Primary Loca in Vver-1000 by Pressurizer Porv Failure

BNS International Conference “Nuclear Energy in Europe – Present and Future”,

18-21 September 2013, Bulgaria

PRIMARY LOCA IN VVER-1000 BY PRESSURIZER PORV FAILURE

L. Sabotinov1, S. Lutsanych2, I. Kadenko3

1 Institut de Radioprotection et Sûreté Nucléaire (IRSN),

PSN-RES, SEMIA, BAST, Fontenay-aux-Roses 92262 France

2 San Piero a Grado Nuclear Research Group (GRNSPG), University of Pisa

Via Livornese 1291, 56122, San Piero a Grado, Pisa, Italy

3 International Nuclear Safety Centerof Ukraine (INSC), University of Kiev

Prospekt Akad. Glushkova 2, Kiev, 01033, Ukraine

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ABSTRACT

The paper presents the calculations and analysis of the design basis accident of a standard VVER-1000/V320 reactor with inadvertent opening and stuck in open position of the pressurizer pilot operated relief valve (PORV). The objective is the independent assessment of this accident applying the French best estimate thermal-hydraulic computer code CATHARE 2 and verification to meet the safety criteria for such kind of the accident.

The “Inadvertent opening and stuck in open position of PORV” is a design basis accident classified as Medium Break Loss of Coolant Accident (MB LOCA) with the equivalent diameter of the break D= 68 mm. This accident is particularly interesting to be calculated and analyzed, because it took place at operating NPP with VVER-1000 reactors (Rovno NPP) in 2009.

The calculations have been carried out with conservative conditions as usual for DBA analysis. The NPP model corresponds to a real VVER-1000/V320 configuration and comprises all safety systems, adopted for one of the latest CATHARE 2 versions.

The results of CATHARE 2 calculations are compared with available results of RELAP 5 calculations. There is similarity of the thermal-hydraulic parameters behavior, but also some differences can be observed basically due to the break flow prediction, which causes differences in primary pressure evaluation. Both calculations show that there is no boiling crisis in the reactor core and reliable cooldown is achieved.

The calculations performed with CATHARE 2 code demonstrate the ability of the code to predict reasonably the break flow, pressures, temperatures etc. for considered LOCA scenario and to be applied for safety studies.

1.  INTRODUCTION

The study of MB LOCA with CATHARE 2 code for standard VVER-1000/V320 reactor with conservative conditions of Design Basis Accident (DBA) was carried out. The objective of the study was to perform independent in-depth safety analysis with CATHARE 2 code during MB LOCA accident with inadvertent opening and stuck in open position of the PORV regarding the safety criteria and compares the results with the RELAP 5 predictions of INSC-Ukraine.

2.  CATHARE 2 COMPUTER CODE

CATHARE 2 (Code for Analysis of Thermal-Hydraulics during an Accident of Reactor and safety Evaluation) [1, 2] was developed to perform best-estimate calculations of pressurized water reactor accidents: PWR loss of coolant (large or small break, primary and secondary circuit). Specific modules have also been implemented to allow modeling of other reactors like VVER, BWR, sodium and gas cooled reactors. It was developed in Grenoble by the Commissariat à l’Energie Atomique (CEA), Electricité de France (EDF), AREVA and Institut de Radioprotection et de Sûreté Nucléaire (IRSN).

CATHARE 2 includes several independent modules [3, 4] that take into account any two-phase flow behavior like thermal and mechanical non-equilibrium, vertical co- or counter-current flow, flooding counter-current flow limitation (CCFL), horizontal stratified flow, critical flow, cold water injection, super-heated steam, reflooding etc. and all flow regimes along with all heat transfer mechanisms.

In order to take into account these phenomena the CATHARE 2 code is based on a two-fluid six equation model with a unique set of constitutive laws. Various modules offer space discretization adapted to volumes (0D), pipes (1D) or vessels (3D) ready to assemble for any reactor design or test facility. CATHARE 2 is limited to accidents during which no severe damage with fuel melting occurs.

3.  MODELING OF VVER-1000 (V320) BY CATHARE

The CATHARE 2 Input Data Deck of VVER-1000 reactor system for MB LOCA is based on IRSN models, developed, verified and used for calculation of various accidents in order to assess the safety aspect of the VVER reactor facilities [5, 6,7].

The general layout of the primary circuit is presented in Figure 1.

Figure 1: Primary circuit of simulated VVER-1000/V320

Figure 1 shows loop №1 of main coolant circuit (MCC). The hot leg nozzle (BC1C-BC4C) for each loop of VVER-1000/V320 is located on the reactor pressure vessel directly above the cold leg nozzle for each corresponding primary loop. The distance between the centers of the hot and cold leg is 1800 mm. The hot leg provides coolant to the SG hot collector GV1IN1.

Primary coolant flows from the cold collector (GV1OUT) to upper part of the annular collector (VOLDOWN). Then it follows to the downcomer and lower plenum (PLENINFI). PLENINFS is a volume between PLENINFI and the core.

