By: Eugenijus Uspuras Algirdas Kaliatka Sigitas Rimkevicius ASME 2012 Verification & Validation Symposium May 2-4, 2012, Las Vegas, NV

Transcription

1 Computer codes validation for transient analysis at nuclear power plants with RBMK-1500 reactor By: Eugenijus Uspuras Algirdas Kaliatka Sigitas Rimkevicius ASME 2012 Verification & Validation Symposium May 2-4, 2012, Las Vegas, NV

2 OUTLINE Introduction RBMK specific Modelling of RBMK-1500 Reactor Cooling Circuit and Accident Localisation System Validation of developed Reactor Cooling Circuit model: Modelling of unstable coolant flow in the RBMK-1500 fuel channels Modelling of critical heat flux in the RBMK-1500 fuel channels Modelling of multiple fuel channels rupture Validation of developed Accident Localisation System model: Benchmarking analysis of unintended Main Steam Release Valve opening Modelling of heat transfer processes in the Condensing Tray Cooling System Conclusions

3 INTRODUCTION (1) Ignalina NPP is the only nuclear power plant in Lithuania consisting of two units, commissioned in 1983 and Both units are equipped with channel-type graphite-moderated boiling water reactors RBMK-1500.

4 Reactor INTRODUCTION (2) At Ignalina NPP Unit 1 was shutdown for decommissioning at the end of 2004 and Unit 2 at the end of At present time no plans are made to build new RBMK type reactors, but 11 RBMK reactors are still operating in Russia: 4 reactors in Saint Petersburg, 3 reactors in Smolensk, 4 reactors in Kursk RBMK heat flow diagram: 1 graphite moderator; 2 - fuel channel; 3 control rod; 4 - Drum Separator (DS); 5 - turbine; 6 - generator; 7 - condenser; 8 - condensate pump; 9 - deaerator; 10 - feedwater pump; 11 - main circulation pump 11 9

5 RBMK SPECIFIC (1) Comparison of BWR and RBMK reactor parameters No. Parameter BWR * RBMK-1000 RBMK Thermal power, MW Core diameter m Core height, m Core volume, m Mean specific power per core volume, MW/m Mean specific power per fuel quantity, MW/t Mean power per fuel rod length, kw/m * General Electric design The fuel clusters are loaded into individual channels rather than a single pressure vessel The coolant loop, consists of 1661 parallel fuel channels The plant can be refuelled on-line The neutron spectrum is thermalized by a massive graphite moderator block

6 RBMK SPECIFIC (2) General view of the RBMK-1500 reactor 1 - graphite stack, 2 - fuel channel feeder pipes, 3 - water pipes, 4 group distribution header, 5 - emergency core cooling pipes, 6 - pressure pipes, 7 - main circulation pump, 8 -suction pipes, 9 - pressure header, 10 - bypass pipes, 11 - suction header, 12 - downcomers, 13 - steam and water pipes, 14 - steam pipes, 15 - refueling machine, 16 - drum separator

7 RBMK SPECIFIC (3) Simplified schematic of the accident confinement system 1 - fuel channel; 2 - main circulation pump; 3 - suction header; 4 - pressure header; 5 - group distribution header; 6 - ECSS header; 7 - suppression pools; 8 - ALS heat exchanger; 9 - air discharge pipe section; 10 - steam pipe from MSVs and SDV-A; 11 - air-gas mixture from the reactor cavity

8 MODELLING OF RBMK-1500 REACTOR COOLING CIRCUIT RELAP5 Ignalina NPP model: 1 DS, 2 downcomers, 3 MCP suction header, 4 MCP suction piping, 5 MCPs, 6 MCP discharge piping, 7 bypass line, 8 MCP pressure header, 9 GDHs, 10 lower water piping with isolation control valve, 11 reactor core inlet piping, 12 fuel channels, 13 reactor core outlet piping, 14 steamwater communication line, 15 steam line, 16 check valve

