RELAP5/MOD3.2 INVESTIGATION OF A VVER-440 STEAM GENERATOR HEADER COVER LIFTING

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1 Science and Technology Journal of BgNS, Vol. 8, 1, September 2003, ISSN RELAP5/MOD3.2 INVESTIGATION OF A VVER-440 STEAM GENERATOR HEADER COVER LIFTING Pavlin P. Groudev, Rositsa V. Gencheva, Antoaneta E. Stefanova Institute for Nuclear Research and Nuclear Energy Bulgarian Academy of Sciences Tzarigradsko Shaussee 72, Sofia 1784, Bulgaria pavlinpg@inrne.bas.bg roseh@inrne.bas.bg antoanet@inrne.bas.bg ABSTRACT This report discusses the results of the thermal-hydraulic analysis of an accident at the Rivne NPP. The accident was caused by primary-to-secondary reactor coolant leakage as a result of full and partial steam generator header covers lifting. The initiating event is full hot collector cover lifting in one of the sixth steam generators (SG#5) with equivalent diameter 107 mm. Hot collector cover lifting in other three SGs #1, #3 and #4 follows this event. Such accident provides a direct release path for contaminated primary coolant to the environment via the secondary side. RELAP5/MOD3.2 computer code [1], [2] has been used to simulate the transient at the Rivne VVER440 Nuclear Power Plant (NPP). A model of the Rivne Unit 1 has been developed based on RELAP5/MOD3.2 thermal-hydraulic code at the National Taras Shevchenko University of Kiev [9] and has been given to the INRNE staff for performing of PRISE analyses. This investigation is a process that compares the analytical results obtained by the RELAP5 computer model of VVER440 mentioned above against the experimental transient data received from the Rivne VVER440 Nuclear Power Plant, Unit 1 [5], [6]. The results of this investigation provide an integrated evaluation of the complete RELAP5 VVER440/V213 model. As it seen from the results RELAP5 predicts correctly the behavior of main plant parameters. Key words: Rivne NPP, VVER440, PRISE accident, transient conditions, plant data 1 INTRODUCTION Since experimental facilities are usually scaled down models of real plants, there is an additional need to evaluate accident analysis code performance in actual plant conditions. This task was enveloped in the International Nuclear Safety Program defining a benchmark problem for validation of thermal-hydraulic codes for application to Soviet-designed VVER440 reactors based on actual plant data. Therefore, this plantbased standard problem is a valuable addition to the validation database. The reference power plant for this analysis is

