PSA ANALYSIS FOCUSED ON MOCHOVCE NPP SAFETY MEASURES EVALUATION FROM OPERATIONAL SAFETY POINT OF VIEW

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1 International Conference Nuclear Energy in Central Europe 2001 Hoteli Bernardin, Portorož, Slovenia, September 10-13, 2001 www: tel.: , fax: Nuclear Society of Slovenia, PORT2001, Jamova 39, SI-1000 Ljubljana, Slovenia PSA ANALYSIS FOCUSED ON MOCHOVCE NPP SAFETY MEASURES EVALUATION FROM OPERATIONAL SAFETY POINT OF VIEW ABSTRACT Ivan Cillik, Lubos Vrtik VUJE - Trnava Inc. Division of Nuclear Safety Okruzna 5, Trnava, Slovak Republic cillik@vuje.sk, vrtik@vuje.sk Mochovce NPP consists of four reactor units of WWER 440/V213 type and it is located in the south-middle part of Slovakia. At present first unit operated and the second one under the construction finishing. As these units represent second generation of WWER reactor design, the additional safety measures (SM) were implemented to enhance operational and nuclear safety according to the recommendations of performed international audits and operational experience based on exploitation of other similar units (as Dukovany and J. Bohunice NPPs). These requirements result into a number of SMs grouped according to their purpose to reach recent international requirements on nuclear and operational safety. The paper presents the bases used for safety measures establishing including their grouping into the comprehensive tasks covering different areas of safety goals as well as structural organization of a project management of including participating companies and work performance. More, results are given regarding contribution of selected SMs to the total core damage frequency decreasing. 1 INTRODUCTION It is known that to finish Mochovce NPP construction a list of SMs was created to improve plant operational safety and reliability. The SMs list was based on WWER 440/V213 (J. Bohunice and Dukovany) operational experience and different international audits performed directly on the site. Below are given the most important documents as the result of audits and recommendations to satisfy the international requirements on nuclear safety and operational reliability: 1. Mission for safety improvements checking on Mochovce NPP, IAEA WWER-SC-102, September Safety improvements evaluation for Mochovce NPP rev. 16, Risk audit report, IPSN/GRS, December

2 SAFETY MEASURE LIST The list of Safety Measure Technical Specifications was originally prepared by VUJE Trnava Inc. together with Mochovce utility to give comprehensive overview of the work to be done. The final result represents a product covering all areas of the plant design including different types of analyses, operational procedures, data collection and information flows as well as monitoring of plant influence on environment and post accident management. To close contracts with different suppliers technical specifications of SMs were converted into SM contracts. Different items were grouped into the areas of technical interest with the precise definition of SM interfaces and information flows. SMs represent one part of the all work to be done on the site covering analytical and design work and supporting designers, suppliers and constructors activities. The major part of the work was performed before plant construction finishing and start-up to power operation. Final list of subcontractors consists of the main Slovak companies SE Inc., VUJE-Trnava Inc., VUEZ-Tlmace, Germany company Siemens, France company FRAMATOME, Czech companies SKODA-Prague, SKODA-Plzen, Energoprojekt-Prague and selected Russia companies. Siemens and FRAMATOME created consortium EUCOM working mainly under Germany budget. The list of SMs is presented below according to areas of common technical background: A. GENERAL G01: Classification of components G03: Reliability analysis of safety class 1 and 2 system G02: Qualification of equipment B. REACTOR CORE RC01: Prevention of uncontrolled boron dilution C. SYSTEM INTEGRITY CI01: RPV embrittlement and its CI02: Non-destructive testing monitoring CI03: Primary pipe whip restrains CI04: Steam collector integrity CI05: SG tubes integrity CI06: SG feedwater distribution integrity D. SYSTEMS S01: Primary circuit cold overpressure protection S02: Mitigation of a SG primary collector break S04: PORV and PSRVs qualification for water flow S03: Reactor coolant pump seal cooling system S05: ECCS sump screen blocking risk S06: ECCS suction line integrity S07: ECCS heat exchanger integrity S08: Power operated valves on the ESSC injection lines S09: Feedwater supply vulnerability S10: SG relief valves qualification for valves S11: SG relief valves performance at low pressure S12: Emergency feedwater make-up procedure S13: SG level control valves qualification on water flow S14: Primary circuit venting under accident conditions S15: Essential service water system S16: Main control room ventilating system S17: Hydrogen removal system E. I&C I&C01: I&C reliability I&C02: Safety system actuation design

