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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 [email protected], [email protected], [email protected] ABSTRACT In the paper the results of RELAP5/MOD3.3 analysis of R eactor C oolant P ump (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 M ain S team I solation V alves (MSIVs) isolation was performed. According to the plant event report the S team G enerator (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 A uxiliary F eedwater (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 P robability S afety A ssessment (PSA) is used for a quantitative assessment of the safety significance of operational events. The method uses the concept of C onditional C ore D amage P robability (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

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Page 1: RELAP5/MOD3.3 Analysis of the Reactor Coolant Pump Trip ... · RELAP5/MOD3.3 Analysis of the Reactor Coolant Pump Trip Event ... of the plant because of the increase of the temperature

International ConferenceNuclear 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 [email protected], [email protected], [email protected]

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

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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”

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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 determine the consequences of inability to establish the AFW flow throughout the transient, the case: RCP trip at 28 % power based on realistic event and with AFW flow not credited was analyzed, Figure 3. During the first hour the heat generated in the reactor core is removed by heat transfer in the SG 1 and afterwards in the SG 2. After both SGs had been depleted, the heat removal takes place by relieving the RCS inventory through the pressurizer PORV. The safety injection signal was not actuated. At time=9700 sec when the RCS inventory decreased to app. 50 tons, the core dry-out occurred.

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

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0 500 1000 1500 2000 2500 3000 3500 4000 4500Time [s]

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sure

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6.4

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7

7.2

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8

RCP Trip Event: Realistic Analysis

PLOT FER V2W 16:47:01, 21/07/05

calculation measurement

Figure 1: RCP trip realistic analysis: SG 1 secondary pressure, power reduction (0-3020

sec)

Mas

s flo

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1500 2000 2500 3000 3500 4000 4500T I ME (sec)

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5060

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SG 1 level-measurement SG 1 level-calculation AFW 1 flow

RCP T rip Event: Real ist ic A nalysis

R5PLOT FER V1.3 13:26:10, 25/08/2005

Figure 2: RCP trip realistic analysis: SG 1 level and AFW 1 mass flow rate (0-3020 sec)

Tem

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0 1000 2000 3000 4000 5000 6000 7000 8000 9000 10000 11000t ime (sec)

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RCS mass SG 1 mass SG 2 mass PRZR PORVs mass Cladding temperature

RCP T rip at 28 %, AFW not credi ted

R5PLOT FER V1.3 13:33:10, 25/07/2005

Figure 3: RCP trip at 28 % power at t=0, AFW not credited: Core cladding temperature

(N=11), RCS mass, RPV mass, SG mass, Integrated PRZR PORV mass

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

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3.2 Analysis of the Reactor Coolant Pump Trip Event at 102 % Power for Different Transient Scenarios

The initial and boundary conditions resulting in the most conservative core cooling conditions were assumed in the analysis (102 % reactor power, RCP trip and MSIV trip at the beginning of the transient). The maximum value for the initial reactor power (nominal value plus 2 % uncertainty) is conservative because of maximum decay heat and the least amount of the SG mass, Table 1. As previously discussed, the heat produced in the core is removed through the PRZR PORV first after both SGs are depleted. The selection of the transient scenarios and operator actions that were analyzed was performed in two steps. In the first step the transient scenarios based on events identified in [2] to contribute to CCDP were analyzed. The results of the analyses performed in the first step were used to model transient scenarios (Step 2) that may result in the core damage earlier in the transient or have other consequences important for safety than those identified in [2].