9 MODELLING OF RBMK-1500 ACCIDENT LOCALISATION SYSTEM ENVIR AVC-L13 & AVC-L14 GDC1 GRC1 GRC2 GDC2 AVC-R13 & AVC-R14 AVC-L11 & AVC-L12 PSSL5 COLL-L5 TSRC-L TSRC-R COLL-R5 PSSR5 AVC-R11 & AVC-R12 PSSL4 AVC-L9 & AVC-L10 AVC-L7 & AVC-L8 PSSL3 COLL-L4 COLL-L3 BSRC-L9 & BSRC-L10 BSRC-L7 & BSRC-L8 Reactor BSRC-R9 & BSRC-R10 BSRC-R7 & BSRC-R8 COLL-R4 COLL-R3 PSSR3 PSSR4 AVC-R9 & AVC-R10 AVC-R7 & AVC-R8 AVC-L5 & AVC-L6 PSSL2 PSSL1 COLL-L2 BSRC-L5 & BSRC-L6 PBB5 PBB21 Rupture discs PBB22 PBB10 BSRC-R5 & BSRC-R6 COLL-R2 PSSR2 PSSR1 AVC-R5 & AVC-R6 AVC-L3 & AVC-L4 COLL-L1 BSRC-L3 & BSRC-L4 BSRC-R3 & BSRC-R4 COLL-R1 AVC-R3 & AVC-R4 HCC-L BSRC-L1 & BSRC-L2 PBB14 PBB9 PBB11 BSRC-R1 & BSRC-R2 HCC-R Model of entire Accident Localization System for COCOSYS code

10 to AVC MODELLING OF RBMK-1500 ACCIDENT LOCALISATION SYSTEM ENVIR V1 V2 GRC GDC -L13 -L14 COLL1 TSRC COLL2 -L11 -L12 V5 PSS1 PSS2 P3SP HCC C3 V3 P1P C1 V4 P2P C2 Model of activated left side tower of ALS in the case of unintended Main Steam Release Valve opening

11 VALIDATION OF DEVELOPED REACTOR COOLING CIRCUIT MODEL

12 Flow rate, kg/s Temperature, o C Modelling of unstable coolant flow in the RBMK-1500 fuel channels (1) The slow decreasing of coolant flow rate in single GDH with constant power was modelled. It leads to the flow oscillation in fuel channels and fuel claddings and fuel channels wall temperature excursion Lower water piping Centre of the core m from the bottom 2.25m from the bottom 3.25m from the bottom 4.25m from the bottom 5.25m from the bottom 6.25m from the bottom 1 2 Beginning of flow instability Time, s Time, s Coolant flow rate in the FC of maximum power Fuel cladding temperature behaviour in the FC of maximum power: 1 saturated nucleate boiling flow regime, 2 saturated film boiling regime, 3 supercritical twophase or single phase steam flow regime

13 Flow rate, kg/s Modelling of unstable coolant flow in the RBMK-1500 fuel channels (2) The coolant flow instabilities depend not only on the power of fuel channel, but also on the pressure loss in Isolation and Control Valve (ICV). In case if ICV is open, the pressure drop on it is smaller, the coolant flow rate through the channel increases, and the influence of core resistance to the coolant flow rate instability increases. In opposite case (ICV partially closed) - the influence of core resistance to the coolant flow rate instability decreases Channel power, MW Coolant flow stability boundaries in fuel channel of RBMK-1500: 1, 2 RBMK designers data, 3 data calculated using RELAP5 code (for completely opened ICV), 4 data calculated using RELAP5 code (for partially closed ICV)

14 Modelling of critical heat flux in the RBMK-1500 fuel channels (1) The slow decreasing of coolant flow rate in single GDH with constant power was modelled. It leads to the flow oscillation in fuel channels and fuel claddings and fuel channels wall temperature excursion. The simulation results showed that in the wide region of coolant quality (x = ) the RELAP5 calculation provide more conservative results, in comparison to RBMK designers experimental data. The values of CHF, calculated using RELAP5 is similar to values, calculated using Hench-Levy correlation, included in ATHLET thermal-hydraulic best estimate code. To improve situation, it is recommended to use Osmachkin correlation for the determination of CHF, instead of the Groeneveld table lookup method. However, in the case of high coolant quality (x > 0.6) the results of RELAP5 calculations show a good agreement with the experimental data. It happens because the boiling crisis at high coolant quality leads to coolant flow instability in the parallel fuel channels. Many authors underline that oscillatory coolant flow in boiling channels leads to significant early appearance of CHF in the entire length of the fuel channels.

