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9 th Workshop

9 th Workshop on European Collaboration for Higher Education and Research in Nuclear Engineering & Radiological Protection Salamanca, Spain 5-7 June 2013. Probability Risk Assessment course for CHERNE students Sebastián Martorell, José Ródenas Departamento de Ingeniería Química y Nuclear

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9 th Workshop

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  1. 9th Workshop on European Collaboration for Higher Education and Research in Nuclear Engineering & Radiological Protection Salamanca, Spain 5-7 June 2013 Probability Risk Assessment course for CHERNE students Sebastián Martorell, José Ródenas Departamento de Ingeniería Química y Nuclear Universidad Politécnica de Valencia (Spain)

  2. Contents of the course • Introduction to fundamentals and procedures to develop a PRA • Introduction to LWR technology (Elements, PWR, BWR) • Overview of the PRA • Accident identification • Accident sequence modeling • Data assessment • Accident sequence quantification • Practical application to a PWR Nuclear Power Plant • Large Break Loss of Coolant Accident (LBLOCA) – Level 1 PRA • Use of Software tools

  3. Logistics of the course • Development of the course • Introduction to fundamentals and procedures - Lectures • Practical application – Computer room • Application tools • Level 1 PRA - LBLOCA documentation • FaultTree+ , IAEA Databank,… • Evaluation of the course • Case Study (portfolio) • Reference Procedures for conducting PSA of NPP (Level 1). Safety Series Nº 50-P-4. International Atomic Energy Agency. IAEA Vienna. 1992.

  4. Timetable (tentative)

  5. More information • Venue: Department of Nuclear Engineering at the Universitat Politècnica de València (SPAIN) • Tentative Dates: 27-31 January 2014 • Pre-registration deadline: 30 October 2013 • 2 ECTS can be obtained after positive evaluation • Additional data of interest: • Fees: 100 € (maximum) • Minimum number of students: 10 • Maximum number of students: 20 • Selection at home institutions

  6. PRA background • Reactor Safety Study (WASH-1400) USNRC, 1975 • First comprehensive application of the methods and techniques • Has become standard tool in safety evaluation of Nuclear Power Plants. Provide insights into: • Dominant risk contributors • Plant design • Limiting Condition for Operation (LCO, eg. ST and AOT) … • Helps Risk Informed Decision Making • Methodological approach to accident sequences identification, modeling and quantification • Provides numerical estimates of risks, frequency and damage

  7. PRA levels In international practice three levels of PRA evolved • Level 1: The assessment of plant failures leading to the determination of core damage frequency • Level 2: The assessment of containment response leading, together with Level 1 results, to the determination of containment release frequencies • Level 3: The assessment of off-site consequences leading, together with the results of Level 2 analysis, to estimate public risks

  8. PRA steps and tasks 6 3 4 5 1 2 Data assessment Accident sequence modeling Accident sequence quantification Report of the analysis Identification of hazards & accident initiators Management and organization Procedures for conducting PSA of NPP (Level 1). Safety Series Nº 50-P-4. International Atomic Energy Agency. Vienna. 1992.

  9. Accident identification 6 3 4 5 1 2 Data assessment Accident sequence modeling Accident sequence quantification Report of the analysis Identification of hazards & accident initiators Management and organization Familiarization with the plant Identification of hazards (sources of radioactive release) Selection of plant operational states Definition of consequences and damage states Identification of accident initiators Determination of safety functions and plant systems

  10. Accident modelling 6 3 4 5 1 2 Data assessment Accident sequence modeling Accident sequence quantification Report of the analysis Identification of hazards & accident initiators Management and organization Event sequence modeling (Event Tree Analysis) System modeling (Fault Tree Analysis) Human Performance Analysis Qualitative dependences analysis Classification of accident sequences into plant damage states

  11. Accident data assessment 6 3 4 5 1 2 Data assessment Accident sequence modeling Accident sequence quantification Report of the analysis Identification of hazards & accident initiators Management and organization Assessment of the frequency of initiating events Assessment of RAM of components Reliability, maintainability and availability models Data bases Assessment of human error probabilities

  12. Accident quantification 6 3 4 5 1 2 Data assessment Accident sequence modeling Accident sequence quantification Report of the analysis Identification of hazards & accident initiators Management and organization Qualitative analysis (Boolean equations) Quantitative analysis of frequencies of accident sequences Importance and sensitivity analysis Uncertainty analysis

  13. Examples

  14. Thank you for your attention

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