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  • International Conference Nuclear Energy for New Europe 2005

    Bled, Slovenia, September 5-8, 2005

    005.1

    RELAP5/MOD3.3 Analysis of the Reactor Coolant Pump Trip Event at NPP Krško for Different Transient Scenarios

    Vesna Benčik, Tomislav Bajs, Nenad Debrecin Faculty of Electrical Engineering and Computing Zagreb

    Unska 3, 10000 Zagreb, Croatia vesna.bencik@fer.hr, tomislav.bajs@fer.hr, nenad.debrecin@fer.hr

    ABSTRACT

    In the paper the results of RELAP5/MOD3.3 analysis of Reactor Coolant Pump (RCP) Trip event at NPP Krško for different transient scenarios are presented. The RCP Trip event occurred at NPP Krško on 25.02.2002 at night when the shift crew began to reduce the power of the plant because of the increase of the temperature reading of the upper radial bearing of the RCP 2. After about 50 minutes when the reactor power was reduced to 28 %, the operators shut down the reactor and the RCP 2 since the temperature reading was still high. Following reactor trip an unexpected steam leak has been identified and the Main Steam Isolation Valves (MSIVs) isolation was performed. According to the plant event report the Steam Generator (SG) Power Operated (PORV) valve did not open at the setpoint. The SG 1 safety valve (SV) 1 opened twice. For the second time the SG 1 SV 1 opened at lower pressure than the nominal setpoint. The motor driven Auxiliary Feedwater (AFW) pumps were stopped due to overheating of the axial bearings and the turbine driven AFW pump was started. After stabilizing the steam line pressure below the SG 1 safety valve opening setpoint and with leaking path on the secondary side isolated, the MSIVs were opened. In the paper different transient scenarios and operator actions were analyzed using RELAP5/MOD3.3 computer code in order to support the resolution of potential safety issues identified as a result of an event and to analyze the severity of consequences in the event of additional failures.

    1 INTRODUCTION

    Best-estimate computer codes are frequently used as a tool for a full understanding of operational events as part of the feedback of operational experience. In support to such purpose, the RCP trip event that occurred at NPP Krško has been previously analyzed using RELAP5/MOD3.3 code and the results were published in [1]. The realistic analysis drew attention to the complexity of the RCP trip event scenario, as well as of interference between inherent plant behavior and operator actions. Additional analyses with different transient scenarios were found necessary to support the resolution of potential safety issues in the event of additional failures. Beside that, a precursor event analysis method has been performed for RCP trip event at NPP Krško, [2]. The precursor event analysis method which was established upon Probability Safety Assessment (PSA) is used for a quantitative assessment of the safety significance of operational events. The method uses the concept of Conditional Core Damage Probability (CCDP) that measures how far, in the PSA model is the event which is being evaluated from the core damage scenario. As a result, for the RCP trip event, the accident

    mailto:tomislav.bajs@fer.hr

  • 005.2

    sequences potentially leading to unacceptable consequences (precursor events) and the resulting CCDPs have been evaluated. The precursor events that may contribute to CCDP according to the PSA evaluation are: 1) Secondary side steam line break and the reduced availability of the main feedwater system due to consequential damage, 2) Steam generator PORVs not opening during the initial pressure peak and 3) Malfunction of AFW motor driven pumps 1 & 2.

    The RELAP5/MOD3.3 has been used to analyze the RCP trip event at NPP Krško for different transient scenarios and operator actions. Transient scenarios that lead to the core damage or other consequences important to safety were identified. A quantitative and qualitative assessment of the consequences of the analyzed transient scenarios has been performed. Following three groups of RCP trip transient cases were analyzed: 1) RCP trip at 28 % power: a) realistic transient analysis and b) AFW not credited 2) RCP trip at 102 % power transient scenarios based on PSA evaluation of CCDP, [2],: a) Case 1: RCP 2 trip, MSIV trip, AFW not credited, b) Case 2: Case 1, MFW trip, c) Case 3: Case 2, SG PORVs not credited, SG SV 1 stuck open in both SGs and d) Case 4: Case 3, SG 1 SV 1 stuck open. 3) RCP trip event cases (Case 5 to Case 27). The results of the CCDP analyses (Case 1 to Case 4) were used to model transient scenarios that may result in the core damage or have other consequences important for safety.