Analysis of the RCP trip event for transient scenarios based on PSA evaluation of CCDP

The transient scenarios that contribute to CCDP based on PSA evaluation of the RCP trip event have been analyzed. The following is the list of precursor events identified in [2] to contribute to CCDP for the RCP trip event at NPP Krško: 1. Secondary side steam line break and a reduced availability of the main feedwater system due to consequential damage., 2. Steam generator PORVs not opening during the initial pressure peak. The most noticeable contribution to CCDP is coming from stuck open safety valves not re-closing or stuck open., and 3. Malfunction of AFWS motor driven pumps 1 & 2. The following four cases were analyzed: 1) Case 1: AFW not credited (base case), 2) Case 2: Case 1, MFW trip at the beginning of the transient, 3) Case 3: Case 2, SG PORVs not credited, SG SV 1 stuck open in both SGs and 4) Case 4: Case 3, SG 1 SV 1 stuck open. The results are shown in Figure 4 to Figure 9 and the main events are summarized in Table 2. The transient is initiated by RCP 2 trip and MSIV isolation. In the cases 2, 3 and 4 an immediate isolation of main feedwater was assumed. In the Case 1 the SG mass increased because of continuing addition of the main feedwater whereas the steam mass flow is isolated, Figure 4. The major part of the heat produced in the core is transferred in the SG 1 since the RCP pump in the loop 2 is stopped. The SG 1 PORV and SG 1 safety valve open in the cases 1 and 2 while in the cases 3 and 4 only SG 1 safety valve opens. After SG 1 had been depleted the heat removal takes place in the SG 2 where either PORV opens (Case 1 and 2) or SV 1 (Case 3 and 4). For the cases 3 and 4, a Low-2 steam line pressure signal has been generated due to stuck open safety valve that actuates the SI signal. The amount of the injected mass from SI depends on RCS pressure, Figure 5, Figure 6. Owing to a very intensive heat removal a rapid decrease of coolant temperature and the increase of coolant density occur in the cases 3 and 4. As a result, an outsurge from the pressurizer with a rapid RCS pressure drop occurs along with steam flow through the SG safety valve 1. In the Case 3, SG 2 SV 1 was also assumed stuck open, thus causing a second RCS pressure drop and considerably more injected SI mass into the RCS than in the Case 4. The RCP 1 was tripped in the Case 3 and Case 4 due to signal: subcooling less than required coincident with SI signal. After both SGs had been depleted the heat produced in the reactor core was no longer removed by heat transfer to the secondary side. RCS temperature rise together with coolant expansion resulted until pressurizer PORVs opened, Figure 6. The inventory relieved through the pressurizer PORVs was pure liquid for all analyzed cases, Figure 7. As a result of the loss of RCS inventory, the reactor pressure vessel (RPV) mass decreased, Figure 8. Finally, fuel cladding temperature increase and core dry-out occurred, Figure 9. The analyzed cases diverge into two groups. Fuel cladding temperature, Figure 9, started to rise when RPV mass dropped to app. 22.5 tons for the cases

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

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005.6

with SI and both RC pumps stopped (Case 3 and 4), Figure 8, and to app. 10 tons in the cases where RC pump 1 was still running (Case 1 and 2), respectively. The temperature rise for the Case 1 and 2 was much slower than in the Case 3 and 4. In the cases 1 and 2 the calculation was stopped because of the temperature increase of SG structure material above 700 K. The shortest time to core dry-out (6565 sec) was obtained for the Case 4 (SG PORVs not credited, SG 1 SV 1 valve stuck open). In the Case 3 much more mass from SI was injected into the RCS than in the Case 4. Consequently, more RPV mass was available for cooling thus resulting in the core dry-out later in the transient than in the Case 4, although the SGs were depleted earlier. A very small difference for the time of the start of the core cladding temperature rise between the cases 4 and 2 can be attributed to the SI injection in the Case 4, as well. The presented results suggest that for this transient, a fast depletion of SGs due to stuck open safety valves, does not unambiguously lead to more adverse cooling conditions in the core. The difference for the time of the start of the core cladding temperature rise between the cases 1 and 2 can be attributed to the difference in the time of MFW trip only. Additional safety concern in all analyzed cases is related to liquid discharge through pressurizer PORVs. First, water discharge through pressurizer PORVs or safety valves may lead to overpressurization of the pressurizer relief tank (PRT), i.e., rupture of the PRT rupture disks. Secondly, the pressurizer PORVs and safety valves are not qualified for water relief. Additional concern is that pressurizer safety valves may fail to properly reseat when exposed to water discharge causing the event to progress to small break LOCA accident.