15 Coolant flow rate, m 3 /h Modelling of critical heat flux in the RBMK-1500 fuel channels (2) Critical heat flux, MW/m Experimental data RELAP5 Hench-Levy Osmaskin Fuel channel power, MW The boundary of CHF in the fuel channels of RBMK-1500: 1 CHF boundary according RBMK designers experimental data, 2 CHF starts to appear in the top part of fuel channel (calculation data using RELAP5 model), 3 CHF appears in the entire length of the fuel channel (calculation data using RELAP5 model) Quality Distribution of the Critical Heat Flux (CHF) predicted with simulation using RELAP5 and RBMK designers experimental data

16 Modelling of multiple fuel channels rupture (1) During the history of RBMK type plant operation there were three FC rupture incidents due to overheating of single FC in RBMK-1000 reactors. The first incident occurred at the Leningrad NPP Unit 1 in 1975, the second one at the Chernobyl NPP Unit 1 in 1982, caused by a flow-rate decrease following staff mistake. The third incident occurred in 1992 in the Leningrad NPP Unit 3. The measured data of last incident are used for validation of system thermal-hydraulic codes. Before the incident, the reactor operated at 3150 MW thermal power, the initial power of affected channel was assumed equal to 2.0 MW, and the coolant flow rate through channel was 5.47 kg/s. The decrease of coolant flow rate through channel leads to fuel channel overheating and channel tube rupture. Further investigations showed that the channel had ruptured in the upper part of the core approximately 6 m above the bottom. The graphite rings around the rupture location were partially destroyed and the graphite block was damaged. A breach in the channel tube and the surrounding graphite had appeared. The fuel rods were mechanically bowed in direction to the breach.

17 Temperature, o C Temperature, o C Temperature, o C Fuel temperature in pellet centre Reactor shutdown Modelling of multiple fuel channels RELAP5/Mod3 KOBRA2 SIMMER-II 4.25m from bottom (LEI model) 5.25m from bottom (LEI model) Time, s RELAP5/Mod3 KOBRA2 SIMMER-II 1.75m from bottom (LEI model) 4.25m from bottom (LEI model) rupture (2) RELAP5/Mod3 KOBRA2 SIMMER-II 1.75m from bottom (LEI model) 4.25m from bottom (LEI model) Time, s Fuel cladding temperature Time, s Fuel Channel tube temperature Leningrad-3 RBMK NPP: sketch of the damaged fuel channel following the 1992 accident

18 VALIDATION OF DEVELOPED ACCIDENT LOCALISATION SYSTEM MODEL The transient event at Ignalina NPP (Unintended Main Steam Release Valve opening) is analysed. Comparison between the calculated results and the available data measured during this transient event is presented.

19 Water temperature, 0 C CTCS water temperature, 0 C Water temperatures in the ALS Water temperature in the top (5 th ) condensing pool 5000 W/(m 2 K) 1000 W/(m 2 K) 2500 W/(m 2 K) Time, min TN26T52 TN26T53 Water temperature at the outlet from heat exchangers of Condensing Tray Cooling System (CTCS) W/(m 2 K) 5000 W/(m 2 K) 2500 W/(m 2 K) Symbol measured; Solid line Calculated Time, min TN60T02 The calculated water temperature in the condensing pool and at the outlet from CTCS gives the best correspondence with measured results in the case of heat transfer coefficient of 2500 W/m 2 K in the heat exchangers of the CTCS. This value of heat transfer coefficient was used for best-estimate analysis of ALS behaviour in the case of loss of coolant accidents in the frames of the Safety Analysis Report of Ignalina NPP.

20 Water level, m The water level behaviour in the top (5 th ) condensing pool of ALS TN26L52 RALOC4, BAL_DRAIN RALOC4, DRAIN_BOT COCOSYS Time, min Symbol measured; Solid line Calculated by RALOC and COSOSYS codes The improved water flow model in COCOSYS code allowed to simulate the water flow through the gaps above the floor correctly even when they are totally under the water.

21 CONCLUSIONS The models of RBMK-1500 Reactor Cooling Circuit and Accident Localisation System were developed using RELAP5 and COCOSYS codes The validation of developed models were performed before the application for safety assessment of Ignalina NPP The RELAP5 Mod3.2 model of Reactor Cooling Circuit was validated for the unstable coolant flow and critical heat flux in the parallel fuel channels phenomena and multiple fuel channels rupture event in the Leningrad NPP Unit 3 Comparison between the calculated results and the available data measured during the transient event at Ignalina NPP (unintended MSV opening) showed that COCOSYS code adequately represents the occurring transient processes and it may be applied for the analysis of Accident Localisation System of Ignalina NPP. The good agreement of calculation results with the measurement data and results of other authors calculation, shows that the main physical phenomena are adequately modelled, and the developed models are suitable for the analysis of processes in the RBMK-1500 fuel channels, reactor cooling system and accident localisation system.

22 Thank you for your attention Questions?