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Ref. to Ilona Lindholm, Keijo Valtonen

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1 Ref. to Ilona Lindholm, Keijo Valtonen
STUK- independent Regulatory Body for nuclear and radiation safety in Finland and the use and role of TSO’s VN/RA/01 Task 1&2 Workshop Hanoi, October 2012 Confidential Ilari Aro Ref. to Ilona Lindholm, Keijo Valtonen STUK

2 Nuclear power programme in Finland
Fennovoima Ltd New utility, no operating reactors DiP approved for FA1, Hanhikivi Site Loviisa NPP (Fortum) 2 operating units – VVERs Interim Spent Fuel Storage at site L/ILLW repository Olkiluoto NPP (TVO) 2 operating units - ABB BWRs OL3 (EPR) under construction DiP approved for OL4 Interim Spent Fuel Storage at site L/ILLW repository Posiva “Onkalo” Photo: Fortum Photo: TVO

3 STUK - Radiation and Nuclear Safety Authority
Mission: Protecting people, society, environment, and future generations from harmful effects of radiation

4 Organisation Regulatory Body Research Centre Expert organisation
use of nuclear reactors nuclear waste and materials use of radiation Research Centre health effects of ionising and non-ionising radiation natural radiation - occurrence and prevention environmental research radiation threats and preparedness for accidents dosimetry and metrology medical use of radiation non-ionising radiation Expert organisation national preparedness for radiation accidents information and advice to public and other authorities contracted expert services radiation measuring and calibration services Organisation Figures indicate staff number (356) at the end of 2011.

5 Organisation of Nuclear Reactor Regulation

6 Statement on nuclear safety
Main organizations involved in licensing and safety assessment in Finland Government Ministry Licenses STUK Safety assessment Statement on nuclear safety Main TSO: VTT Safety analysis

7 Typical TSO’s in Finland for safety analysis and research
VTT – Technical Research Centre of Finland Branches for all technical fields of society For nuclear and waste safety about 100 experts (altogether about 2000 staff) Main TSO for STUK Geological Institute and seismological institute for siting studies Meteorological institute for meteorology and emergency response Universities (e.g. Helsinki, Aalto and Lappeenranta universities) e.g. radiochemistry, nuclear waste, thermohydraulics and severe accident research 23 August 2010

8 Review and Assessment by the Regulatory Body
Licensee is responsible to demonstrate safety and fulfilment of safety requirements (requirements are presented in the YVL guides) STUK conducts independent review and assessment in licensing steps (before license is granted) and before modifications are implemented at the plants Focus and scope of STUK’s review and assessment depends on the licensing phase Safety assessment tools such as deterministic and probabilistic safety analyses are utilised in STUK’s review and assessment STUK has established internal guidance for review work to ensure consistent review and assessment process and application of graded approach Technical Support Organisations are utilised in specific areas (e.g. comparative accident and transient analyses), but STUK makes decisions on the safety case

9 DiD-levels, event categories and frequency of events belonging to each category
Normal operation (DBC 1) Level 2 Anticipated operational occurrences (DBC 2) f > 10-2/a Level 3a Class 1 postulated accidents (DBC 3) 10-2/a > f > 10-3/a Class 2 postulated accidents (DBC 4) f < 10-3/a Level 3b Design extension conditions (DEC) CCF rare events Level 4 Severe accidents (SA) safety goals CDF <10-5/a, LRF < 5x10-7/a Ilari Aro, 31 October 2010

10 Acceptance criteria for radioactive releases / max doses to general public
DBC 1, Normal operation radiation dose limit 0,1 mSv / year for the entire site DBC 2, Anticipated operational occurrences radiation dose limit 0,1 mSv DBC 3, Class 1 postulated accidents radiation dose limit 1 mSv DBC 4, Class 2 postulated accidents radiation dose limit 5 mSv DEC, Design extension conditions radiation dose limit 20 mSv SA, Severe accidents release < 100 TBq Cs-137 equivalent no acute health effects Ilari Aro, 31 October 2010