Table 2: Time table of main events for the CCDP analyses

Event Case 1 Time(sec)

Case 2 Time (sec)

Case 3 Time (sec)

Case 4 Time (sec)

RCP 2 trip, MSIV isolation 0 0 0 0 Reactor trip, turbine trip 2.5 2.5 2.5 2.5 Main feedwater isolation 57.6 0 0 0 Safety injection - - 380.9 380.9 RCP 1 trip - - 5184.5 4652.6 Core dry-out 9000 6635 6965 6565 Pressurizer liquid solid 6285 4044 3165 3780

0 1000 2000 3000 4000 5000 6000 7000 8000 9000 10000 11000 12000T IME (sec)

Mas

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kg)

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20000

30000

40000

50000

60000

Case 1: SG 1 Case 1: SG 2 Case 2: SG 1 Case 2: SG 2 Case 3: SG 1 Case 3: SG 2 Case 4: SG 1 Case 4: SG 2

RCP Trip Event: CCD P analyses

R5PLOT FER V1.3 16:20:10, 31/08/2005

Figure 4: RCP trip CCDP analysis - SG mass

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

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0 1000 2000 3000 4000 5000 6000 7000 8000 9000 10000 11000 12000T IME (sec)

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Pa)

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13

13.5

14

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Case 1 Case 2 Case 3 Case 4

RCP Trip Event: CCD P analyses

R5PLOT FER V1.3 16:32:13, 31/08/2005

Figure 5: RCP trip CCDP analysis - Pressurizer pressure

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Case 1: RCS Case 1: PRZR PORVs Case 2: RCS Case 2: PRZR PORVs Case 3: RCS Case 3: PRZR PORVs Case 4: RCS Case 4: PRZR PORVs

RCP T rip Event: CCD P analyses

R5PLOT FER V1.3 16:43:43, 31/08/2005

Figure 6: RCP trip CCDP analysis - RCS mass and integrated PRZR PORVs mass flow

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RCP Trip Event: CCD P analyses

R5PLOT FER V1.3 16:23:18, 31/08/2005

Figure 7: RCP trip CCDP analysis - Liquid fraction at pressurizer top

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RCP Trip Event: CCD P analyses

R5PLOT FER V1.3 16:35:20, 31/08/2005

Figure 8: RCP trip CCDP analysis - Reactor pressure vessel mass

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RCP Trip Event: CCD P analyses

R5PLOT FER V1.3 16:37:51, 31/08/2005

Figure 9: RCP trip CCDP analysis - Fuel cladding temperature (N=11)

Analysis of the RCP trip event for selected transient scenarios based on evaluation of

the results of the CCDP analyses Two safety concerns have been identified in the CCDP analyses to be relevant in the

RCP trip event. First, the loss of heat sink caused by loss of feedwater flow and the resulting depletion of the SGs and of the RCS lead to inadequate core cooling and core dry-out. Secondly, the pressurizer fills with liquid and the liquid discharge through the pressurizer PORVs occurs. The analysis of the cases 3 and 4 has pointed to the interaction of the heat removal rate from the secondary side by means of SG PORV and safety valves and the safety injection and its influence on the time of the core dry-out. It has been demonstrated that consequences of different scenarios cannot be estimated without analyzing each case. For this purpose, the transient scenarios for different combinations of status of the SG PORV and SV 1 have been analyzed. The SG SV 1 valve not credited (NA) case was not considered.

Transient results for the critical parameters (time to core dry-out, time for pressurizer liquid solid and liquid discharge) are summarized in Table 3. The initial conditions for the analysis were based on Case 2 (RCP trip from 102 % power, MFW trip, MSIV isolation at the beginning of transient). The cases with SG 2 SV 1 stuck open and SG 2 PORV operating were not analyzed since the SG 2 SV 1 does not open in that case, i.e., Cases 5 and 17 are identical.

