PRA R&D at NRC: Selected Topics › docs › ML1627 › ML16273A584.pdf · RG 1.174. ASME/ANS. PRA Standard. Revised. Reactor Oversight. Level 3 PRA. What: NRC Applications of PRA

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PRA R&D at NRC:Selected Topics

N. Siu and J. ChangOffice of Nuclear Regulatory Research

NRC/KAERI Cooperative Research MeetingDaejeon, Korea

October 10, 2016

Topics

• Organizational overview• Site risk assessment• External hazards• HRA• Digital I&C reliability analysis• Emerging issues in PSA• Advanced PSA• Fire PSA

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PRA at the NRC

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The who, what, when, where, how, and why

Why PRA: 1995 PRA Policy Statement

• “The use of PRA technology should be increased in all regulatory matters to the extent supported by the state-of-the-art in PRA methods and data and in a manner that complements the NRC’s deterministic approach and supports the NRC’s traditional defense-in-depth philosophy…”

• A probabilistic approach extends a traditional, deterministic approach to regulation, by:(1)Allowing consideration of a broader set of potential

challenges to safety, (2)providing a logical means for prioritizing these challenges

based on risk significance, and (3)Allowing consideration of a broader set of resources to

defend against these challenges.

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PRA at the NRC

Who/Where: NRC, Contractors, and Others

• NRC (HQ and Regions)– Analysts– Reviewers– Policy and decision makers

• National Laboratories• Private Firms• Universities

– Contracts– Grants– Fellowships

• Cooperating Organizations– Other government agencies– Industry (licensees, owners groups,

R&D)– International (IAEA, OECD/NEA)

• Standards Organizations• Public

– Industry– PRA community– General public

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PRA at the NRC

NRR

NRO

NSIR

NMSS

RES

Regions

How: Risk-Informed Decision Making (RIDM)

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PRA at the NRC

6

Adapted from RG 1.174

ChernobylTMI

When: A PRA Timeline

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1940 1950 19701960 1980 1990 20102000 2020

PRA at the NRC

NUREG-1150

AECcreated

WASH-740

Fukushima

IndianPoint

WASH-1400

NRCcreated

IPE/IPEEE

Atomic Energy Act“No undue risk”

SafetyGoalPolicy

PRAPolicy

Price-Anderson(non-zero risk)

RG 1.174

ASME/ANSPRA Standard

RevisedReactor Oversight

Level 3 PRA

What: NRC Applications of PRA

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Risk Assessment

Regulations and Guidance

Licensingand

Certification

Oversight

Operational Experience

PRA at the NRC

Fire Protection (“NFPA 805”)• Post-Browns Ferry deterministic fire

protection (10 CFR Part 50, App R)• Risk-informed, performance-based fire

protection (10 CFR 50.48, NFPA 805)– Voluntary alternative to Appendix R– Deterministic and performance-based

elements– Changes can be made without prior

approval of Authority Having Jurisdiction (AHJ)

– Ensure risk is “acceptable” to AHJ– As of January 2016:

• 24 units (16 sites) transitioned• 22 units (13 sites) in process

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From Cline, D.D., et al., “Investigation of Twenty-Foot Separation Distance as a Fire Protection Method as Specified in 10 CFR 50, Appendix R,” NUREG/CR-3192, 1983.

PRA at the NRC

Changing Plant Licensing Basis (RG 1.174)

• Voluntary changes: licensee requests, NRC reviews

• Small risk increases may be acceptable

• Change requests may be combined

• Decisions are risk-informed

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U.S. Nuclear Regulatory Commission, “An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant Specific Changes to the Licensing Basis,” Regulatory Guide 1.174, Revision 2, 2011.

PRA at the NRC

Projected Risk-Informed Licensing

• Projected ~60% increase in new risk-informed submittals (FY15/16 to FY17/18)

• Submittals of greatest increase:– Seismic PRA – 50.69 – Risk-informed GSI-191 – Tornado Missile– TSTF-505

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PRA at the NRC

Reactor Oversight Program

• Inspection planning• Determining significance of findings

– Characterize performance deficiency– Use review panel (if required)– Obtain licensee perspective– Finalize

• Performance indicators

• Non-reactor: Risk-Informed Fuel Cycle Oversight Program (RFCOP) under development

