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29 - 31 March 20041 “Experience Gained from the Mexican Nuclear Regulatory Authority in the Probabilistic Safety Assessment Level 2 Development for Laguna.

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Presentation on theme: "29 - 31 March 20041 “Experience Gained from the Mexican Nuclear Regulatory Authority in the Probabilistic Safety Assessment Level 2 Development for Laguna."— Presentation transcript:

1 29 - 31 March 20041 “Experience Gained from the Mexican Nuclear Regulatory Authority in the Probabilistic Safety Assessment Level 2 Development for Laguna Verde NPP” Cologne, Germany INTERNATIONAL WORKSHOP ON LEVEL 2 PSA AND SEVERE ACCIDENT MANAGEMENT OECD/NEA

2 29 - 31 March 20042 PSA for Laguna Verde NPP Individual Plant Examination Individual Plant Examination Mexican Regulatory Authority developed an independent PSA level 2 model. Mexican Regulatory Authority developed an independent PSA level 2 model. Review Process Review Process

3 29 - 31 March 20043 Level 2 PSA Methodologies Utility used the methodology of small event tree-large fault tree developed by the Electrical Power Research Institute (EPRI) of the USA Utility used the methodology of small event tree-large fault tree developed by the Electrical Power Research Institute (EPRI) of the USA CNSNS used: CNSNS used:  NUREG-1150 (Accident Progression Event Tree, APET)

4 29 - 31 March 20044 Laguna Verde Plant / Containment Characteristics Two BWR/5, 2027 MWt Two BWR/5, 2027 MWt  Unit 1 began its commercial operation in 1990 and Unit 2 in 1995. Fuel mass: 92 041 kg Fuel mass: 92 041 kg Zirconium mass: 34908 kg Zirconium mass: 34908 kg MARK II Containment MARK II Containment

5 29 - 31 March 20045 Laguna Verde Nuclear Power Plant Mark II Containment

6 29 - 31 March 20046 Laguna Verde. Plant / Containment Characteristics Containment volume 10907 m 3 Containment volume 10907 m 3 Power / Containment volume: 0.18 MWt/m 3 Power / Containment volume: 0.18 MWt/m 3 Power / Suppression pool volume: 0.63 MWt/m 3 Power / Suppression pool volume: 0.63 MWt/m 3 Several systems to supply coolant injection to the core. Several systems to supply coolant injection to the core. Depressurization of the vessel Depressurization of the vessel

7 29 - 31 March 20047 Laguna Verde. Plant / Containment Characteristics Internal design pressure 3.16 kg/cm 2 Internal design pressure 3.16 kg/cm 2 Estimated containment failure pressure: Estimated containment failure pressure: 11.6 kg/cm 2 Venting pressure 4.2 kg/cm 2 Venting pressure 4.2 kg/cm 2 3 Diesel Generators 3 Diesel Generators

8 29 - 31 March 20048 Regulatory Authority PSA level 1 PSA LEVEL 1 PSA LEVEL 1  CDF: 5.65E-5 reactor / year  Station Black Out (43%) LEVEL 1 / 2 INTERFACE LEVEL 1 / 2 INTERFACE  25 Plant Damage States, which resume the possible plant states at the moment of core damage.

9 29 - 31 March 20049 CNSNS, PSA level 2 Methodology Accident Progression Event Tree Accident Progression Event Tree  131 Questions about possible events  Phenomenological aspects  Systems availability  Operator interactions  Conditions before core damage  Containment conditions before and after vessel breach  Containment failure modes MELCOR code was used to support the APET. MELCOR code was used to support the APET.

10 29 - 31 March 200410 Accident Severe Phenomenology for LV NPP In-vessel In-vessel  Generation of hydrogen  Melt progression Ex – vessel Ex – vessel  Direct Containment Heating  Steam Explosions  Core – Concrete Interaction  Fission Products Transport  Mitigation by suppression pool scrubbing

11 29 - 31 March 200411 Structural Containment Analysis Containment Failure Modes Leak Leak  Depressurization of the containment after 2 hrs. (A=92.9 cm 2 ) Rupture Rupture  Depressurization of the containment before or at 2 hrs. (A=929 cm 2 ) Temperature ( o K) Pressure (Kg/cm 2 ) 30311.6 40311.1 48911.1 5736.8 Ultimate Capacity of the Containment

12 29 - 31 March 200412 Accident Progression Event Tree The quantification process was performed by means of the EVNTRE computer code The quantification process was performed by means of the EVNTRE computer code  Event Progression Analysis Code, NUREG/CR-5174, Sandia National Laboratories. More than 1000 accident progression paths More than 1000 accident progression paths

13 29 - 31 March 200413 Accident Progression Event Tree Binning/Rebinning Binning/Rebinning  33 initial characteristics of accident progression paths.  13 final characteristics (bins).