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

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The shortest time to start of the core dry-out (6290 sec) was obtained for the Case 6 (SG 1 PORV stuck open). The fastest depletion of the steam generator was obtained for the case 20 (SG 1 PORV stuck open, SG 2 PORV not credited, SG 1 SV 1 and SG 2 SV 1 stuck open). As previously discussed, an intensive heat removal from the primary side leads to a considerable decrease in primary pressure and the large amount of the injected SI mass. Therefore, for the Case 20, the largest amount of the injected SI mass as well as the longest time to core dry-out (7000 sec) among the analyzed cases was obtained. The hypothetical case with SI disabled and with fastest loss of SGs inventory resulting in the most adverse cooling conditions was modeled (Case 21). The core dry-out in this case (6180 sec) occurs only 110 sec earlier than in the Case 6. Another consequence of the largest amount of the injected SI mass in the Case 20 was the early liquid solid condition in the pressurizer (at time=3015 sec).

Table 3: Time of events: RCP trip event (RCP 2 trip, MSIV isolation, MFW trip) – SG PORV and SV 1 behavior

Case

Core dry-out

Time (sec)

PRZR liquid solid (discharge) Time (sec)

Case 2 base case (RCP 2 trip, MSIV, MFW trip, AFW NA) 6635 4020 (4000) Case 3 SG 1 PORV, SG 2 PORV NA, SG 1 SV 1, SG 2 SV 1

stuck open

6965

3165 (3145) Case 4 SG 1 PORV, SG 2 PORV NA, SG 1 SV 1 stuck open 6565 3780 (3760) Case 5 SG 1 PORV NA 6390 3865 (3840) Case 6 SG 1 PORV stuck open 6290 3910 (3880) Case 7 SG 2 PORV NA 6365 3965 (3940) Case 8 SG 2 PORV stuck open 6330 3740 (3720) Case 9 SG 1 SV 1 stuck open 6540 3755 (3740) Case 10 SG 1 PORV NA, SG 1 SV 1 stuck open 6495 3750 (3730) Case 11 SG 1 PORV NA, SG 2 PORV stuck open 6340 3880 (3840) Case 12 SG 1 PORV NA, SG 2 PORV, SG 1 SV 1 stuck open 6700 3585 (3565) Case 13 SG 1 PORV, SG 2 PORV NA, SG 2 SV 1 stuck open 6700 3760 (3740) Case 14 SG 1 PORV, SG 2 PORV NA 6300 3830 (3815) Case 15 SG 1 PORV, SG 1 SV 1 stuck open 6630 3720 (3700) Case 16 SG 1 PORV, SG 2 PORV stuck open 6550 3985 (3920) Case 17 SG 1 PORV NA, SG 2 SV 1 stuck open 6390 3865 (3840) Case 18 SG 1 PORV, SG 2 PORV, SG 1 SV 1, SG 2 SV 1 stuck

open

6770

3470 (3445) Case 19 SG 1 PORV stuck open, SG 2 PORV NA 6300 3920 (3890) Case 20 SG 1 PORV stuck open, SG 2 PORV NA, SG 1 SV 1, SG

2 SV 1 stuck open

7000

3015 (3000) Case 21 Case 20, no SI (hypothetical case) 6180 4000(3985) Case 22 SG 1 PORV stuck open, SG 2 PORV NA, SG 1 SV 1

stuck open

6620

3740 (3725) Case 23 SG 1 PORV stuck open, SG 2 PORV NA, SG 2 SV 1

stuck open

6720

3645 (3625) Case 24 SG 1 SV 1 stuck open, SG 2 PORV NA 6535 3780 (3750) Case 25 SG 1 SV 1 stuck open, SG 2 PORV stuck open 6705 3585 (3550) Case 26 SG 1 SV 1 stuck open, SG 2 PORV NA, SG 2 SV 1 stuck

open

6950

3135 (3115) Case 27 SG 2 PORV NA, SG 2 SV 1 stuck open 6780 3790 (3760)