11 Acceptance criteria for fuel
DBC 1 , Normal operation DBC 2 , Anticipated events 95/95 confidence with respect DNB or dry-out, no (internal) fuel melting, nor damage due to pellet-cladding mechanical interaction. DBC 3 “Class 1” postulated accidents number of rods in heat transfer crisis < 1%, PCT < 650 °C, and extremely low probability of fuel damage by the mechanical interaction between fuel and cladding DBC 4 ”Class 2" postulated accidents the higher the frequency of a postulated accident, the smaller the number of damaged fuel rods. Number of damaged fuel rods < 10%. Max PCT < 1200 C. Limited enbritlement. Enthalpy limit 140 cal/g for failure (230 cal/g not be exceeded). Enthalpy limits are valid for fuel burnups up to 40 MWd/kgU. Limits for higher burnups shall be justified by experiments. No danger to long-term coolability DEC, Design extension conditions DEC A Max PCT < 1200 C. Limited enbritlement. Enthalpy limit 140 cal/g for failure (230 cal/g not be exceeded). No danger to long-term coolability DEC B, STUK September 2011/ I Aro

12 10 % FUEL FAILURE LIMIT STUK September 2011/ I Aro

13 Analyses of plant behaviour: Examples of cases to be analysed
Examples of initiating events to be analysed are: DBC 2 disturbance in the reactor power control or other disturbance, which causes a change in reactivity disturbance in primary circuit flow, pressure control or water volume control disturbance in steam pressure or steam flow disturbance in feedwater flow or feedwater temperature DBC 3,4 leaks from the primary circuit during power operation, change in operational state, refuelling and/or outage leak from secondary circuit (PWR) leak from primary to secondary circuit (PWR) STUK September 2011/ I Aro

14 Design Extension Conditions (DEC)
DEC A includes conditions in which a common cause failure (CCF) in a safety system is assumed during anticipated operational occurrence (DBC 2) or class 1 accident (DBC 3), overall frequency of an event ~ as an example ATWS station black out total loss of feed water LOCA together with the complete loss of one emergency core cooling system total loss of the CCWS total loss of the RHR loss of ultimate heat sink loss of fuel pool cooling realistic assumptions are applied for accident analysis single failure is assume in safety systems DEC B includes complex sequences and rare external events multiple stem generator tube rupture (~10) extreme weather condition large airplane crash STUK September 2011/ I Aro

15 Analyses of plant behaviour: Examples of severe accidents
Severe accident analyses shall be used to study factors which affect containment integrity, leak tightness and the operability of containment systems. They could include i.e.: total, long lasting loss of AC power total loss of feedwater leak of primary coolant without emergency cooling during power operation or a maintenance, refuelling or other outage leak of primary coolant and blockage of coolant recirculation STUK September 2011/ I Aro

16 Analyses of releases and radiation doses : Examples of postulated accidents
Separate radiation dose analyses shall be made if the dose upper limit cannot be concluded from the results of other analyses. Some examples are: Large leak of coolant from the primary circuit during power operation. A typical example of accidents during which radioactive substances are first released into the containment and gradually leak out. Leak of reactor coolant outside the containment due to an instrument line rupture Leak from steam generator primary to secondary side. The total rupture of one or multiple steam generator tubes shall be analyzed by assuming that also the safety valve of the steam generator has stuck open in a case it is expected to open. Also a leak larger than the one mentioned above shall be analyzed if estimated possible on the basis of the structure of the steam generator. Leak out of the primary circuit during a maintenance, refuelling, or other outage. STUK September 2011/ I Aro

17 Analyses of releases and radiation doses : Examples of postulated accidents (continued)
Leak outside the containment in an unisolated steam line connecting to a steam generator in which, before the initiation of the accident, the largest primary to secondary circuit leak (PWR) allowable in the Technical Specifications has occurred. Leak in a steam line outside the containment or in a reactor coolant purification line (BWR). Damage outside the containment in a system containing radioactive gases. Damage outside the containment in a system containing radioactive liquids. Damage of a fuel assembly which has been removed from the reactor. Dropping of a transfer or transport cask containing spent fuel during hoisting, in a situation where the cask is not tightly closed, or dropping of the fuel cask during transfer. Dropping of a heavy object on top of stored fuel or an open reactor. Severe accidents Analyses shall be carried out for cases which on the basis of containment behaviour and conditions and the concentration of radioactive substances in the containment are estimated to cause the most extensive releases. STUK September 2011/ I Aro

