9th Workshop on European Collaboration for
Higher Education and Research in Nuclear Engineering &
Radiological ProtectionSalamanca, Spain
5-7 June 2013
Probability Risk Assessment course for Probability Risk Assessment course for CHERNE studentsCHERNE students
Sebastián Martorell, José RódenasDepartamento de Ingeniería Química y Nuclear
Universidad Politécnica de Valencia (Spain)
CHERNE-9CHERNE-9
1.1. Introduction to fundamentals and procedures to develop a PRAIntroduction to fundamentals and procedures to develop a PRAIntroduction to LWR technology (Elements, PWR, BWR) Introduction to LWR technology (Elements, PWR, BWR) Overview of the PRAOverview of the PRAAccident identificationAccident identificationAccident sequence modelingAccident sequence modelingData assessmentData assessmentAccident sequence quantificationAccident sequence quantification
2.2. Practical application to a PWR Nuclear Power PlantPractical application to a PWR Nuclear Power PlantLarge Break Loss of Coolant Accident (LBLOCA) – Level 1 PRALarge Break Loss of Coolant Accident (LBLOCA) – Level 1 PRAUse of Software toolsUse of Software tools
Contents of the courseContents of the course
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1.1. Development of the courseDevelopment of the course-Introduction to fundamentals and procedures - Introduction to fundamentals and procedures - LecturesLectures-Practical application – Practical application – Computer roomComputer room
2.2. Application toolsApplication tools-Level 1 PRA - LBLOCA documentation Level 1 PRA - LBLOCA documentation
-FaultTree+ , IAEA Databank,…FaultTree+ , IAEA Databank,…
3.3. Evaluation of the courseEvaluation of the course-Case Study (portfolio) Case Study (portfolio)
4.4. ReferenceReference
Procedures for conducting PSA of NPP (Level 1). Procedures for conducting PSA of NPP (Level 1).
Safety Series Nº 50-P-4. International Atomic Energy Agency. Safety Series Nº 50-P-4. International Atomic Energy Agency. IAEAIAEA Vienna. 1992. Vienna. 1992.
Logistics of the courseLogistics of the course
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MondayMonday 27 27 Tuesday 28Tuesday 28 Wednesday 29Wednesday 29 Thursday 30Thursday 30 Friday 31Friday 31
10-1110-11 IntroductionIntroduction Accident Accident
SequenceSequence
ModelingModelingData Data
AssessmentAssessment
Accident Accident Sequence Sequence QuantificationQuantification EvaluationEvaluation11-1211-12 PRAPRA
OverviewOverview12-1312-13
13-1513-15 Lunch breakLunch break
15-1615-16 Accident Accident
Sequence Sequence IdentificationIdentification
Accident Accident
SequenceSequence
ModelingModelingData Data
AssessmentAssessment
Accident Accident Sequence Sequence QuantificationQuantification FreeFree16-1716-17
17-1817-18
Timetable (tentative)Timetable (tentative)
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More informationMore information
Venue: Department of Nuclear Engineering at the Venue: Department of Nuclear Engineering at the Universitat Politècnica de València (SPAIN)Universitat Politècnica de València (SPAIN)
Tentative Dates: 27-31 January 2014Tentative Dates: 27-31 January 2014 Pre-registration deadline: 30 October 2013 Pre-registration deadline: 30 October 2013 2 ECTS can be obtained after positive evaluation2 ECTS can be obtained after positive evaluation Additional data of interest:Additional data of interest:
Fees: 100 € (maximum)Fees: 100 € (maximum)Minimum number of students: 10Minimum number of students: 10Maximum number of students: 20Maximum number of students: 20Selection at home institutionsSelection at home institutions
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1.1. Reactor Safety Study (WASH-1400) USNRC, 1975Reactor Safety Study (WASH-1400) USNRC, 1975 First comprehensive application of the methods and First comprehensive application of the methods and
techniquestechniques
2.2. Has become standard tool in safety evaluation of Nuclear Has become standard tool in safety evaluation of Nuclear Power Plants. Provide insights into:Power Plants. Provide insights into:
Dominant risk contributorsDominant risk contributors Plant designPlant design Limiting Condition for Operation (LCO, eg. ST and AOT) …Limiting Condition for Operation (LCO, eg. ST and AOT) …
3.3. Helps Risk Informed Decision MakingHelps Risk Informed Decision Making Methodological approach to accident sequences identification, Methodological approach to accident sequences identification,
modeling and quantificationmodeling and quantification Provides numerical estimates of risks, frequency and damageProvides numerical estimates of risks, frequency and damage
PRA backgroundPRA background
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In international practice three levels of PRA evolvedIn international practice three levels of PRA evolved Level 1:Level 1:
The assessment of plant failures leading to the The assessment of plant failures leading to the determination of core damage frequencydetermination of core damage frequency
Level 2:Level 2:
The assessment of containment response leading, The assessment of containment response leading, together with Level 1 results, to the determination of together with Level 1 results, to the determination of containment release frequenciescontainment release frequencies
Level 3:Level 3:
The assessment of off-site consequences leading, The assessment of off-site consequences leading, together with the results of Level 2 analysis, to together with the results of Level 2 analysis, to estimate public risksestimate public risks
PRA levelsPRA levels
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Management and
organization
1
Identification
of hazards& accidentinitiators
Accidentsequencemodeling
3
Data assessme
nt
4
Accidentsequence
quantification
5
Reportof the
analysis
6
Procedures for conducting PSA of NPP (Level 1).
Safety Series Nº 50-P-4.
International Atomic Energy Agency. Vienna. 1992.
2
PRA steps and tasksPRA steps and tasks
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Management and
organization
1
Identification
of hazards& accidentinitiators
2
Accidentsequencemodeling
3
Data assessme
nt
4
Accidentsequence
quantification
5
Reportof the
analysis
6
Familiarization with the plantIdentification of hazards (sources of radioactive release)Selection of plant operational statesDefinition of consequences and damage statesIdentification of accident initiatorsDetermination of safety functions and plant systems
Accident identificationAccident identification
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Management and
organization
1
Identification
of hazards& accidentinitiators
2
Accidentsequencemodeling
3
Data assessme
nt
4
Accidentsequence
quantification
5
Reportof the
analysis
6
Event sequence modeling (Event Tree Analysis)System modeling (Fault Tree Analysis)Human Performance AnalysisQualitative dependences analysisClassification of accident sequences into plant damage states
Accident modellingAccident modelling
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Management and
organization
1
Identification
of hazards& accidentinitiators
2
Accidentsequencemodeling
3
Data assessme
nt
4
Accidentsequence
quantification
5
Reportof the
analysis
6
Assessment of the frequency of initiating eventsAssessment of RAM of components Reliability, maintainability and availability models Data basesAssessment of human error probabilities
Accident data assessmentAccident data assessment
CHERNE-9CHERNE-9
Management and
organization
1
Identification
of hazards& accidentinitiators
2
Accidentsequencemodeling
3
Data assessme
nt
4
Accidentsequence
quantification
5
Reportof the
analysis
6
Qualitative analysis (Boolean equations)Quantitative analysis of frequencies of accident sequencesImportance and sensitivity analysisUncertainty analysis
Accident quantificationAccident quantification
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ExamplesExamples
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Thank you for your attentionThank you for your attention