4 CONCLUSION

Different transient scenarios for the RCP trip event at NPP Krško using RELAP5/MOD 3.3 code have been analzyed. The analyzed transient scenarios are divided into three groups: 1) RCP trip at 28 %: a) realistic transient analysis and b) AFW flow not credited, 2) RCP trip at 102 % power for transient scenarios based on PSA evaluation to contribute to CCDP, and

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

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3) RCP trip at 102 % for transient scenarios not identified in the PSA based evaluation of CCDP but that lead to core damage and other unacceptable consequences. The following conclusions can be drawn from the analysis of the RCP trip event for analyzed scenarios: • At the beginning of the transient the heat produced in the core is removed by opening of the SG PORV and SV 1 valve. After both SGs had been depleted the RCS temperature rise together with coolant expansion resulted until pressurizer PORVs opened. Unless the controlled operator action is taken, the heat removal using PRZR PORVs leads to loss of coolant inventory, inadequate core cooling and core dry-out. The unavailability of the AFW flow causing the loss of heat sink is of prime importance in the transient. The RCP trip event at 28 % with AFW flow not credited led to the core dry-out 9700 sec after transient begin. • Four transient cases selected to contribute to CCDP based on PSA evaluation have been analyzed. There are two distinct groups among the analyzed cases. In the cases where the RCP 1 was not stopped (Cases 1 and 2) core dry-out occurred at much lower RPV mass (10 tons) and at slower rate than for the Cases 3 and 4 (at RPV mass equal to 22.5 tons). In the Case 3 (both SG 1 and SG 2 SV 1 valve stuck open) more SI mass and a later core dry-out (6965 sec) was obtained than in the Case 4 (6565 sec) although the loss of heat sink occurred earlier. The results of the CCDP analyses have indicated that consequences for different scenarios cannot be estimated without analyzing each case. Another safety concern arising in the analyzed transients is the pressurizer liquid solid condition together with the liquid discharge through the pressurizer PORVs. • Transient scenarios for different combinations of the status of the SG PORV and SV 1 for the most conservative assumptions (102 % power, MSIV and MFW trip at the beginning of the transient, AFW not credited, no operator actions) were analyzed. The critical parameters determined for the analyzed cases are the time to core dry-out and the time to pressurizer liquid solid conditions and liquid discharge, respectively. The time to core dry-out for all analyzed cases is less than 7000 sec. The shortest time to core dry-out (6290 sec) was obtained for the Case 6 (stuck open SG 1 PORV). In general, the later start of the core dry-out was obtained for the cases with stuck open SG SV 1 because of the larger amount of injected SI mass. However, there is a relatively small benefit of the SI injection on the time of the core dry-out (7000 sec for the Case 20 against 6180 sec for the hypothetical Case 21 – no SI). The pressurizer liquid solid conditions were first attained for the cases with large amount of the injected SI mass (the cases with both SG 1 SV 1 and SG 2 SV 1 stuck open). The shortest time to pressurizer liquid solid conditions was obtained for the Case 20 (3015 sec) and the latest time was obtained for the Case 2 (4020 sec).

REFERENCES

[1] V. Benčik, N. Debrecin, D. Feretić, RELAP5/MOD3.3 Analysis of Reactor Coolant Pump Trip Event at NPP Krško, Proc. Int. Conf. Nuclear Energy for New Europe 2003, Portorož, Slovenia, September 8-11, 2003, 210

[2] Precursor analyses – The use of deterministic and PSA based methods in the event investigation process of nuclear power plants, IAEA-TECDOC-1417, September 2004.

[3] D. Grgić at al., NEK RELAP5/MOD3.3 Nodalization Notebook (2000 MWt and new SGs), Report number (NEK ESD TR 09/03)

[4] T. Bajs, V. Benčik, S. Šadek, NEK RELAP5/MOD3.3 Steady State Qualification Report (Based on NEK ESD TR 09/03), Report number (NEK ESD-TR-10/03)

[5] Reactor Coolant Pump Trip event description and plant event PIS data, NEK, 2002.

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