14 29 - 31 March 200414 Source Term Analysis  LVSOR (series XSOR)  Parametric equation based on mass conservation PhenomenonPhenomenon Events related with the accident progressionEvents related with the accident progression

15 29 - 31 March 200415 Releases Categories Timing Timing  Early (before 6 hours)  Intermediate (6 to 24 hours)  Late (after 24 hours) A mount of fission product releases A mount of fission product releases  High (more than 10% of Cs-I)  Moderate (1% to 10%)  Low (less than 1%)

16 29 - 31 March 200416 Individual Plant Examination 9 PDS 9 PDS Containment Event Tree developed for the Accident Progression Analysis. Containment Event Tree developed for the Accident Progression Analysis. 50 to 400 accident progression paths for every Containment Event Tree 50 to 400 accident progression paths for every Containment Event Tree Codes: CAFTA and MAAP Codes: CAFTA and MAAP

17 29 - 31 March 200417 Containment Failure Mode for LVNPP (CNSNS study)

18 29 - 31 March 200418 LVNPP PSA Level 2 Results CNSNSIPE Containment failure frequency 5.25E-52.59E-5 Conditional probability of vessel breach at high pressure given core damage 0.140.27 Conditional probability of containment failure by leak or rupture 0.550.6 Conditional probability of containment venting 0.370.06 Conditional probability of no containment failure 0.070.19 Conditional probability of containment bypass --0.14

19 29 - 31 March 200419 LVNPP PSA Level 2 Results CNSNSIPE Conditional probability of containment failure before vessel breach 0.360.15 Conditional probability of containment failure at vessel breach 0.14--- Conditional probability of containment failure after vessel breach 0.420.66 Conditional probability of no containment failure 0.070.19

20 29 - 31 March 200420 LVNPP PSA Level 2 Results CNSNSIPE Large Early Release Frequency 1.02E-083.4E-7 Conditional probability of Cs and I release greater or equal to 10%: 0.740.36 Conditional probability of Cs and I release between 1% and 10%: 0.050.13 Conditional probability of Cs and I release lower than 1%: 0.130.32

21 29 - 31 March 200421 LVNPP PSA Level 2 Results CNSNSIPE Conditional probability of releases in an early period 0.070.05 Conditional probability of releases in an intermediate period 0.70.67 Conditional probability of releases in a late period 0.170.09 Conditional probability of no containment failure 0.070.19

22 29 - 31 March 200422 Categories of fission products releases IPE HE 1% HI 49% LL 17% NL 16% HL 7% LI 8% LE 1% MI 0% ME 1% ML 0% CNSNS HE 0% HI 59% HL 16% ME 3% MI 2% ML 0% LE 4% LI 9% LL 0% NL 7%

23 29 - 31 March 200423 Experience Gained During the Review Process The experience gained during the development of the regulatory PSA allow us to focus the review process on those important features of the back end analysis. k Timing of phenomenological issues k Parameter figures k The usage of simulations code results k Containment structural analysis k Source term analysis

24 29 - 31 March 200424 Review Process Modifications and improvements: Modifications and improvements:  CET structure  Fault tree models  Input deck of MAAP code  Conatainment filure modes inclution

25 29 - 31 March 200425 CONCLUSIONS Largest source terms in both studies are associated to Station Blackout scenarios Largest source terms in both studies are associated to Station Blackout scenarios The overpressure is the dominant containment failure mode, and the location is in the drywell. The overpressure is the dominant containment failure mode, and the location is in the drywell. In general both studies show similar trends in the evolution of the accident progression and source term released. In general both studies show similar trends in the evolution of the accident progression and source term released. The small containment event tree method is more traceable, and considerably easier to review The small containment event tree method is more traceable, and considerably easier to review IPE did not characterize uncertainty. IPE did not characterize uncertainty.

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