3 I&C03: Review of reactor scram initiating signals I&C04: Physical and functional separation of main and emergency control room I&C06: Primary circuit diagnostic systems I&C05: Mechanical equipment condition monitoring I&C07: Reactor vessel head leak monitoring I&C08: Accident monitoring system instrumentation I&C09: Technical support centre I&C10: Water chemistry control and monitoring I&C11: Changing of HINDUKUS and VK3 F. ELECTRIC SYSTEMS El01: Start-up logic for the emergency El02: Diesel generators reliability diesels El03: Protection signals of emergency El04: On-site power supply accident diesel generators management El05: Emergency battery discharge time El06: Reliability of outside top transformer El07: Reliability of common top transformer system G. CONTAINMENT CONT01: Bubbler condenser strength response (max. pressure difference) under LOCA CONT02: Bubbler condenser thermodynamic response CONT03: Containment leak rates CONT04: Maximum pressure differences on walls between compartments of hermetic boxes CONT05: Containment peak pressure & activation of sub-atmospheric pressure after blowdown CONT06: Experimental verification of thermodynamic phenomena and of BC under LOCA H. INTERNAL HAZARDS IH01: Systematic fire hazards analysis IH02: Fire prevention IH03: Fire detection IH04: Extinguishing IH05: Mitigation of fire effects IH06: Systematic flooding analysis IH07: Turbine missiles IH08: Internal hazards due to high energy pipe breaks IH09: Heavy load drop I. EXTERNAL HAZARDS EH01: Seismic design EH02: Analysis of plant specific natural external EH03: Man induced external events J. ACCIDENT ANALYSIS AA01: Scope and methodology of accident analysis AA02: QA of plant data used in accident analysis AA03: Computer code and plant model validation AA04: Availability of accident analysis results for supporting plant operations AA05: Main steam line break AA06: Overcooling transients related to pressurized thermal shock AA07: Steam generator collector rupture AA08: Accidents under low power and shutdown (LPS) conditions AA09: Severe accidents AA10: Probabilistic safety assessments (PSA) AA11: Boron dilution accidents