18 Reliability of the analysis methods
Methods of analysis mean i.a. methods based on hand calculations, computer programs and the application of experimental data. The reliability of the analysis methods used shall be justified. A description of the analysis methods used shall be given, including their general principles as well as the physical models and numerical methods used. The experimental correlations used in the calculations shall be justified by presenting the measurement data from which the correlations have been derived. If the correlation is commonly known and the measurement data are publicly available, a bibliographic reference is sufficient. The analysis methods shall be adequately verified for the treatment of the events in question. Both numerical methods and physical models shall be verified. Numerical methods shall be verified by adequate reference calculations. Physical models shall be verified by demonstrating their ability to depict suitable separate effects tests or integral tests for complete systems or nuclear power plant transients. In addition, comparison with other, earlier verified models may be utilised. If sufficiently reliable calculation methods are not available, the analysis shall be justified by experiments. This requirement applies especially to most phenomena essentially relating to severe accident management, for example, the long term coolability of reactor core debris after a severe accident. October 2012 / IA

19 Performing safety analysis
Limiting transients or accident cases are selected from the point of view of fuel design criteria or pressure vessel design criteria or for containment analysis purposes. The purpose is to check that design values are according to the regulatory requirements, input data has been properly selected and that there are enough conservatism and design margin in design. Also the analysis models are checked by using comparison analyses with different analysis tools. The ideal case is to use different calculation codes and models. In most cases this also takes place. In this respect, STUK has been quite successful. VTT in Finland has the role of developing analysis tools or transform them into the Finnish conditions, test their validity and verification and also to perform safety analysis when needed. VTT validates the codes e.g. by performing some benchmark calculations against experiments or as international co-operation. In Finland, originally, many computer codes used were of US design and it was necessary to apply them into the reactor systems that were different from the original purposes. For this reason, VTT was needed for this basic design and verification work. Currently, VTT has developed its own analysis tool system that is efficient in the safety analysis. October 2012 / IA

20 Performing safety analysis
Analysis tools are mainly provided by VTT or via VTT to STUK. STUK has also developed some computer codes by itself: e.g. for PSA and for radiation safety calculations. If necessary, STUK requires from the licensee also the analysis tools and the input data used for its review in addition to the analysis reports and results. For the safety assessment purposes, different computer codes and models are mainly used. Power companies typically use codes developed by the plant vendor e.g. ABB computer codes for BWR or they have supported the development of own codes as Fortum/VTT codes for VVER. STUK mainly uses codes developed by VTT or which have been received through international co-operation e.g. with US NRC. October 2012 / IA

21 Codes used for Olkiluoto BWR
ANALYSIS LICENSEE REGULATOR Core calculation Polca / Updat not performed Transient analys. Bison/Ramona(3D) TRAB / Ramona4 LOCA analysis Coblin / Dragon RELAP5 Containment Copta CONTEMPT Structural analys. commercial codes Severe accident MAAP / MELCOR MELCOR Radioact. release Meteorology TUULET ARANO, VALTO SILAM (FMI) PSA SPSA October 2012 / IA

22 Codes used for LOVISA - VVER
ANALYSIS LICENSEE REGULATOR Core calculation HEXBU not performed Transient analys. HEXTRAN (3D) TRAB (3D) LOCA analysis APROS RELAP5 / (APROS) / FRAP-TRAN Containment CONTEMPT Structural analys. commercial codes Severe accident MAAP / MELCOR MELCOR Radioact. release Meteorology TUULETV2001/MERI2002 / TUULET (acc) ARANO, VALTO SILAM (FMI) PSA RiskSpectrum / SPSA SPSA October 2012 / IA

23 What to assess - plant level
Nuclear safety analysis Plant functional design (anticipated transients, design basis accidents, design extention conditions, severe accidents, PSA, radioactive releases, environmental effects) Site evaluation and environmental aspects Plant design - system analysis system design features and description - compliance with design criteria safety classification, QA/QC requirements, regulatory control system failure analysis - redundancy, diversity, separation, interaction with other systems, risk assessment system operation - functional requirements; operation during different plant states including accidents environmental conditions & classification operational limits and conditions, surveillance and testing October 2012 / IA