2 Unit 1 at the Rivne NPP site. Operational data from Rivne NPP [6] are available for the purpose of assessing how the RELAP5 model compares against the plant data. The following sections of this report include brief description of Rivne VVER440 power plant, description of the test being studied, description of the RELAP5/MOD3.2 input model, results, and conclusions. 2 RIVNE NPP UNIT 1 GENERAL DESCRIPTION Rivne NPP Unit 1 with VVER440/V213 pressurized water reactor was commissioned on December 22, Unit 1 of the Rivne NPP has six coolant loops, isolation valves and horizontal steam generator on each loop. The unit includes two 220 MW steam turbines. Unit 1 has the standard emergency core cooling system (ECCS) including high-pressure pumps, low-pressure pumps and hydro-accumulators. Primary pressure maintenance system consists of pressurizer, two pilot operating relief valves (PORVs), bubbler tank, piping, control and cut-off valves. Emergency core flooding system consists of four hydro-accumulators, piping and control and cutoff valves. Reactor pressure vessel (RPV) has six inlet and six outlet nozzles (500 mm diameter). The outlet nozzles are located at a higher elevation than the inlet nozzles. Reactor core: composed of 276 fuel assemblies, 37 movable control assemblies and 36 shield assemblies located in the periphery of the core. The fuel assemblies contain 126 fuel rods each, arranged in a triangular grid with a rod pitch of 12.2 mm. Reactor Coolant system (RCS) transports heat from the reactor core to the steam generators that provides steam to the turbine generators through the main steam lines. Six primary coolant loops have common flow paths through the reactor vessel. Each RCS loop includes a horizontal steam generator, a main circulation pump and Main Gate Valves (MGV). The primary coolant flows from the reactor outlet nozzle to the steam generators and then is pumped by the reactor coolant pumps through the MGV placed on the cold leg to the reactor inlet nozzle. Pressurizer, which maintains overall system pressure (12.5 MPa) and compensates the changes in the primary coolant volume, is connected to the RCS between the reactor outlet nozzle and the hot leg isolation valve in one of the loops. The pressurizer is connected with the cold leg of the primary loop piping at the reactor coolant pump discharge by an injection pipeline. The steam generators are horizontal units, with submerged tube bundles and a built-in steam separator. Each unit includes a cylindrical horizontal shell, two vertical inlets and two horizontal tube bundles of stainless steel, U shaped tubes of 16 mm diameter and 1.4 mm thickness. The primary coolant flows through the tube side, the feed water is delivered to the shell side. 3 RELAP5/MOD3.2 MODEL This section describes the modeling assumptions and nodalization for the development of a RELAP5/MOD3.2 model for VVER440, Unit 1 Rivne NPP. The Baseline input deck for VVER440/V213 Rivne Nuclear Power Plant Unit 1 [4], [9] is provided by National Taras Shevchenko University of Kiev. A 6-loop primary system model of the Rivne Unit 1 has been developed for the systems thermal hydraulic code RELAP5/MOD3.2. The RELAP5 model of Rivne NPP was developed for analyses of operational occurrences, abnormal events, and design basis scenarios. The model provides a significant analytical capability for the specialists working in the field of NPP safety. Data and information for the modeling of these systems and components were obtained from the Rivne documentation and

3 from the power plant staff. The input model has been transformed into a RELAP5 input deck, which is composed of the following groups of data: job control cards, trip and control system data, hydrodynamic component data, heat structure data, reactor kinetics data, and general tables. In the baseline input model there have not been made changes in the nodalization scheme. The KU model has been stabilized on 82% power by INRNE team. Only the differences in the safety systems behavior are implemented in the Baseline input deck for VVER440/V213. For INRNE calculation performing it has been created a small restart input, which uses the results of the steady state. In this small restart input there have been modeled the behaviors of Main Gate Valves (GZZs), BZOKs, SI system. The breaks also have been modeled there. The approach of modeling GZZs was done according to the actual work of the valves - discussed with NPP operators. In the input deck all four SGs hot collector cover lifting have been modeled as a trip valves. Flow energy loss coefficients were accepted by engineering judgment to be 1.0 for both: forward and reverse flow. Based on plant data, the flow discharge coefficients at the breaks are assumed to be 1.0 for both SGs (#5 and # 1), with full hot collector cover lifting. For SGs #3 and #4 it is determined partially hot collector cover lifting in the plant data. By engineering judgment it has been estimated the flow discharge coefficient for SG #3 to be 0.25 and respectively 0.17 for SG#4. Figure 1. Nodalization scheme of the reactor VVER-440/213 Figure 2. Nodalization scheme of steam generator Figure 3. Nodalization scheme of the break 4 DESCRIPTION OF THE TRANSIENT At the beginning of the accident the thermal power was MW (82 % nominal