The reactor core consists of 3 main parts: average core (ACORE), hot assembly (HCORE) and bypass (CBYPASS). ACORE is modeled by CATHARE with axial element and represents the 162 thermal-hydraulic channels. The axial element is of the type rod bundle with 40 axial segments. The thermal-hydraulic weight of ACORE is 50544. HCORE depicts the hottest assembly of reactor and has the thermal-hydraulic weight equal to 312.

From reactor core, the heated coolant flows to upper plenum (PLENSUP). The four main coolant pumps (MCP) are modeled by CATHARE with one-node pump sub-module using homologous characteristics for VVER-1000 pumps.

Pressurizer (PRESSU) is represented by a volume element with total height of 11.176 m (recalculated from the total volume 79 m3 and d=3 m). The bottom of the volume is connected to the hot leg (BC1C) by the surge line (EXPANS) while at the top of the pressurizer there is a spray system with piping connected to a cold leg and the make-up system.

Figure 2: Secondary circuit of simulated VVER-1000/V320

Figure 2 represents loop №1 of secondary circuit. The SG is a horizontal U-tube type with recirculation. It is modeled by two parts: axial GVDOWN1 (downward flow and riser part with heat exchange) and a volume element GVVOL1 (volume above the perforated plate).

4.  Inadvertent opening and stuck in open position of PORV accident

The “Inadvertent opening and stuck in open position of PORV” – is design based accident classified as medium coolant leak (MB LOCA). Diameter of the leak from PORV is 68 mm.

Both pressurizer safety relief valves (SVPRZ1 and SVPRZ2) are modeled with EXTERNAL SINK SAFETY VALVE sub-module based on Gros d'Aillon correlation for critical mass flowrate. Transient with opening of SVPRZ1 (Figure 3) was carried out.

Figure 3: Modeling of “Inadvertent opening of PORV” accident

4.1.  Initial conditions and the steady state calculations

For modeling of the accident, the following conservative assumptions are considered:

-  The initial reactor power is equal to 104% of nominal power (takes into account maintenance accuracy 2% and measurement error 2%);

-  Mass flow of the reactor coolant is minimal from the operational range (it provides higher core heatup)

-  Pressure in primary circuit is chosen to be minimal (takes into account pressure measurement error -0.3 MPa);

-  Water level in pressurizer is the minimum of the operational range (takes into account maintenance accuracy -150 mm and measurement error -150mm). It will provide minimal initial water inventory in primary circuit;

-  Secondary pressure in SG is chosen to be maximal (takes into account pressure measurement error +0.2 MPa);

-  Water level in SG is the minimum of the operational range. This conservative assumption provides minimal amount of water in SG and reduces heat removal from the primary side.

Table 1 contains the nominal and conservative initial conditions of NPP equipped with standard VVER-1000/V320. Also the steady state results of CATHARE 2 calculations are provided.

Table 1

Nominal and conservative initial conditions. CATHARE steady state results

Parameter / Units / Nominal operation / Initial conservative conditions / CATHARE 2 steady state calculation
Reactor thermal power / MW (%) / 3000 (100) / 3120 (104) / 3120 (104)
Maintenance accuracy / ± 60(2)
Measurement error / ± 60(2)
Reactor outlet pressure / MPa / 15.9 ± 0.3 / 15.6 / 15.52
Reactor coolant flow rate / m3/h / 84800-4800+4000 / 80000 / 80010
Reactor coolant inlet temperature / °C / 289.7+2 / 291.7 / 291.66
Reactor coolant outlet temperature / °C / 320+4 / 324 / 323.99
Reactor heat-up / °C / 30.3±3 / 33.3 / 33.13
Maximum cladding temperature / °C / 350 / 350 / 351.06
Pressurizer level / m / 8.77±0.15 / 8.47 / 8.46
Measurement error / 0.15
SG water level (wide range) / m / 2.10±0.05 / 2.10±0.05 / 2.136
SG pressure / MPa / 6.3±0.2 / 6.5 / 6.5
Feed water flow in SG1/2/3/4 / kg/s / 408.3±28.6 / 436.9 / 434.51
Steam temperature at SG outlet / °C / 278.5±2 / 280.5 / 280.87
Feed water temperature / °C / 220±5 / 227.0 / 226.68
Measurement error / ±2

4.2.  Boundary conditions and assumptions

The following conservative assumptions are considered:

-  NPP operates at full power until the reactor scram;

-  No operator actions are considered;

-  Loss of offsite power is assumed at the reactor scram signal;

-  Single failure is applied. Failure of one diesel generator is assumed. Consequently, there is dependent failure of one channel of the safety systems;

-  The pressurizer electrical heaters are not operating;

-  Two high pressure safety injection (HPSI) pumps available with a delay of 40 sec after the loss of offsite power. ECCS starts, when the primary pressure < 108.9 Bar. Conservative flow characteristics of HPSI pumps with a minimum flow rate are assumed;

-  Initial pressure in hydroaccumulator (HA) is minimal from the operational range (57 Bar). The water inventory of HA is 48.4 m3 (minimal from the range);

-  Auxiliary feedwater system (AFWS) and steam dump valve to condenser (BRU-K) are not considered.