24 Criteria for safety assessment (plant level)
General design criteria basis for safety assessment review report YVL guides YVL 1.0 Safety criteria for design of nuclear power plants YVL 2.2 Transient and accident analyses for justification of technical solutions at nuclear power plants YVL 2.8 Probabilistic safety analyses (PSA) YVL 6.2 Fuel design limits and general design criteria YVL 7.1 Limitation of public exposure in the environment of and limitation of radioactive releases from nuclear power plants YVL guide 2.0 gives criteria for the design of safety systems YVL 2.1 cover safety classification YVL 2.7 cover failure criteria YVL 1.4 cover QM YVL 2.5 cover pre-operational and start-up testing of NPP YVL guide system provide detailed criteria for structures, systems and components October 2012 / IA

25 System design requirements
Safety classification based on PSA insights N+2 failure criterion for systems that deal with design basis events; besides redundancy, also adequate diversity and segregation Proven technology properly evaluated operational experience experimental demonstration & analysis (novelties, such as “passive” systems) Adequate demonstration of performance and safety margins on the basis of deterministic studies and PSA October 2012 / IA

26 Design basis for containment
1) Large break LOCA adequate capacity to carry pressure loads and to limit radioactive releases must be shown in conditions expected after a LB LOCA containment this gives a sound basis to manage also severe accidents 2) Severe accidents all foreseeable loads threatening the containment integrity in connection with a severe core damage must be identified, and necessary protection (prevention or mitigation) must be provided against each load in Olkiluoto 3, pressure caused by hydrogen burn is the limiting design event for containment 3) External events potential external events must be identified and protected against October 2012 / IA

27 Features required to mitigate severe accidents
Severe accident management strategy is mandated in containment design high pressure failure of reactor vessel prevented by dedicated depressurization system hydrogen management with autocatalytic recombiners to prevent detonation low pressure melt arrested in a core catcher, with passive long-term cooling containment integrity against dynamic loads containment pressure management in long term containment leak tightness criteria from release limits AC power supply systems and I&C systems dedicated to support severe accident management are required For systems dedicated for protection against severe accidents, single failure criterion applies October 2012 / IA

28 Protection against external threats
After September 11, 2001: political and public will was expressed to improve protection against terrorist actions Reconsideration of aircraft crash design basis consider large passenger and military aircrafts no immediate release of significant amount of radioactive substances initiation and maintenance of key safety functions in spite of the direct consequences of the event (penetration of structures by impacting parts, vibration, explosion, fire) Microwave and biologic weapon consideration October 2012 / IA

29 Structure of the new YVL Guides
A Safety management of a nuclear facility B Plant and system design C Radiation safety of a nuclear facility and environment D Nuclear materials and waste E Structures and equipment of a nuclear facility A.1 Regulatory control of the safe use of nuclear energy A.2 Siting of a nuclear facility A.3 Management systems of a nuclear facility A.4 Organisation and personnel of a nuclear facility A.5 Construction of a NPP A.6 Operation and accident management of a NPP A.7 Risk management of a NPP A.8 Ageing management of a nuclear facility A.9 Reporting on the operation of a nuclear facility A.10 Operating experience feedback of a nuclear facility A.11 Security arrangements of a nuclear facility B.1 Design of the safety systems of a nuclear facility B.2 Classification of systems, structures and equipment of a nuclear facility B.3 Safety assessment a NPP B.4 Nuclear fuel and reactor B.5 Reactor coolant circuit of a NPP B.6 Containment of a NPP B.7 Preparing for the internal and external threats to a nuclear facility B.8 Fire protection of a nuclear facility C.1 Structural radiation safety of a nuclear facility C.2 Radiation protection and dose control of the personnel of a nuclear facility C.3 Control and measuring of radioactive releases to the environmental of a nuclear facility C.4 Radiological control of the environment of a nuclear facility C.5 Emergency preparedness arrangements of a NPP D.1 Regulatory control of nuclear non-proliferation D.2 Transport of nuclear materials and waste D.3 Handling of spent nuclear fuel D.4 Handling of low- and intermediate-level waste and decommissioning of a nuclear facility D.5 Final disposal of nuclear waste E.1 Manufacture and use of nuclear fuel E.2 Construction plan of the mechanical components and structures of a nuclear facility E.3 Regulatory control of the mechanical components and structures of a nuclear facility E.4 Verification of strength of pressure equipment of a nuclear facility E.5 In-service inspections of the mechanical components and structures of a nuclear facility E.6 Buildings and structures of a nuclear facility E.7 Electrical and I&C equipment of a nuclear facility