4 power). The emergency core cooling system with high-pressure injection pumps and emergency core cooling pumps were in the standby mode according to the plant design. The initiating event is steam generator header cover lifting. It results in abrupt primary pressure and pressurizer water level decreasing, initiated by the leak of the primary coolant to the cavity of SG#5 vessel through opening of the hot header cap. It is determined, that the size of the hot header cap opening was equivalent to a 107 mm diameter pipe. After 12 seconds the reactor SCRAM is actuated due to primary pressure decreasing to 95 kgf/cm 2. During the first minute the rate of primary pressure drop is 1 kgf/cm 2 per second. According to the designed mode 30 sec after the signal for accident initiation all three safety systems are activated diesel generators with the automatic loading system, the emergency core cooling system with borated water reserve equal 900 m 3 with temperature of +25 C, +30 C, and also hydro-accumulators passive injection, when the primary pressure decreases to 60 kgf/cm 2. The operator closes Turbine Stop Valves of the Turbine Generator (TG) -2. Pressure in the main steam pipe lines decreases from 47 kgf/cm 2 to 42 kgf/cm 2 and the operator manually shut down the TSV of the Turbine Generator (TG) -1. When the primary pressure reaches the set point 60 kgf/cm 2 the hydro-accumulators 1, 2, 3 start to inject in the downcomer and upper reactor part. When the primary pressure is stabilized at value 40 kgf/cm 2 water supply from hydro-accumulators was stopped. At 300 sec during the transient time following the start-up of diesel-generators the Makeup pump-3 (with capacity of 50 m 3 /h ) starts to deliver into the sixth loop. At 780 sec during the transient time the operator stops the Reactor Coolant Pump (RCP) 5 and closes main gate valves (GZZs), but the coolant leakage to SG#5 vessel continues. After the activating of the emergency core cooling system with high pressure injection pumps the primary pressure was stabilized at the level of approximately 40 kgf/cm 2 i.e. equal the secondary pressure. At 900 sec during the transient time the hydro-accumulator 4 starts to inject borated water with temperature 40 C into the primary system. At 1800 sec during the transient time an uncontrolled increasing of SG#3 water level starts. As a result of this RCP#3 is stopped and the main gate valves (GZZs) on the loop #3 are closed. Due to leaks localization and isolation primary loops #3 and #5 primary pressure starts to increase and at 2340 sec reaches 105 kgf/cm 2. After that moment it starts to decrease abruptly to 40 kgf/cm 2 because of failure of hot primary header s cap on SG#1 and leak of the primary coolant continues. The operator stops RCP#1 and closes the main gate valves on the same loop#1, but the primary pressure and the pressure inside SG#1, #3, #5 vessels continues to change until the equilibrium pressure conditions at the both sides of the rupture are established. This confirms the fact that GZZs are not completely tight. At 2700 sec safety isolated valves of SG#1 open. At 3000 sec because of discrepancy of the feedwater flow, necessary to support the nominal level in SG#4, the RCP#4 is stopped and main gate valves on the loop#4 are closed. During this period the highpressure injection pumps are in operation and through unsealed main gate valves(gzzs) the leak of the coolant continues. During the period from 3540 sec to 4200 sec safety relief valves of steam generators #1, #3 and #5 were periodically opened. At 3900 sec for certain reasons RCP#2 and #6 stop and in period of 1620 sec the forced coolant circulation stops. The fuel cooling was performed by the natural circulation of the coolant under the injection of the boric water by the highpressure injection pumps. At 5580 sec operator starts up RCP#6 and at 7560 sec starts up RCP#2. 5 INITIAL CONDITIONS The comparison between the initial

5 conditions of the plant data parameters before initiation of the accident and the RELAP5 calculation at 82% reactor power (steady state conditions at 82% reactor power) is shown in Table 1. Table 1. Steady state conditions at 82% reactor power Parameter Dimensi on RNPP value RELAP5 value Thermal reactor power MW Pressure at the reactor MPa outlet Pressurizer level M Reactor coolant temperature: Cold leg: Hot leg: Pressure drop across the MCPs Loop 1 Loop 2 Loop 3 Loop 4 Loop 5 Loop 6 K kpa no data no data No data MSH pressure MPa Primary coolant K average temperature Pressure drop across kpa 287, the reactor Water temperature in ECCS storage tanks K Parameter Auxiliary feedwater temperature Emergency feedwater temperature Dimensi on RNPP value RELAP5 value K K RESULTS AND DISCUSSION The transient test scenario is modeled using the RELAP5/MOD3.2 computer code and the VVER440/V213 input model for Rivne NPP, Unit 1 [9]. As the results show, RELAP5 predicts the plant behavior correctly. The chronological sequence of transient events is shown in Table 2. Table 2. Chronological sequence of transient events Description of event Measured value, s 1. SG-5 hot collector cover lifting (107 mm equivalent diameter) RELAP5 Calculated results, s Reactor SCRAM ECCS actuation. Primary Letdown system closure 4. Turbine-2 stop valve closure 5. Turbine-1 stop valve closure 6. HA-1(2,3) start injection 7. HA-1(2,3) stop injection 8. Start of CVCS pump injection into loop 6 (G=50 m 3 /h) 9. MCP-5 switching off. Closure of GZZs on the loop HA-4 start injection

