Reactor Oversight Process
Mark A. Cunningham
NRRC Workshop onRisk-Informed Decision Making: A Survey of U.S. Experience
Tokyo, Japan, June 1-2, 2017

Nuclear Risk Research Center, CRIEPIThe B. John Garrick Institute for the Risk SciencesUCLA

Introduction
The USNRC modified its reactor inspection program in the late 1990’s• In response to substantive external concerns• In a manner that

• Emphasizes quantitative analyses (when possible)• Is publicly reviewed and discussed (and adapted, as needed)• Has successfully matured.

Risk analyses provide key information, used to• Define what is routinely inspected and how often• Assess performance of certain key equipment• Determine the significance of inspection findings.

Periodic public reviews• Provide information on individual licensee performance• Support decisions on• Incremental inspection activities• Related regulatory actions.

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Some Important History• Reengineering of USNRC’s inspection process• In the late 1990’s and early 2000’s• NRC and the US nuclear industry had developed into a generally stable industry• Expanded use of probabilistic risk analysis was being encouraged• Concerns were being raised about the NRC inspection and enforcement processes• at times not clearly focused on the most safety important issues,• consisted of redundant actions and outputs, and• were overly subjective with NRC action taken in a manner that was at times neither scrutablenor predictable.

• NRC undertook a major reengineering effort• New “reactor oversight process” was the result

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3The B. John Garrick Institute for the Risk SciencesUCLAKey ROP Elements
The Reactor Oversight Process uses a top-down regulatory
framework to define how licensee performance will be assessedAssesses each licensee’s performance using• quantitative and• qualitative information

Uses the “action matrix” to determine regulatory actions
Implements actions - making changes to the future oversight of eachlicensee

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Overall safety
missionProtect public health and safety as a result ofcivilian nuclear power operation

Strategicperformance areas
Reactor safety

CornerstonesInitiating
events

Cross-cutting issues

Mitigating
systems

Barrier
integrity

Humanperformance

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Radiation safety

Emergency
preparedness

Safety conscious
workenvironment
5

Occupationalradiation
safetyPublicradiation
safetySafeguards

Physicalprotection

Problem
identification
and resolution
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Reviewing Inspection Results & Performance Indicatorsand Defining SignificanceCornerstone

Baseline inspection
results
The Significance
Determination
Process (SDP)

Performanceindicator results

PRA-relatedindicators∆ CDF < 1X10-6
Significancethreshold

SPAR models usedhere

1X10-6 < ∆ CDF < 1X10-5
1X10-5 < ∆ CDF < 1X10-4

SignificanceThreshold∆ CDF > 1X10-4

Action matrix

Regulatory response

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6The B. John Garrick Institute for the Risk Sciences

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Reactor safety

Initiating
eventsMitigating
systems

Barrier
integrity

Emergency
preparedness

Unplanned
scramsEmergency ACpower

Reactor coolant
system specificactivity

Drill/exercise
performance

Scrams with lossof normal heatremovalHigh pressure
injection

Reactor coolant
system leakage

Heat removalUnplanned
power change

Residual heatremovalSupport coolingwater

Radiation safety

ERO drill participation

Safeguards

Occupationalradiation
safetyPublicradiation
safetyPhysicalprotection

Occupational
exposure control
effectivenessRadiologicaleffluents

Protected areaequipment

Alert and notification
system

PRA-related
indicators

Safety system
functional
failures

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Assessing the Importance of Inspection Findings –the Significance Determination Process (SDP)Objectives• To characterize the safety or security significance of inspection findings, using
best available risk insights as appropriate• To provide all stakeholders an objective and common framework forcommunicating the potential safety or security significance of inspectionfindings• To provide a basis for timely assessment and/or enforcement actions
associated with inspection findings

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Significance Determination ProcessProcess• Develop inspection findings• Characterize significance (initial staff assessment) [using SPAR models]• Obtain licensee perspectives on initial characterization• Finalize staff’s significance determination [using SPAR models]• Issue final determination letter• Provide licensee appeal opportunity

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Risk Analyses and SPAR Models
Available licensee PRAs in late 1990’s had important shortcomings• Multiple risk approaches and software tools• PRA standards were just beginning to appear• Considerable licensee-to-licensee variability• Not required to be submitted to USNRC

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Risk Analyses and SPAR Models
Using risk analyses to assessing significance of inspection findings
introduced important constraints• “user-friendly” models• Consistent modeling approaches• Additional staff• Capabilities (senior reactor analysts)• Training

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Risk Analyses and SPAR Models
SPAR models have evolved significantly• Initial development pre-dated ROP changes• ”Simplified” models for assessing implications of generic (not plant-specific) issues• With ROP change• SPAR models offered better alternative• Model evolution continues• External hazards (found to be important)• Shutdown accidents• Accident consequences (Level 2 PRA)

• Comparisons now made to improve consistency with• Plant-specific PRAs
• Standards

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Integrating Information and Defining Actions

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Decreasing licensee performance
Licensee
response

All assessment
inputs green

Regulatory
response

Degradedcornerstone

Multiple/repetitive
degradedcornerstone

Unacceptableperformance

One white input
or two white
inputs in different
cornerstones

Two white inputs
in single
cornerstone
One yellow input;
Three white
inputsMultiple yellow
inputs
One red input

Plants notallowed tooperate

Normal baseline
inspection

Increasing additional NRC inspection ~ 40 hours~ 200 hours

5X

Nuclear Risk Research Center, CRIEPI> 10x15< 3000 hours

Orders tomodify,suspend, or
revoke licensed
activities

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Taking Actions
USNRC’s inspection manual (Section 0305) defines follow-up actions
for each Action Matrix column• Licensees responsible for correcting identified issues• USNRC inspectors monitor and verify corrective actions• Licensees in column 3 may be required to take additional actions• Safety culture• Licensees in column 4 expected to have third-party safety culture evaluation• Lack of timely action can result in additional regulatory actions

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16The B. John Garrick Institute for the Risk Sciences

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The US Reactor Oversight Process has been a successful use of risk analyses• To align USNRC inspection resources• To monitor key equipment performance• To assess significance of inspection findings

Quantitative approach has improved credibility• Some important aspects are not amenable to quantitative analysis, so handled qualitatively

Key ROP elements include• A strategically driven regulatory framework• Significance determination process (using SPAR models)• Action matrix

Use of SPAR models• Provides important measure of consistency in SDP evaluations• Reflects limitations in licensee PRA models (in late 1990’s)

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