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∆CDF < 1E-6∆LERF < 1E-7

1E-6 < ∆CDF < 1E-51E-7 < ∆LERF < 1E-6

1E-5 < ∆CDF < 1E-41E-6 < ∆LERF < 1E-5

∆CDF > 1E-4∆LERF > 1E-5

CDF = Core damage frequencyLERF = Large early release frequency

PRA at the NRC

Accident Sequence Precursor Program• Program recommended by WASH-

1400 review group (1978)• Provides risk-informed view of

nuclear plant operating experience– Conditional core damage probability

(events)– Increase in core damage probability

(conditions)• Supported by plant-specific

Standardized Plant Analysis Risk models

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3(≥ 10-1)

5 (10-2 to 10-1)

26 (10-3 to 10-2)

171 (10-4 to 10-3)

260 (10-5 to 10-4)

316 (10-6 to 10-5)

64,446 Total LERs Reviewed

Licensee Event Reports 1969-2010(No significant precursors since 2002)

significant

PRA at the NRC

Accident Sequence Precursor Program

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PRA at the NRC

Other Risk-Informed Applications

• Risk Prioritization InitiativeSRM-SECY-15-0050 (ADAMS ML15237A142) - Commission did not approve RPI activities, but supported the consideration of risk insights in regulatory decision-making through existing processes

• DC/COL-ISG-028, “Technical Adequacy of the Advanced Light-Water Reactor Probabilistic Risk Assessment”

Draft for use http://pbadupws.nrc.gov/docs/ML1423/ML14230A111.pdf

• Debris Accumulation on PWR Sump Performance, GI-19150.46c - provision allowing, on a case-by-case basis, licensees to use risk-informed alternatives to address containment debris

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PRA at the NRC

Day-to-Day Models and Tools• SPAR* Models

− 79 operating plant models (event tree/fault tree)

− 4 new reactor plant models

• SAPHIRE** code− Idaho National Laboratory (NRC-

sponsored)− Features to support event and

condition analysis

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*Standardized Plant Analysis Risk **Systems Analysis Programs for Hands-on Integrated Reliability Evaluation

PRA at the NRC

Role of NRC R&D

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Adapted from National Research Council, “World-Class Research and DevelopmentCharacteristics for an Army Research, Development and Engineering Organization,”National Academy Press, Washington, DC, 1996, ISBN 0-309-05589-X.

PROCESSING SYSTEMRECEIVING

SYSTEM

OUTPUT OUTCOMESINPUT

MEASUREMENT AND FEEDBACK


R&D needsResources- People- Funds- Infrastructure- Information

Facts/KnowledgeRecommendationsMethodsModelsToolsDataGuidance

SafetySecurityEnvironment

MEASUREMENTAND FEEDBACK

NRC/RES

ActivitiesTests/experimentsSurveys/reviewsTechnical analysesDevelopment

…

Nuclear:NRC

U.S. IndustryInternational Orgs

Other:U.S. Congress

OGAsTechnical Community

General Public

PoliciesDecisionsActionsInformation


LicensedFacilities

&Activities

Other

THE WORLD

RESPONSES

PRA at the NRC

For Further Reading*• “A Proposed Risk Management Regulatory Framework,” NUREG-2150, 2012.• “Status of the Accident Sequence Precursor Program and the Standardized

Plant Analysis Risk Models,” SECY-15-0124, 2015. • “Annual Update of the Risk-Informed Activities Public Web Site,” SECY-15-

0135, 2015. (http://www.nrc.gov/about-nrc/regulatory/risk-informed/rpp.html)• “Probabilistic Risk Assessment and Regulatory Decision Making: Some

Frequently Asked Questions,” NUREG-2201, 2016.• “Historical Review and Observations of Defense-in-Depth,” NUREG/KM-0009,

2016.• Upcoming CSNI/WGRISK report on PSA use and development.

18*References can be found at www.nrc.gov

PRA at the NRC

http://www.nrc.gov/about-nrc/regulatory/risk-informed/rpp.html

http://www.nrc.gov/

Site Risk Assessment

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Level 3 PRA project overview

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• Develop a Level 3 PRA generally based on current state of practice that

– reflects technical advances since the last NRC-sponsored Level 3 PRAs – addresses scope considerations not previously considered

• Extract new insights to enhance regulatory decisionmaking and to help focus limited agency resources on issues most directly related to the agency’s mission

• Enhance PRA staff capability and expertise and improve documentation practices

• Demonstrate technical feasibility and evaluate the realistic cost of developing new Level 3 PRAs

Project Objectives


Project Scope

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• All major site radiological sources (all reactor cores, spent fuel pools, and dry storage casks).

