLOCA with LOACP