Source Term Determination Methods of the Slovenian Nuclear Safety Administration Emergency Response Team

IAEA TM on Source Term Evaluation for Severe Accidents, Vienna, 21-23 October 2013 Source Term Determination Methods of the Slovenian Nuclear Safety Administration Emergency Response Team Tomaž Nemec Slovenian Nuclear Safety Administration Litostrojska 54, Ljubljana, SLOVENIA tomaz.nemec@gov.si

Nuclear facilities in Slovenia Nuclear Power Plant Krško, Westinghouse 2 loop PWR, 676 MWe, with spent fuel pool storage (~1100 spent FA) On-site emergency plan for protection and rescue Emergency centers onsite (TSC, OSC) and offsite (EOF) Annual emergency response exercises with NPP full scope simulator with modelling of severe accidents Research Reactor TRIGA Mark II of the Jožef Stefan Institute, GA, 250 kwt, in Brinje near Ljubljana Only local radiological consequences possible Central interim storage of radioactive waste in Brinje - Storage of radioactive waste produced in industry, research and medicine near Ljubljana (next to the TRIGA RR) Limited off-site consequences in case of fire 2

Location of nuclear facilities in Slovenia 3

EMERGENCY RESPONSE IN SLOVENIA - ROLE OF THE SNSA 4

Emergency Response organisations in Slovenia National Emergency Response Plan for Nuclear and Radiological Accidents (2010) ACPDR - Administration for Civil Protection and Disaster Relief responsible for population protection and for the organization of civil protection units in nuclear installations Coordination of civil protection activities SNSA - Slovenian Nuclear Safety Administration responsible for on-site procedures and measures related to the onsite emergency plan 1. Gathering information on NPP conditions : ERDS & KSID 2. Assessment of Krško NPP status in case of emergency events 3. Source term evaluation in severe accidents for NPP & spent fuel 4. Recommendations for protective measures 5. International reporting in case of emergencies 5

SNSA emergency response organisation SNSA inspector on duty 24/7 At SNSA emergency response center: Emergency director Nuclear accident analysis group (source term evaluation) Dose assessment group (radioactive releases to environment, evaluation of doses to population) Group of communicators (EMERCON, ECURIE, public) Technical support SNSA representatives in: Krško NPP EOF (in Ljubljana) Headquarters for Civil Protection (HCP) 6

Nuclear accident analysis group (SSAJN) review of emergency classification determined by NPP operators assessment of plant conditions determination of source term based on the status of radiological barriers conservative prognosis of scenario evaluation in near future preparation of input (calculated source term) to the SSOD for RODOS, DOZE, INTERRAS models 7

Dose assessment group (SSOD) Tools & models: INTERRAS (RASCAL 3.0.3) GAUSS model RODOS DOZE RIMPUFF model Enables prognosis of weather development Rich set of results based on database on NPP characteristics used only for emergencies in Krško NPP plan specific Lagrange model with site local site characteristics data Diagnostic Input data on source term determined by the SSAJN 8

SNSA Emergency director and communicators Gathering information from NPP and national stakeholders MKSID communication tool (2008) International reporting to neighbouring countries (bilateral agreements) IAEA (EMERCON) EU (ECURIE) Public information for Slovenia Answering questions from the public and requests for information Preparing recommendations for protective measures for population and advising to the HCP and the government 9

NUCLEAR ACCIDENT ANALYSIS 10

Nuclear accident analysis group (SSAJN) Review of emergency classification by the operator Continuous assessment of: barriers: cladding, primary circuit boundary, containment plant conditions and scenario in progress determination of radioactive release path to environment any probable worst case development based on status of critical safety functions, safety systems availability, challenge to barriers Actions of plant operators and emergency response teams Calculation of source term: Leakage of primary coolant Core damage assessment Status of containment or bypass of contaiment Decrease of radioisotopes concentration in containment Integration of radioactivity that was released to the environment 11

PWR / Krško NPP overview Figure adapted from US NRC Response Technical Manual, 1993 12