• All internal and external hazards, all modes of plant operation. Excludes initiating events involving malevolent acts.

• Incorporates improvements in PRA technology and changes in plant operational performance and safety since completion of NUREG-1150

• Single multi-unit site; results in some limits in the general applicability of risk insights


Vogtle Electric Generating Plant

Technical Approach Philosophy

• Generally based on the current state of practice.• Considers:

1) ASME and ANS PRA standards2) Results of earlier scoping study (SECY-11-0089)3) Interactions with NRC experts4) Input from internal Technical Advisory Group (TAG)

• Peer Reviewed

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Other Considerations• Tools and models

– SAPHIRE 8– MELCOR– MACCS2– SPAR– Informed by licensee model

• Risk metrics– Public health effects and offsite economic costs– Core damage frequency and large early release frequency

• Communication plan• Documentation

– Top tier (NUREG report) will be publicly available– Lower tier (interim deliverables) will likely contain proprietary information

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Key Elements• Reactor, at-power, Level 1

– Internal events and floods– Internal fires– Seismic events– High winds, external flooding, and

other hazards• Reactor, at-power, Level 2• Reactor, at-power, Level 3• Reactor, low power and shutdown• Spent fuel pool (SFP)• Dry cask storage (DCS)• Integrated site risk

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Major Accomplishments

• Completed initial version of model and PWROG-led, ASME/ANS PRA standard-based peer review for eight scope elements:– Reactor, Level 1 PRA for internal events and floods, high winds,

and other hazards– Reactor, Level 2 PRA for internal events and floods– Reactor, Level 3 PRA for internal events and floods

• Completed initial version of reactor, Level 1 PRA for internal fires and seismic events

• Completed initial version of dry cask storage Level 1, 2, and 3 PRA

• Completed expert elicitation for frequency of interfacing systems LOCA

• Completed substantive update of reactor, Level 1 PRA for internal events and internal floods

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Other Key Activities

• Based on internal and external feedback, complete substantive updates to the following models:– Reactor, Level 2, internal event and flood PRA– Reactor, Level 3, internal event and flood PRA

• Complete initial models for:– Reactor, LPSD, Level 1, internal event PRA

• Complete revised models for:– Reactor, Level 1, seismic PRA– Reactor, Level 1, internal fire PRA

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External Hazards

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Seismic, flooding, and other hazards

Post-Fukushima Activities

• Most Phase 1 seismic and flooding re-evaluations using current guidance and methods (Stage 1) are complete.

• Licensees assessing plant response to reevaluated hazards (Stage 2) as necessary– Seismic: limited scope assessments or seismic PRA– Flooding: focused evaluations or integrated assessments

(NEI-16-05, as endorsed by JLD-ISG-2016-01)• Other external hazards

– “Assessment of Fukushima Tier 2 Recommendation Related to Evaluation of Natural Hazards Other Than Seismic and Flooding,” SECY-16-0074, June 2, 2016.

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External Hazards

Related Activities

• PFHA– Workshop (2013, summary: ML13057A642)– Ongoing: PRA analysis of total plant response to

flooding• CSNI Working Group on External Events

(WGEV)• Expert elicitation guidance development (SRM-

COMGEA-11-0001, SRM-SECY-11-0172)

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External Hazards

Potential Challenges for the Future

• PSA– Multiple, correlated hazards– Multiple mechanisms– Scale of analysis– Regional sources– Multiple units and sites– Human effects– Data (hazard and site)

• Risk-informed decisionmaking– Aggregation– Credit for portable equipment– New methods

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External Hazards

http://www.mvn.usace.army.mil/Missions/Mississippi-River-Flood-Control/Mississippi-River-Tributaries/Mississippi-Drainage-Basin/

St Lucie – Local Intense Precipitation

• Heavy rainfall January 9, 2014• Reactor Auxiliary Building flooding (some

conduits missing flood seals)– 16:10: Flooding detected– 18:03: Unusual Event (UE) declared– 00:01 January 10: UE terminated

• Potential challenge to ECCS pumps, no actual loss of safety-related accident mitigation or safe shutdown equipment

• Concern: missing flood barriers below design flood height, not identified during flooding walkdowns (NTTF Recommendation 2.3)

• “White” finding(3E-6/yr < ∆CDF < 1E-5/yr)