6 SG-3 hot collector 11. cover lifting 12. MCP-3 switching off. Closure of GZZs of loop 3 SG-1 hot collector 13. cover lifting 14. MCP-1 switching off. Closure of GZZs of loop 1 SG-1 SVs opening 15. SG-4 hot collector 16. cover lifting 17. MCP-4 switching off. Closure of GZZs of loop 4 SG-3(4,5) SVs 18. opening MCP-2(6) switching off. HPIS switching off End of calculation No data No data No data Figure 5. Secondary side pressure The transient calculations are compared with the plant data in Figure 1 through Figure 9. The calculation was performed up to sec of the transient time. Before running of the investigated transient event the RELAP5 model was run in real plant equilibrium conditions to establish steady state conditions at 82% reactor power. Figure 6. Hot and cold leg temperatures in the Loop #1 Figure 4. Primary side pressure Figure 7. Hot and cold leg temperatures in the Loop #4

7 Figure 8. Head of MCP#3 Figure 9. Head of MCP#6 Figure 10. Reactor pressure drop Figure 11. Average primary side temperature The most important parameters behavior is shown in Figures from 4 to 11 for the both cases: RELAP5 calculation and plant measurements. The calculation was performed up to approximately 90 min (5500 sec) into the transient time. The initiating event of this analysis is hot collector cover lifting in SG#5. Since the primary system pressure is initially much greater than the steam generator pressure, reactor coolant flows from the primary into the secondary side of the affected steam generator. In response of this loss of reactor coolant, pressurizer level and RCS pressure decrease. Fast depressurization of the primary circuit and rapid increasing of the water level in the affected SG#5 characterize the initial phase of transient. Figure 4. present measured primary pressure during the plant transient event and calculated one. The figure indicates good agreement between the plant data and calculated pressure behavior. In accordance to the designed mode hydroaccumulators start to inject borated water with temperature 40 0 C when the primary pressure decreases to 60 kgf/cm 2. This event comes at 60.0 sec. Three hydro-accumulators #1, #2 and #3 start to inject borated water to downcomer and upper reactor volume. After approximately sec. primary side