Krško NPP release paths monitoring (ERDS) Releases from reactor coolant system: LOCA leak/break detection: RM-2, RM-7, RM-11, RM-12, RM-22 LOCA with degraded core conditions: RM-9 in RM-10 SGTR: RM-19, RM-23, RM-31, RM-32 Intersystem LOCA: RM-4 Leak to the CC (RM-17) and SW (RM-20) systems NEW release through filtered containment venting system in SA (RM to be installed in the future) Other releases: Release path through relief/safety valves on main steam line: RM-33 in RM-34 Release path through plant vent: RM-14, RM-14, RM-21, RM-27; accidental RM-24.1 in RM-24.2 Releases from spent fuel storage: RM-5; accidental RM-5.2, RM-28 Release path from condenser RM-15, accidental RM-25.1, RM-25.2 13

ERDS Emergency Response Data System set of parameters based on NUREG 1394 and expanded with some Krško NPP parameters 192 plant/sim parameters online: Reactor coolant pressure Reactor coolant temperature Core exit thermocouples (39) Subcooling margin Pressurizer level Reactor coolant system charging and letdown Reactor coolant system flow Reactor power Steam generators level Steam generators pressure Main feed water flow Auxiliary feed water flow Reactor vessel level (RVLIS) Hydrogen concentration in cont. ECCS flow (SI, RHR) RWST level Containment pressure Containment temperature Containment sump level Containment radiation Condenser radiation Plant vent radiation Process radiation Vent/exhaust flow Wind, atmospheric stability Simulated weather ERDS will be expanded with additional parameters and trending capabilities 14

Krško NPP reporting by fax or MKSID Additional information Notification on emergency Time of reactor shutdown time of emergency classification Emergency class determination criterion, EAL Radioactivity release to environment Status of safety systems Status of critical safety functions Operational procedures in course Operability of emergency centers (TSC, EOF) Plant conditions and actions in course External support to the plant Content and location of radioactive releases to environment 15

CALCULATION OF SOURCE TERM - IN AN EMERGENCY 16

Input parameters Pre-selected data: Activity of primary coolant Inventory of isotopes in reactor core Reactor power MWe (normalisation) Variables status of barriers (core, RCS, containment) Primary coolant leakage rate (RCS barrier conditions) Cladding failure (%) (core conditions) Core melt (%) (core conditions) Reactor pressure vessel melt-through Reduction factor for containment atmosphere radioactivity Containment leakage / bypass (containment conditions) 17

Excel table for source term calculation 38 isotopes with predetermined concentrations/inventories 18

Reactor coolant activity / Core inventory Generic data from IAEA TecDoc 955 for PWR Reactor coolant concentrations Core fission product inventory (FPI) EOL 18-months cycle, 676 MWe PWR (=Krško NPP) In the Krško NPP reactor core fission product inventory is calculated in real time using measured reactor power data In 2004 at cycle extension 12->18 months the core fission product inventory was checked by calculations with ORIGEN and compared with the generic data (IAEA TecDoc 955) 19

Equations for calculation of typical releases Release of normal reactor coolant H ni =M A si F Release at core damage H ci = A ri F Release at SG tube rupture (containment bypass) H ui = A si M f M leakage rate A activities of reactor coolant / reactor core F RDF, release reduction factor f activities transfer to secondary circuit at SG tube rupture 20

Source term calculation Variables used in source term calculation (IAEA TecDoc 955) Core release fractions (CRF): normal coolant leakage, cladding failure (gap release), core melt Release reduction factor (RDF) for particulates/aerosols Escape fractions (EF): containment leakage, SG tube rupture 21

Core damage assessment IAEA TecDOC 955 Several methods (limits adapted to Krško NPP): Core temperature (Core exit thermocouples; CET) Time of core uncovery (Reactor vessel level; RVLIS) Containment radiation monitors readings (PARMS) Primary coolant activities (PASS) Hydrogen concentration inside containment Results calculated by these methods need to be intercompared and confirmed 22