• LER 335-2014-001

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External Hazards

St. Lucie Plant

Human Reliability Analysis

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Current activities, emerging issues

HRA Topics

• NRC-KAERI recent interactions on HRA methods and data

• The IDHEAS HRA method developmental plan• The SACADA simulator data collection• Emerging items

– Minimum Joint HEP– Crediting FLEX equipment in risk-informed applications– HRA guidelines for control room abandonment in fire

events

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IDHEAS: Integrated Human Event Analysis SystemSACADA: Scenario Authoring, Characterization and Debriefing Application

HRA

Recent NRC-KAERI Interactions

• NRC commented on the KAERI/TR-6401/20161 draft report and received the final report from KAERI.

• KAERI commented on NRC’s IDHEAS-G draft report.• KAERI and NRC co-authored an RESS paper2 on KAERI’s

analysis of Korean NPP simulator records using the SACADA taxonomy.

• NRC (Dr. Chang) visited KAERI in 7/2016 for SACADA outreach.• KAERI invited Dr. Chang to attend the PSAM 13 and the first HRA

Society East Asia chapter meeting. • This technical information exchange meeting.• NRC plans to host an HRA data workshop in the summer of 2017.

Invitations, when available, will include KAERI and KINS.

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HRA

1KAERI/TR-6401/2016 “A framework to estimate HEPs from the full-scope simulators of NPPs: unsafe act definition, identification and quantification”

2Title:“The use of the SACADA taxonomy to analyze simulator records: insight and suggestions”

IDHEAS Strategic Development Plan

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IDHEAS-G(NUREG-2198, Vol.1)

IDHEAS-Internal, At-Power(NUREG-2199, Vol.1)

Use of Expert Judgment to Estimate HEPs

(NUREG-2199, Vol.2)

Formal Method Testing(NUREG-2199, Vol.3)

Users’ Guide(NUREG-2199, Vol.4)

Empirical Data for HEP Estimation(NUREG-2198, Vol.2)

Other Application-Specific HRA Methods

Technical Basis

General Guidance

Application Specific HRA

Methods

SupportingDocuments

SACADA Data for HEP Estimation

IDHEASSoftware

MODEL/METHOD IMPLEMENTATION

Literature Review (NUREG-2114 and others)

Operations experience,Human factors practices,Selected HRA methods

HRA


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(NUREG-2199, Vol.2)





Technical Basis

General Guidance


Methods

SupportingDocuments


IDHEASSoftware




HRA

NUREG-2114 “Cognitive Basis for HRA”: • A large psychological literature review to establish the technical

basis and the cognitive framework for the IDHEAS development• A hierarchical cognitive framework

– Macrocognitive functions: Detection, Understanding, Decision-making and planning, Action execution, and Teamwork

– Proximate causes: e.g., cues or information not perceived, and cues or information not attended to

– Mechanisms: e.g., Cue saliency, vigilance, attention, expectation, and working memory

– Performance Influencing Factors (PIFs): e.g., human system interface


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(NUREG-2199, Vol.2)





Technical Basis

General Guidance


Methods

SupportingDocuments


IDHEASSoftware




HRA

IDHEAS-G General Methodology:Provides framework and guidance for developing application-specific HRA methods.• Qualitative analysis: expected to be the same for all HRA methods to be

developed, e.g.,– Scenario narrative, event timeline, relevant operating experience,

context analysis, HFE and critical task identification, etc.• Quantitative analysis: provide guidance on HEP quantification and a

Basic Quantification Structure with a basic set of cognitive failure modes and a comprehensive list of PIFs; Application-specific quantification methods can be derived from the Basic Quantification Structure.


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(NUREG-2199, Vol.2)





Technical Basis

General Guidance


Methods

SupportingDocuments


IDHEASSoftware




HRA

Empirical Data for IDHEAS-G:Experimental and operational data, nuclear and non-nuclear, about the effects of PIFs on HEPs in psychological literature and other sources.


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(NUREG-2199, Vol.2)





Technical Basis

General Guidance


Methods

SupportingDocuments


IDHEASSoftware




HRA

SACADA Data:• Collect licensed operator

simulator training performance data

• Designed to be used by nuclear power plants’ simulator training program for a long term data collection


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(NUREG-2199, Vol.2)





Technical Basis

General Guidance


Methods

SupportingDocuments


IDHEASSoftware




HRA

IDHEAS Internal, At-Power Method(NUREG-2199, Vol.1 ):• Jointly developed by NRC and EPRI for

NPP internal events at-power HRA applications

• HEP calculation is based on crew failure modes (CFMs) and PIFs

• The CFMs and PIFs coverage is a sub-set of IDHEAS-G’s CFMs and PIFs.