    2 CALCULATIONAL MODEL FOR NPP KRŠKO

    The standard RELAP5/MOD 3.3 nodalization for NPP Krško developed at Faculty of Electrical Engineering and Computing (FER) was used in the analysis, [3] and [4]. The RELAP5 model has 469 volumes and 497 junctions. The total number of heat structures is 378 with total number of mesh points of 2107. The developed RELAP5/MOD 3.3 input data set contains the models of the NPP Krško monitoring as well as protection and control systems. The RELAP5 model contains the detailed models of Safety Injection (SI) system, MFW and AFW system as well as of control systems (Automatic rod control, Pressurizer (PRZR) pressure and level control, Steam dump control and Steam generator level control). The initial conditions for the RCP trip event analysis are summarized in Table 1. The initial conditions for the RELAP5 realistic analysis were obtained as a result of power reduction: 100%-28% calculation.

    Table 1: Initial conditions for Reactor Coolant Pump Trip Event at NPP Krško Parameter Unit Measurement

    (RCP trip at 28 %) RELAP5 analysis

    (28 % power) RELAP5 analysis

    (102 % power)

    Pressurizer pressure MPa 15.5 15.5 15.5 Steam generator pressure MPa 7.27/7.3 7.23/7.23 6.227/6.226 RCS average temperature K 569.79/569.65 569.3/569.3 577.8/577.7 Feedwater temperature K 452.25/451.05 452.25/451.05 492.8/492.79 Feedwater mass flow rate kg/s 160.06/158.81 154.79/154.8 551.93/555.27 Main steam line mass flow kg/s 109.8/125.8 136.48/136.27 551.93/555.26 Pressurizer level % 36.6 38.9 55.27 SG narrow range level % 69.9/69.5 68.45/68.47 69.3/69.3 RCS mass tons - 131.51 131.55 SG mass tons - 62.93/62.94 48.59/48.45 Reactor core power MW (%) 558.32 (28 %) 558.1 (27.99 %) 2033.88 (102 %)

    Proceedings of the International Conference “Nuclear Energy for New Europe 2005”

  • 005.3

    3 RELAP5/MOD3.3 ANALYSIS OF THE REACTOR COOLANT PUMP TRIP EVENT

    3.1 Analysis of the Reactor Coolant Pump Trip Event at 28 % Power

    The results of the analysis of the RCP trip event at 28 % power are presented in Figure 1 to Figure 3. The realistic transient analysis has been performed and the results were published in [1]. The RCP trip event consists of the a) power reduction: 100-28 % transient and b) RCP trip event analysis. The power reduction transient lasted for approximately 50 minutes (3020 sec). It was initiated on the secondary side by a reduction of turbine steam flow. The realistic plant data obtained from Process Instrumentation System (PIS) data, [5], were assumed for the power reduction transient scenario and for main feedwater temperature, respectively. The RCP trip transient was initiated at time=3020 sec when nuclear power was reduced to 28 % by the sequence of the following events: reactor trip, turbine trip and RCP 2 trip. Following the RCP 2 trip, a part of the cold leg flow from the unaffected loop (1) bypassed reactor core and flew to the cold leg of the affected loop (2) so that flow in that loop reversed. Hence, the imbalance in the transferred heat on the secondary side between the two loops was established. Thus, the major part of the heat produced in the core was transferred in the SG 1. Following reactor trip nuclear power was quickly reduced. The average temperature decreased to the value for the Low Tavg & Reactor trip signal, which actuates main feedwater isolation. Owing to steam dump flow, which is determined by the steam dump control – turbine trip mode, the secondary pressure decreased, Figure 1. However, because of MSIV isolation initiated by the operator, the steam dump flow was terminated and the SG pressure increased again. Following the MSIV isolation, the only means of heat removal to the secondary side was by SG PORV and safety valves. The analysis of the measured data has suggested that the SG 1 PORV valve did not open at the setpoint pressure. Also, a nonstandard behavior of SG 1 SV 1 was ascertained. Due to a very late start of the Auxiliary Feedwater (AFW) flow, a low SG 1 level, Figure 2, could be observed in the measurement. Following assumptions were used in the RELAP5 model of the realistic RCP trip event: 1) The SG 1 PORV valve did not open at the setpoint at the beginning of the event. 2) The SG 1 SV 1 valve operating setpoint and valve area were tuned from nominal values to achieve measured behavior. After second SG 1 SV 1 closing signal, the steam line pressure was controlled manually by adjusting the SG 1 PORV valve setpoint. 3) The AFW flow obtained from the measurement was used as the input in the RELAP5 analysis. The role of AFW flow is twofold. First, spray-type injection of cold AFW water efficiently reduces SG pressure due to steam condensation on water droplets. Secondly, the AFW flow recovers SG inventory so that the heat sink is preserved. In the realistic analysis a very good agreement of the calculation with measurement for both phases (power reduction transient and RCP trip event) was obtained, Figure 1, Figure 2.

    In order to determi

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