4 305.4 AA12: Spent fuel cask drop accidents AA13: ATWS K. OPERATION OP01: Procedures for normal operation OP02: Emergency operating procedures OP03: Limits and conditions OP04: Safety culture OP05: Experience feedback OP06: Quality assurance program OP07: Data and document management OP08: Philosophy on use of procedures OP08: Philosophy on use of procedures OP09: Communication system OP10: Radiation protection and monitoring OP11: Emergency centre 3 ASSESSMENT OF SAFETY MEASURES ON THE BASES OF PSA MODELS Detail Level 1 FPSA for Mochovce unit 1 was performed to present contribution of implemented SMs from the above list to the core damage frequency (CDF) decreasing. Three statutes of Mochovce plant upgrades were considered for power level 1 PSA model performance as follows: 1. The first PSA model was prepared for Mochovce unit 1 status before SMs implementation (original plant design) considering old event oriented emergency operating procedures without feed & bleed procedure. Feed & bleed procedure was finally considered as well to show positive contribution of this reactor core cooling and residual heat removal procedure to the total core damage frequency. 2. The second PSA model was prepared for Mochovce unit 1 status after the unit start-up including implemented SMs and new EO EOPs considering feed & bleed procedure (EO EOP-HP201). This model represent unit 1 status during the first year of its operation. 3. The third PSA model was realised for Mochovce unit 1 status after the unit first refuelling including all implemented SMs and considering its operation under SB EOPs which were trained by operators one year before their implementation. 3.1 Analysis of Benefit of Safety Measures from the Risk Decrease Point of View SMs analysed by the Mochovce FPSA model are presented above. Mochovce FPSA model considered internal events as LOCA accidents, transients, internal fires and floods. The list of the considered initiating events is given in the table below. This list of initiating events is the same for all analysed PSA models as it was presented above. Level 1 PSA model was calculated using WWER specific data based mainly on V-2 plant operational experience. Table 1: Initiating event categories for the Mochovce PSA study IE Category Frequency (epy) Description L6-loop 1,2,3,5,6 5.60E-5 Large LOCA ( mm) loop 1,2,3,5,6 L6-loop E-5 Large LOCA ( mm) loop 4 L5 8.00E-5 Large LOCA ( mm) L4 1.50E-4 Medium LOCA ( mm) L3 1.50E-4 Medium LOCA ( mm) L2 5.50E-4 Medium LOCA (20-60 mm) L1 1.10E-3 Small LOCA (0-20 mm) PSL 4.10E-4 Pressurizer steam LOCA IFSL 5.00E-4 Interfacing system LOCA PLOCA 1.00E-4 Interfacing pool LOCA SGTR 8.90E-4 Steam generator tube rupture 2TG 3.80E-1 Both TG trip

5 305.5 IE Category Frequency (epy) Description IRT 1.80E-1 Inadvertent reactor trip LMF 9.20E-2 Loss of main feedwater LMF(FWHB) 1.40E-3 Main feedwater header break LMF(FWTB) 1.60E-3 Feedwater tank break LOF6 1.40E-1 Loss of four or more MCPs LOCW 1.20E-1 Loss of circulating cooling water RAT 2.00E-2 Reactivity addition transient SHB 1.40E-3 Steam header break SLBI 1.30E-3 Steam line break inside confinement SLBO 9.00E-4 Steam line break outside confinement LOP 3.10E-2 Loss of offsite power FIRE E-2 Fire in the TG hall before improvements pre-modification state FIRE-490 post-modification state 6.80E-4 Fire in the TG hall after improvements 3.2 Results of Pre-modification Status The results of the first pre-modification phase of PSA modelling are presented for two basic cases: without and with the consideration the primary feed & bleed operation. In the second case non-qualified pressurizer safety valves would be used for feed & bleed given total loss of primary to secondary side heat removal. Results without the Consideration of the Primary feed & bleed Operation Using the PSA model, the following mean core damage frequency was calculated without consideration the primary feed & bleed operation: 1.03E-3 y -1 Selected contribution of initiating events, categories of basic events and systems to the core damage frequency is presented below: Contribution of Dominant Initiating Events Main feedwater header break LMF(FWHB) 30.19% Fire in the TG hall (FIRE-490) 25.15% Steam header break (SHB) 21.75% Steam line break outside confinement (SLBO) 13.98% Contribution of Dominant Basic Event Categories Post trip operator actions are represented in cut sets that account for 33.9% of the total CDF: operator fails to initiate the EFW system 32.1% manual fire suppression in the TG hall 26.2% operator fails to initiate demi-water 1 MPa system 2.6% Another Important Categories of Primary Events turbine hall effect 65.8% hardware failure 2.8% common cause failures 1.4% Contribution of Dominant Systems The importance of system contribution to the core damage frequency is following: Emergency feedwater system 32.3% Demi water system 1 MPa 2.8% Auxiliary feedwater system 1.9%