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(NUREG-2199, Vol.2)





Technical Basis

General Guidance


Methods

SupportingDocuments


IDHEASSoftware




HRA

Use of Expert Judgment to Estimate HEPs (NUREG-2199, Vol. 2):• Expert elicitation was used to estimate

the HEPs of the CFMs in the IDHEAS Internal, At-Power method.

• This report documents the expert elicitation process and results


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(NUREG-2199, Vol.2)





Technical Basis

General Guidance


Methods

SupportingDocuments


IDHEASSoftware




HRA

Formal Testing of IDHEAS method for internal at-power applications (NUREG-2199, Vol. 3):• Five HRA analyst teams of NRC and

U.S. industry tested the IDHEAS Internal, At-Power method

• This report documents the test process and results.


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(NUREG-2199, Vol.2)





Technical Basis

General Guidance


Methods

SupportingDocuments


IDHEASSoftware




HRA

Other Application-Specific HRA Methods:• Priority application is the NRC’s

event and condition analysis application, i.e., the significance determination process and the accident sequence precursor analyses. May include crediting FLEX equipment.

• Other potential applications include spent fuel handling and radiation medical treatments


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(NUREG-2199, Vol.2)





Technical Basis

General Guidance


Methods

SupportingDocuments


IDHEASSoftware




HRA

IDHEAS Software:• To facilitate HRA implementation

and documentation• Based on the IDHEAS-G to provide

“tool sets” and templates, for the HRA method developers to develop the application-specific HRA software for HRA analysts to use.

Schedule

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HRA




(NUREG-2199, Vol.2)





Technical Basis

General Guidance


Methods

SupportingDocuments


IDHEASSoftware




1/2016

9/2016

2017/2018

2017

2017*

2018

2018

2017/2018

Green: Technical work completed; Red: Anticipated technical work completion date;*First focus is on NRC’s Event and Condition Analyses

More About SACADA

• Agreements:– Raw Data: South Texas Project, Halden Reactor Project,

Taiwan Power Company (TPC), and DOE Advanced Test Reactor

– General: KAERI and the ÚJV Řež, a. s.• Recent and planned activities:

– Assist TPC to pilot SACADA (training, Chinese version, and tech support)

– Complete time data collection function– External data analysis: expected to be completed in summer

2017; a workshop in 2017 is planned.– Domestic outreach: Issuing a Regulatory Information Summary

and a poster in the Regulatory Information Conference, 201746

HRA

Minimum Joint HEP (MJHEP)

• Issue: – Some industry PRAs calculate cutsets with multiple

HFEs by directly multiplying their HEPs. This generates extremely low joint HEPs. This could underestimate the actual risk.

– NRC is considering to establish a Minimum JHEP requirement in PRA standards. Inappropriate MJHEP establishment could significantly affect risk assessment results.

• The NRC plans to gain more understanding about MJHEP.

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HRA

Crediting FLEX Equipment in Risk-Informed Applications

• FLEX equipment provided in response to Mitigating Strategies order (EA-12-049)

• U.S. industry wishes to take credit for FLEX equipment in risk-informed applications

• HRA method to analyze FLEX implementation could be useful

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HRA

Main Control Room Abandonment in Fire Events• An NRC-EPRI joint project to provide HRA guidelines

supplementing NUREG-1921.• Two-phase development

– Phase I:• MCRA scenario development, including consideration of the decision to

abandon• Human failure event (HFE) definition and identification• Qualitative HRA specific to MCRA scenarios, including consideration of

performance shaping factors (PSFs) and other influences on operator performance

– Phase II: Quantification• Phase I draft report is out for review.