5.  RESULTS OF CATHARE CALCULATIONS AND COMPARISON WITH RELAP

This section present the results of computational analysis of the “Inadvertent opening and stuck in open position of PORV” accident at the initial and boundary conditions formulated in 4.1 and 4.2. Analysis is performed on a time interval of 3600 seconds without taking into account actions of operating personnel. Predicted chronological sequence of events is presented in the Table 2.

Table 2

Chronological sequence of events for “Inadvertent opening and stuck in open position of the PORV” accident

Time, sec / Event
00.00 / Opening and stuck in open position of PORV. Diameter of the leak is 68 mm.
37.99 / Reactor scram is generated by signal ‘‘Reactor outlet pressure is less than 148-2.5 Bar under the Reactor Power less than 75% and Temperature in any hot leg more than 260 °C’’
Loss of offsite power (postulated event)
Stop of MCP and MFWP
Loss of power supply to BRU-A №1
39.99 / Signal for DG start up after loss of offsite power with delay 2 sec
40.33 / Closing of Turbine Stop Valve (TSV)
43.05 / Opening of the BRU-A №2/3/4 (TX60, 70, 80S05). Pressure in the Main Steam Lines > 73+1 Bar. The rest of time, BRU-A №2/3/4 control pressure in a secondary circuit
50.71 / Maximal pressure in secondary circuit 82.1 Bar
55.78 / Opening of BRU-A №1. Pressure in the Main Steam Lines > 73+1 Bar. The rest of time, BRU-A №1 controls pressure in a secondary circuit
175.43 / HPSI (TQ13, 33D01) start. Pressure in the primary circuit < 108.9 Bar
3600.00 / End of calculation

Inadvertent opening of PORV leads to primary coolant leakage from pressurizer to the bubbler tank. It causes a rapid decrease of the primary pressure (Figure 4) and intensive coolant boiling (flashing) in the pressurizer. The calculated initial value of coolant flow from PORV is 38 kg/s (see Figure 5).

4

Figure 4: Pressure Upper Plenum, CATHARE / RELAP

Figure 5: PORV and HPSI flows, CATHARE / RELAP

4

At time = 37.99 sec, due to generated signal ‘‘Reactor outlet pressure is less than 148-2.5 Bar, Reactor Power less than 75% and Temperature in any hot leg more than 260 °C’’, reactor scram occurs. After the insertion of control and safety rods into the core, reactor power is reduces to the level of residual energy release (Figure 6). Loss of offsite power supply system occurs simultaneously with scram signal. It causes stop of main coolant pumps (MCP), main feedwater pumps (MFWP), closure of turbine stop valve (TSV) and loss of steam dump valve to atmosphere (BRU-A №1) electrical supply.

4

Figure 6: Relative exchange and reactor power, CATHARE / RELAP

Figure 7: Reactor outlet/saturation temperatures, CATHARE / RELAP

4

After the scram occurs, temperature of the primary coolant starts to decrease as well as coolant saturation temperature (Figure 7). With the break opening flashing process in PRZ occurs and swelling level rises until the reactor scram. Then some level decrease is observed. Decrease of the primary pressure < 108.9 Bar results in operation of HPSI pumps (t=175.43 sec), which try to compensate the leakage from the primary circuit (Figure 5). Therefore water level in PRZ starts rising (Fig.9). Pressure in primary circuit start to increase until the water level in PRZ reach maximum. Liquid and gas flow rates through the opened PORV are illustrated in Figure 8.

4

Figure 8: Liquid and gas flows through the opened PORV,

Figure 9: Pressurizer water level,

4

In Figure 10 is presented the secondary pressure, which increases rapidly after the closure of the MFWP and TSV. Consequently, pressure reaches the opening set-point of BRUA-2, 3, 4 (74 Bar). At time = 55.78 sec, the power supply is recovered. BRUA-1 activates due to pressure in MSL №1 > 73+1 Bar. Pressure in MSL doesn’t reach the opening set-points of safety valves of steam generator SGSV-1, 2.

The maximum cladding temperature is the initial temperature 347.9 ºC (Figure 11).

4

Figure 10: Pressure MSL №1, CATHARE / RELAP

Figure 11: Maximum cladding temperature, CATHARE / RELAP

4

Observed differences in the results are basically due to the different break flow prediction, which causes difference in primary pressure evaluation and following systems operation.

Results of CATHARE calculation show that during the accident “Primary LOCA with pressurizer PORV failure” the reliable core cool-down is observed. Minimum DNBR = 1.74 at time = 40 sec (Figure 12).