30 Computer programs for safety analysis in VTT in Finland
ENDF/B, JEF NJOY CASMO libraries Nuclear Data APROS simulation environment RELAP5/Mod3 Thermal hydraulics Plant analyzer Training simulator Thermal hydraulics Assembly wise group constants CASMO 4 SIMULATE rectangular HEXBU-3D hexagonal Severe accidents MELCOR CONTAIN SCDAP/RELAP5 Integral severe accident analyses Containment performance Core and primary system PASULA Pressure vessel integrity in severe accidents Reactivities, power and burnup ENIGMA, FRAPCON, SCANAIR, FRAPTRAN FRAPTRAN-GENFLO Steady state and transient fuel behaviour TRAB-3D BWR HEXTRAN VVER Transient and accident analyses with 3D core model, complete circuit models

31 Reactor analysis calculation system

32 Example on coupling of safety codes used in accident analysis

33 Code development in VTT

34 PSA tool in use in STUK FinPSA is a comprehensive risk and reliability analysis tool intended for full scope PSA/PRA modeling. The tool has been developed and maintained since 1988 by Radiation and Nuclear Safety Authority of Finland (STUK). The tool is designed to support the main activities related to PSA/PRA by easy model creation, efficient and versatile analysis, good traceability, flexible reporting and information exchange capabilities. All these features make your work comfortable in living PSA, plant assessment and operational modifications. From the beginning of 2012, STUK and VTT (Technical Research Centre of Finland) have initiated a project to develop the tool ahead and to train new FinPSA experts. 23 August 2010

35 TSO Support for Olkiluoto 3 Review
Criteria for TSO’s; Competent organisation to carry out EPR analysis Independent from the licensee Codes independent from the licensee’s codes Adequately validated codes for EPR Main organisations used as TSO’s were Technical Research Centre of Finland (VTT) Institute for Safety and Reliability (ISar) Main computer codes used by TSO’s TRAB3D/SMABRE (VTT) APROS (VTT) MELCOR (VTT) ATHLET (ISar)

36 TSO Support for Olkiluoto 3 Review
Main Finnish organisations used were Technical Research Centre of Finland (VTT) and Lappeenranta Technical University (LTU) VTT: Transient analyses with the TRAB3D/SMABRE code Design Basis Accident analyses with the APROS code Independent comparative analyses of primary circuit behaviour during accident situations Small break LOCA Steam Generator Tube Rupture Large break LOCA

37 Tools for Severe Accident Assessment
Integral severe accident analysis tools Used for assessment of progression of whole accident scenario Good for assessment of overall performance of ESF Applicable also for PRA level 2 studies Detailed, separate effects severe accident codes State-of-the-art models Used for in-depth studies of safety critical SA phenomena Generally related to phenomena that cannot be modelled/solved with the methodology used in integral systems analysis codes Experiments Last resort to support or validate analytical results Equipment survivability studies

38 Analytical tools Integral system codes obtained from abroad
Detailed, separate effects SA codes obtained from abroad or developed at VTT Experiments: Participation in major international experimental research programmes Small-scale experiments/testing also in-house

39 Integral Severe Accident Analysis Tools at VTT
MELCOR (version and 2.1) for integral plant analysis, source term, developed in Sandia National Lab’s for USNRC Main severe accident analysis tool since 1990 at VTT Validation: Sandia’s validation and QA PCCS condenser application tested against: Concrete erosion model validated against various OECD/MCCI experiments Plant models for Loviisa 1&2, Olkiluoto 1&2, Olkiluoto 3 ASTEC Integral plant analysis, source term, developed by IRSN (France)/GRS (Germany) Participation in ASTEC users group for validation efforts since 2011 Only RPV lower head melt pool studies performed so far Applications of fission product source term models under way Long-term goal to build full plant inputs for Finnish NPPs