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HRA

Digital I&C Reliability Analysis

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Program overview, current activities

Digital I&C R&D for PRA

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Research Goal:Determine if it is practical and useful to incorporate digital systems into PRAs

Ongoing R&D

• Statistical Testing Method– Use of PRA context to determine operational profile for digital

system (to test system as it is used)– Directly estimate software failure probability using testing

failure data– Completed “black box” testing of nuclear plant control system– See M. Li and K. Coyne, “Risk-Informed Statistical Testing for

Digital I&C Systems,” WGRISK Annual Meeting, March 2016. (ML16069A188)

• Bayesian Belief Network– Use software development environment characteristics (e.g.,

quality attributes) and expert judgment to estimate fault density– Convert fault density to reliability– NUREG/CR in preparation

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Emerging Issues in PSA

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Challenges to the technology

Example Challenges

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Developers

Analysts Users

• Time• Resources• Biases/heuristics• Communication

• Data• Bounding/screening• Guidance• “Holes”• Integration• Imagination

• New science/engineering• Operational experience• Intended users/applications• Computational limits• Rewards

• Understanding• Uncertainties• Heterogeneity and

aggregation• Confidence• Other Factors (e.g.,

DID, safety margins)• Stakeholders

Emerging Issues

New Experiments and Analyses

• High Energy Arc Faults (HEAF) in cabinets• Aqueous transport of accident-generated wastewater• State-of-the-Art Consequence Analysis (SOARCA)

55

Emerging Issues

Operational Experience• 3/11/2011 – Great East Japan Earthquake• Meltdowns at Fukushima Dai-ichi, varying

challenges at Fukushima Dai-ni, Onagawa, Tokai Dai-ni, Higashidori

• PRA technology challenges– Severe external hazards– Human performance– Command and control– Organizational response– Accident-induced environment– Offsite consequences– Searching for failures

• A knowledge management challenge: lessons from past events (e.g., Blayais, 1999)

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TEPCO photos from “The Yoshida Testimony,” Asahi Shinbun, 2014.

Emerging Issues

Integration of Multiple Disciplines

• Technical domains• Culture

– Language– Accepted methods– Unstated assumptions– Views on uncertainty

• Limited total resources

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Here be dragonsLoose nuts

Really big mountains

Emerging Issues

Stakeholder Views

• Provides strategic direction to advance use of risk-informed decisionmaking

• Formed October 2013• Public meetings• Coordinated working groups

– Technical adequacy (including new methods approval)

– Uncertainty in decision making (including aggregation)

– Credit for mitigating strategies

5858

Adapted from RG 1.174

NRC Risk-Informed Steering Committee

Emerging Issues

Advanced PRA

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Methods and tools to help the analyst

Looking Beyond Current Approaches

• Simulation-oriented methods– Discrete dynamic event trees (DDETs)– Direct stochastic simulation

• Information management and utilization– Content analytics– Formal models (e.g., “Open PSA”)

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Advanced PRA

Discrete Dynamic Event Trees

• Rationale– Retains benefits of dynamic approach

• Context for operator actions• “Natural language” framework for integrated

multidiscliplinary analysis• Reduced need for intermediate modeling

approximations– Discrete scenario format supports understanding

• Diagnostics => confidence, contributors• “Story telling” => actionable lessons

– Leverages principal external activities• NRC-supported activities

– Feasibility study (Sandia)– Event analysis (UCLA)

61

Advanced PRA

ImprovedRealism

Direct Stochastic Simulation

• Rationale– Retains benefits of dynamic approach (as with

DDET)– Literal system representation facilitates user

understanding of conceptual model– Natural approach for many engineering disciplines,

leverages• Current education and training• General purpose software packages• Specialized software for analogous applications

• A current application: vulnerability assessments62

Advanced PRA

Information Management and Utilization

• January 14, 2011: Prime-time “Jeopardy!” demonstration of IBM Watson

• Post-show shifting aims to get at “Big Data”– From oracle to aide– Near-term technology (Content Analytics)

• Alternate vision: intelligent personal assistants– Siri (2011)– Google Now (2012)– Cortana (2014)

63

Advanced PRA

IBM, 2011

Risk Information Characteristics

• Risk triplet (qualitative + quantitative)• Systems viewpoint

– Multiple technical disciplines– Problem scale and complexity (multitude of

scenarios, interacting pieces)– Diverse and implicit sources of information (licensing

basis, operating experience, past analyses, …)• Rare events• Broad user base

64

Advanced PRA

Long-Term Research Program Project:“Advanced Knowledge Engineering Tools to Support Risk-Informed Decision Making”

• ObjectiveAssess feasibility/need for additional NRC effort

• Scope– IBM Content Analytics 2.2 (ICA 2.2)– Selected case studies (“use cases”)– Reduced database (330,000 documents)