8 pressure in the both cases (the measured and the calculated one) is stabilized at level of 4.0 MPa (see Figure 4.). The high pressure injection pumps were activated to provide borated water to cold legs of the second, third, and sixth loops with total flow rate of approximately 240 m 3 /h at 4.0 MPa. Due to fast primary pressure decrease, after the signal P 1 <115 kgf/cm 2 there is actuation of the Reactor Scram-2 (AZ-2) and consequent reactor power decreasing this event comes at 2.9 sec. Coolant discharge leads to the further rapid decreasing of primary pressure (see Figure 4.) and 0.51 sec. after the signal P 1 <95 kgf/cm 2 Reactor Scram occurs. In approximately 12 sec all control rods fall to the core bottom. This event occurs at 12.0 sec for the both cases (the plant measured and the RELAP5 calculated one). In accordance with the instructions NPP staff stops Turbine Generator #2 by closing turbine stop valves. In the model Turbine Generator have been isolated at 30 sec. At the 50 sec during the transient time pressure in the main steam pipe lines (at RNPP) decreases from 47 kgf/cm 2 to 42 kgf/cm 2 and NPP staff manually shuts down Turbine Generator #1 by closing the turbine stop valves. Secondary side pressure behavior in SG#1 for the both cases (the measured and the calculated one) is presented in Figures 5. The deviations after the first 3200 sec. in RELAP5 calculations come due to work of safety valves of the corresponding SG. Generally comparison of the measured and the calculated secondary pressure in the SG#1 shows good agreement. In Figure 10. it has been presented Reactor Pressure Drop. As it seen from the figure the behavior of the both curves: measured and the calculated one is similar. The trends of the both curves are almost the same. Another important characteristic is the coolant temperature in the cold and hot legs. Comparison between the temperatures in the loops #1 and #4 is presented in Figures 6. and 7. As it seen from the Figures there is very good agreement between the plant measurements the calculation results. The comparison between the measured and the calculated pump heads for MCPs #3 and #6 is presented in Figures 8. and 9. As it seen from the figures the trends of the measured pump head curves closely follow the calculated. In the Figure 11 there are presented the calculated and the measured primary side average temperature. In the calculation it has been used the common approach for determination of average primary temperature. It has been calculated dividing the sum of all hot and cold legs temperatures by 12. As it seen from the Figure 11 the calculated primary average temperature follows the measured one very close. According to the definition report the measured average temperature has been recorded approximately to 3800 sec. 7 CONCLUSIONS The RELAP5 model developed for the transient analysis of the performance of VVER- 440/V213 nuclear power plants has been used to predict the results obtained during the transient Primary-to-Secondary reactor coolant leakage as result of SG header cover lifting with 107 mm equivalent diameter break at the Rivne NPP, Unit 1. The results provided in this report show that RELAP5 predicts correctly the behavior of main plant parameters. These results provide an integrated evaluation of the complete RELAP5 VVER440/V213 model. 8 REFERENSES [1] The RELAP5 Development Team. August RELAP5/MOD3.2 Code Manual, NUREG/CR-5535, INEL-95/0174, Vol.1, 1995 [2] Guideline For Performing Code Validation Within The DOE International Nuclear Safety Center (INSC), US/Russian International Nuclear Safety Center, Argonne, Illinois and Moscow.

9 September 23, [3] Rivne NPP Unit 1 short description and transient Data Base. May [4] Data base for VVER-440/213 reactor safety analyses. Trnava. SR. [5] RNPP Unit 1 In-depth Safety Assessment Project. NSSS Data Base for Design Basis Accident Analysis. 1D210DL11R09.DOC [6] The Data on Rivne NPP Unit 1 accident 01/23/1982 Leakage from primary to secondary side through steam generators caused by hot cover lifting [7] Leakage of primary circuit coolant into the secondary circuit within the steam generator through the sealing of the primary circuit hot header in a NPP with a WWER-440 Reactor. Rovno-1. Incident 909. Report to IAEA [8] American National Standard for Decay Heat Power in Light Water Reactors. ANSI/ANS American Nuclear Society Standards Committee. Working Group ANS [9] Description of Rivne NPP Unit 1 RELAP Model, Kyiv September, BOA A-R4, Task#5, Deliverable 2.1a. 9 FIGURES LEGEND Figure 1. Nodalization scheme of the reactor VVER-440/ Figure 2. Nodalization scheme of steam generator... 3 Figure 3. Nodalization scheme of the break... 3 Figure 4. Primary side pressure... 6 Figure 5. Secondary side pressure... 6 Figure 6. Hot and cold leg temperatures in the Loop # Figure 7. Hot and cold leg temperatures in the Loop # Figure 8. Head of MCP# Figure 9. Head of MCP# Figure 10. Reactor pressure drop... 7 Figure 11. Average primary side temperature 7 10 TABLES LEGEND Table 1. Steady state conditions at 82% reactor power... 5 Table 2. Chronological sequence of transient events... 5 [10] Draft Rivne NPP VVER-440 Thermal Hydraulics Standard Problem Analysis Report.

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