40 Separate Effects Severe Accident Analysis Tools at VTT (1/4)
PASULA code suite for detailed mechanical analyses of structures Creep rupture of lower head Lower head penetration integrity Ex-vessel cooling of lower head Validation against several Sandia’s LHF and OLHF tests Applied for Loviisa 1 & 2, Olkiluoto1 & 2 and SWR-1000 reactors

41 Separate effects Severe Accident Analysis Tools at VTT (2/4)
CORQUENCH for concrete erosion, melt pool coolability, developed at Argonne National Lab for USNRC Code tested at VTT against various CCI-tests and COMET-tests FLUENT (CFD code) for hydrogen mixing and deflagration combustion, commercial UK code Tested at VTT against THAI-experiments for mixing Tested at VTT against FLAME tests for flame speed and acceleration Applied for Olkiluoto 1,2,3 and Hungarian Paks (VVER 440) TONUS for hydrogen detonation, developed at CEA, France Tested at VTT against ENACEFF tests for flame acceleration Tested at VTT against FLAME experiments for detonation DET3D for hydrogen detonation, developed at FZK, Germany Interface with ABAQUS structural analysis code

42 Separate effects Severe Accident Analysis Tools at VTT (3/4)
TEXAS-V for steam explosions, developed at UW, USA Validated at VTT against KROTOS experiments KS-1 and KS-2 Applied for Olkiluoto 1 & 2 MC3D V3.5 for steam explosions, developed at IRSN, France Validation at VTT ongoing against KROTOS and TROI tests Applied for Olkiluoto 1&2 MEWA for corium particle bed coolability, developed at IKE Stuttgart Validation at VTT against own STYX and COOLOCE experiments Applied for Olkiluoto 1 & 2 CONTAIN 2.0 for containment T/H and source term analyses, developed at Sandia National Lab for USNRC Validation at VTT against Fortum’s VICTORIA tests, HDR tests Applied for Loviisa 1 & 2

43 Separate effects Severe Accident Analysis Tools at VTT (4/4)
CHEMPOOL for calculation of pH in containment pools, developed at VTT Validated against a matrix of titration experiments QA’ed for application in safety related work to US Used for supporting design calculation for design of pH control system of OL1/OL2 Applied also for Olkiluoto 3 and ESBWR (GEH) RADTRAD for dose calculations for DBA cases (NRC Regulatory Guide 1.183); developed for USNRC Applied for Ringhals 2 (Westinghouse PWR)

44 Application example: Olkiluoto unit 1 and 2 BWR 860 MW
RPV failure mode Effects of recriticality Hydrogen Core debris coolability Steam explosions pH control Elastomer survivability Painted wall structure decontamination

45 OLKILUOTO 1 and 2 BWR 860 MW Containment inerted with nitrogen  in-containment hydrogen combustion risk low Hydrogen generation  contributes to containment pressurization and H2 leak to RB Recriticality RPV is assumed to fail  debris coolability and stabilization in the Lower Drywell crucial Flooding of Lower Drywell prior to Pressure Vessel failure  particle bed coolability is an issue Ex-vessel steam explosions Containment pool pH conrol

46 RPV lower head failure mode

47 RPV lower head penetrations
MCP penetration Instrument tube penetration Control rod penetration

48 Penetration failure Penetration weld fails by loss of strength in about 40 s after contact with melt Instrument tube inside nozzle tube (”blue tube”) can fall downward and open a flow path Uncertainties: Falling tube becomes stucked at lower elevation Refreezing of melt blockage in discharge channel Control rod penetration would fail 5.5 hours after melt arrival RCP opening would fail 5.2 hours after arrival of melt Liquid corium, T=2550 K Weld RPV wall Instrument tube

49 Creep rupture of lower head
If corium is not able to discharge through a failed penetration Heat fluxes from the melt pool to the RPV wall obtained from MELCOR calculation Creep rupture occurs near the interface of metallic and oxidic melt layer at 5.5 hours after melt arrival Practically no difference in timing between RPV wall creep rupture, control rod tube failure and RCP opening failure But, the initial hole areas may differ

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