Regulatory Decision Making and PRA

George Apostolakis
Head, Nuclear Risk Research Center
apostola@mit.edu
2nd JEA Nuclear Standards Committee Symposium
TokyoJune 4, 2015
2015

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The Traditional Approach
Prior to Risk Assessment• Management of (unquantified at the time)
uncertainty was always a concern.
• Defense-in-depth and safety margins became
embedded in the regulations.
• “Defense-in-Depth is an element of the NRC’s safetyphilosophy that employs successive compensatory
measures to prevent accidents or mitigate damage if
a malfunction, accident, or naturally caused event
occurs at a nuclear facility.” [USNRC White Paper,February, 1999]

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Problems with the Traditional Approach• There is no guidance as to how much defense indepth is sufficient• DBAs use qualitative approaches for ensuringsystem reliability (the single-failure criterion) whenmore modern quantitative approaches exist• DBAs use stylized considerations of humanperformance (e.g., operators are assumed to takeno action within, for example, 30 minutes of an
accident’s initiation)• DBAs do not reflect operating experience andmodern understanding• Industry-sponsored PRAs showed a variability inrisk of plants that were licensed under the same
regulations.
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Probabilistic Risk Assessment•Study the system as an integrated socio-technicalsystem•

Probabilistic Risk Assessment (PRA) supports
Risk Management by answering the questions:

What can go wrong? (thousands of accident sequencesor scenarios)How likely are these scenarios?

What are their consequences?

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PRA Model Overview and Subsidiary
Objectives
CDF
10-4/ry

LERF
10-5/ry

QHOs

Level I

Level II

Level III

PLANT
MODELCONTAINMENT
MODEL

SITE/CONSEQUENCE
MODEL

Results
Accident
sequencesleading toplantdamagestates
Uncertainties

Results

Results

Containment
failure/release
sequences

Public healtheffects

PLANT MODE
At-power Operation
Shutdown / Transition
Evolutions
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SCOPE
Internal Events
External Events

Risk-Informed Changes to the
Licensing Basis (RG 1.174; 1998)
Comply with
Regulations

Maintain
Defense-inDepth
Philosophy

Maintain
SafetyMargins

Integrated
Decision Making
Risk Decrease,
Neutral, or Small
Increase

Monitor
Performance6CDF

Acceptance Guidelines for Core Damage
Frequency
Region I- No changes
Region IIRegion I- Small Changes- Track Cumulative Impacts
Region III

10-5- Very Small Changes- More flexibility with respect toBaseline- Track Cumulative Impacts

Region II
10-6Region III

10-5

10-4
7CDF

Decision Making• Regulatory decision making (like any decision) should
be based on the current state of knowledge and should
be documented
 The current state of knowledge regarding design, operation,and regulation is key.
 PRAs do not “predict” the future; they evaluate and assess
future possibilities to inform the decision makers’ current
state of knowledge.
 Ignoring the results and insights from PRAs results indecisions not utilizing the complete state of knowledge.
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PRA Adequacy• A full-scope PRA includes all operating modes,
internal and external initiating events and estimatesof the core damage frequency, large early releasefrequency, release categories, and health effects• Most regulatory decisions utilize Level 1 PRAs andLarge Early Release Frequency• Many regulatory decisions do not require a fullscope Level 1 PRA• The level of detail of the PRA is determined by itsintended use
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9PRA Quality• PRA models are ambitious in scope (socio-technicalsystem model)• Many diverse models are employed (systems,human reliability, earthquakes, etc)
• Expert judgment is important, just as it is importantin “deterministic” analyses• Regulators must have confidence that the quality ofrisk information is sufficient to justify its use indecision making
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Regulatory Guidance
USNRC Regulatory Guide 1.200, “AN APPROACH FOR
DETERMINING THE TECHNICAL ADEQUACY OF PROBABILISTICRISK ASSESSMENT RESULTS FOR RISK-INFORMED ACTIVITIES”
 Scope of a PRA
 Technical elements of a full-scope Level 1 and Level 2
PRA and their associated attributes and characteristics
 Level of detail of a PRA
 Development, maintenance, and upgrade of a PRA• The documentation must be sufficient to facilitate
independent peer reviews

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Peer Review• Qualifications of the experts
 independent with no conflicts of interest (i.e., have not
performed any work on the PRA)
 collectively represent expertise in all the technicalelements of a PRA including integration
 expertise in the technical element assigned to review
 knowledge of the plant design and operation
 knowledge of the peer review process• Guidance for reviews
 NEI 00-02, “Probabilistic Risk Assessment Peer ReviewProcess Guidance.”
 NEI 05-04, “Process for Performing Follow-On PRA Peer
Reviews Using the ASME PRA Standard.”
 NEI 07-12, “Fire Probabilistic Risk Assessment (FPRA)
Peer Review Process Guidelines.”
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PRA Standards• “The peer review is to be performed against
established standards” (RG 1.200)• Examples
 ASME/ANS RA-Sa-2009, “Addenda to ASME/ANS RA-S2008 Standard for Level 1/Large Early Release Frequency
Probabilistic Risk Assessment for Nuclear Power PlantApplications,”
 ASME/ANS RA-S-1.4-2013: Probabilistic RiskAssessment Standard for Advanced Non-LWR Nuclear
Power Plants (for trial use)

• Concern about stifling methodological progress
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ASME/ANS RA-Sa-2009 Requirements• High-Level Requirements
 The HLRs set forth the minimum requirements for atechnically acceptable baseline PRA, independent of anapplication.

• Supporting Requirements
 For each Capability Category, the SRs define theminimum requirements necessary to meet that CapabilityCategory• Capability Category II
 Resolution and specificity sufficient to identify therelative importance of the significant contributors at thecomponent level Use of plant-specific data/models

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Example: Initiating Events• HLR-IE-A The initiating event analysis shall provide a
reasonably complete identification of initiating
events.
• Supporting Requirements
 IE-A2 INCLUDE in the spectrum of internal-event
challenges considered at least the following generalcategories: transients, LOCAs, SGTR, ISLOCAs, supportsystem failures
 IE-A4 REVIEW generic analyses and operating
experience of similar plants to assess whether the list ofchallenges included in the model accounts for industry
operating experience
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