6 305.6 Results with the Consideration of the Primary feed & bleed Operation The mean core damage frequency is: 6.54E-5 y -1 The next part of the section presents the contribution of initiating events, categories of primary events and systems to the core damage frequency. Contribution of Dominant Initiating Events Main feedwater header break LMF(FWHB) 22.17% Fire in the TG hall (FIRE-490) 19.11% Steam header break (SHB) 15.90% Steam line break outside confinement (SLBO) 10.26% Small LOCA (L2) 8.53% Interfacing LOCA (IFSL) 7.31% Contribution of Basic Event Categories Post trip operator actions are represented in cut sets that account for 95.8% of the total CDF: operator fails to initiate feed & bleed 55.4% operator fails to prevent overflow of LPSI tanks 35.4% operator fails to initiate the EFWS 23.4% Another Important Categories of Primary Events turbine hall effect 48.6% common cause failures 4.4% hardware failure 3.6% Contribution of Dominant Systems The importance of system contributions to the core melt frequency is following: Primary circuit ( feed & bleed ) 55.5% Low pressure injection system 35.6% Emergency feedwater system 23.6% 3.3 Results of Post-modification Status with EO EOPs The results of the first post-modification phase of PSA modelling are presented in this section. The mean core damage frequency is: 1.91E-5 y -1 The next part of the section presents the contribution of the initiating events, categories of basic events and systems to the core damage frequency. Contribution of Dominant Initiating Events Fire in the turbine hall (FIRE-490) 50.58% Steam generator tube rupture (SGTR) 11.20% Loss of offsite power (LOP) 11.05% Loss of circulating cooling water (LOCW) 7.28% Contribution of Dominant Basic Event Categories Post trip operator actions are represented in cut sets that account for 83.9% of the total CDF. operator fails to initiate the EFW system 71.7% operator fails to initiate primary bleed and feed 69.9% manual fire suppression in the TG hall 50.8% operator fails to isolate the SGTR 7.6% Another Important Categories of Primary Events: hardware failure 33.3% common cause failures 14.4% turbine hall effect 5.6% Contribution of Dominant Systems The importance of system contribution to the total core damage frequency is as follows:

7 305.7 Emergency feedwater system 72.3% Primary circuit 70.1% Emergency power supply of category II 23.6% High pressure injection system 6.7% Demi-water 1 MPa system 6.0% Intermediate cooling system of ECCS 5.4% Auxiliary feedwater system 5.1% 3.4 Results of Post-modification Status with SB EOPs The mean core damage frequency is: 9.13E-7 y -1 The next part of the section presents the contribution of the initiating events, categories of primary events and systems to the core damage frequency. Contribution of Dominant Initiating Events Large LOCA ( mm) (L5) 20.37% Small LOCA (0-20 mm) (L1) 17.52% Interfacing LOCA (IFSL) 11.72% Steam generator tube rupture (SGTR) 10.74% Contribution of Main Basic Event Categories The hardware failures represented by cut sets covering 95.4% of the total CDF: common cause failures of ECCS intermediate cooling system 38.7% single failures of intermediate cooling system of ECCS 19.1% single failures of low pressure injection system 11.7% single failures of engineering safety fast actuating system 8.7% common cause failures of low pressure injection system 7.0% Another important categories of basic events: human errors 23.6% common cause failures 56.3% Contribution of Main Systems The importance of system contribution to the core melt frequency is the following: Intermediate cooling system of ECCS 58.3% Low pressure injection system 20.7% Engineering safety fast actuating system 8.8% Confinement spray system 7.8% 4 CONCLUSIONS The results presented above shows very transparently that the preparation, evaluation, selection and implementation of presented safety measures list to enhance Mochovce safety level based on the different international audits and missions and considering operational experience of VEER reactors substantially contribute to the VVER reactor safety. In the combination with SB EOPs implementation into the plant operation they shift plants with VVER reactors into comparable position with western designed NPPs regarding operational and nuclear safety.

8 305.8 REFERENCES [1] I. Cillik et al., Evaluation of Realized Safety Measures to Mochovce Unit 1 Start-up with Indication of Safety Measure Contribution to Core Melt Frequency, VUJE, Trnava, 1998 [2] VUJE Inc., Level-1 PSA of Mochovce Unit 1 NPP for Post-modification State of the Unit SB EOP, VUJE, May 2000, Report No. 2572/96