Core damage assessment Krško NPP method Several methods (Krško NPP procedure based on WOG Core Damage Assessment Guidance, 1999): Core temperature (Core exit thermocouples; CET) Time of core uncovery (Reactor vessel level; RVLIS) Containment radiation monitors readings (PARMS) Primary coolant activities (PASS) Hydrogen concentration inside containment Results calculated by these methods need to be intercompared and confirmed SNSA performs calculation using generic methods and uses WOG methods for an independent verification of its own assessment results 23

Krško NPP procedure Core damage assessment Some examples: 24

TYPICAL RESULTS OF DBA & SA SOURCE TERM CALCULATIONS 25

Typical source term calculations for Krško NPP SG tube rupture 3 tubes ruptured; f=1.05 Mass of Krško NPP reactor coolant ~ 200 tons Release through main steam line relief valve to environment Release = 15 GBq/s Core uncovery cladding failure 100% failure of fuel cladding Reduction of activities; F=0.36 Containment integrity design leakage rate; EF=0.2% per day Release through containment anulus and plant vent is reduced by HEPA filters Release = 22 GBq/s 26

Typical Krško source term calculations - 2 Core uncovery core melt 100% melt of fuel pellets Reduction of activities; F=0.36 Containment integrity design leakage rate; EF=0.2% per day Release through containment anulus and plant vent is reduced by HEPA filters Release = 350 GBq/s Reactor pressure vessel melt Molten core release into containment Reduction of activities; F=0.03 Containment integrity hydrogen explosion caused a break in containment Release through containment directly into environment; EF=100% per hour Release = 3.5 1E+6 GBq/s 27

Source term for filtered containment venting NEW - Passive containment filtered vent system (PCFVS) instalation in course during the outage - October 2013 Westinghouse dry filter method (DFM) Passive opening of filtered venting at containment pressure 6 bar abs (rupture disk) first venting occurs before MCCI Aerosol filters in containment, iodine filter in aux. building Decontamination factors: noble gases 1, aerosols 10000, elemental iodine 100, organic iodine 10 Filtered release estimate = 3.4 1E+6 GBq/s noble gases, 2500 GBq/s iodine, 100 GBq/s aerosols Non-filtered release = 3.4 1E+6 GBq/s noble gases, 2.3 1E+6 GBq/s iodine, 1.2 1E+6 GBq/s aerosols 28

OTHER SOURCE TERM RESULTS 29

Source term calculations for Krško NPP Design basis events: Krško NPP Safety Analysis Report (USAR) Assumption - 1% defective fuel cladding Core and gap activities at EOL 18-months cycle For a list of DBA, including FA damage in spent fuel pool PSA IPE level 2 for NPP Krško MAAP - analytical tool for PSA level 2 study Internal events and external events (seismic, fire, flood etc) 8 (+4) release categories, 12 fission product groups Results used also in ASTRID Pre-calculated typical source terms for RODOS PSA level 2 source term can be used as input for DOZE 30

Future work Krško NPP will calculate source term for severe accident including MCCI - filtered containment venting release Krško NPP will model release and calculate doses to environment/population for the same SA scenario Emergency protection areas will be re-evaluated New ERDS tool with extended data set will be developed SNSA is following international activities with new tools for quick determination of source term SNSA is following development of generic assessment methods for source term determination (revision of the IAEA TecDoc 955) 31

REFERENCES 1. IAEA TECDOC-955»Generic Assessment Procedures for Determining Protective Actions during a Reactor Accident«, IAEA, 1997 2.»International Response Technical Manual (RTM) for Interim Use and Comment«(RTM-95), Volumes 1-3, U.S. NRC, 1995 3. Krško NPP procedure EIP-17.001»Emergency class determination«4. Krško NPP procedure EIP-17.070»Core Damage Assessment«5. WOG Core Damage Assessment Guidance, WCAP-14696-A, 1999 6. Krško NPP procedures EOP»Emergency Operating Procedures«7. Krško NPP guidelines SAG-17.001»Severe Accident Management Guidelines (SAMG)«8. SSR-NEK-7.10.3»Source term calculations«, Westinghouse, 2001 9. Krško NPP Safety analysis report 10. Determination of source term for Krško NPP extended fuel cycle, T. Nemec et al., Proc. NENE 2004 11. Slovenian Post-Fukushima Action Plan 32