• Participants– Subject matter experts– Software engineers

65

Advanced PRA

Technical Approach

• Use Cases (UC)– 1: Multi-unit events (identify and characterize)– 2: Current plant CDFs (find estimates)– E: Database exploration (open-ended “discovery”)

• Process– Specify search problem– Develop customized search tool (facets/keywords)– Test and refine tool– Demonstrate final tool, compare against alternatives

• Informed search• Basic search

66

Advanced PRA

Conclusions

• ICA 2.2 is a human-in-the-loop tool (not a push-button solution)• Subject matter experts must be involved during tool development• For most tests, ICA 2.2 was effective and efficient, and showed some

advantages over alternative methods (ADAMS P8/Public/Enterprise Search, Acrobat, Google, LERSearch)

– Reasonable speed– Helpful interface (contextual text, query construction and saves)– Supports more open-ended explorations

• Useful side-benefits– Early familiarization and use of next generation ADAMS tools– A more powerful and useful LER search tool– Improved knowledge of use case subject matter

67

Advanced PRA

Fire PSA

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Fire PRA realism, recent experiments

Fire PRA Realism Project*

• Purpose: assess concerns over conservatism and realism of fire PRA potentially affecting– Transitions to NFPA 805– Future risk-informed applications

• Topics explored– Technical maturity– Precursor-based estimation of CDF– Relative contributions to CDF– Qualitative comparisons with operational experience– Technology developments– Applications environment and infrastructure

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Fire PSA

*See N. Siu, K. Coyne, and N. Melly, “Fire PRA Maturity and Realism: A Technical Evaluation,” technical opinion paper, January 2016. (ML16022A266)

Comparing Fire PRA with Precursors

70

(w/Browns Ferry)

(w/o Browns Ferry)

Fire PSA

Changing CDF Contributions

71

Fire PSA

Fire PRA Maturity and Realism

• Conclusions– Fire PRA is in an intermediate-to-late stage of maturity– Results may be conservative; degree is uncertain– Qualitative results compare well with operational

experience; some specific concerns– Improvements are underway; operational experience

reviews should be used to identify/prioritize– Realism is affected by the pool of trained staff and

analyst attitudes/beliefs (e.g., regarding the use of conservative assumptions to compensate for uncertainties)

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Fire PSA

Fire R&D: Recent/Upcoming Products

• “Heat Release Rates of Electrical Enclosure Fires (HELEN-FIRE),” NUREG/CR-7197, 2016.

• “Refining And Characterizing Heat Release Rates from Electrical Enclosures During Fire (RACHELLE-FIRE),” NUREG-2178, 2016.

• “Technical Resolution to Outstanding Issues on Nuclear Power Plant Fire-Induced Circuit Failure,” NUREG/CR-7150, Vol. 3, in preparation.

• “Determining the Effectiveness, Limitations, and Operator Response for Very Early Warning Fire Detection Systems in Nuclear Facilities (DELORES-VEWFIRE),” NUREG-2180, draft for comment, 2015.

• “Verification and Validation of Selected Fire Models for Nuclear Power Plant Applications,” NUREG-1824, Supplement 1, draft for comment, 2014.

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See M. H. Salley, “NRC Fire Research Current Research Activities,” NEI Fire Protection Information, September 12-14, Atlanta, GA. (ML16256A822)

Fire PSA

Response Bias of Electrical Cable Coatings at Fire Conditions (REBECCA-Fire)• Objectives

– Regulatory: provide confirmatory data to support enhancements to NUREG/CR-6850 guidance

– Technical: Determine ignition characteristics, flame spread and electrical performance of electrical cables with fire retardant coating materials

• Small- and intermediate-scale experiments at Sandia and NIST

• Preliminary results (vertical trays)– Flame spread on some uncoated samples, not

on coated samples– Delayed damage time for coated, non-qualified

cables– Some coated, qualified cables may fail earlier

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Fire PSA

NIST vertical flame spread test configuration

High Energy Arc Faults (HEAF)

• Fire PRA guidance (NUREG/CR-6850, Appendix M) – Based on San Onofre (2001)– Single methodology for 480V and above– Assumes next upstream over-current protection

device will trip open (failed at Robinson, 2010)• OECD-FIRE: 48 out of 415 events (through

mid-2012)• Visible contributor to fire CDF for some plants

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Fire PSA

Example HEAF Test

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480V switchgear, 42 kA, 8 secProject information: http://www.oecd-nea.org/jointproj/heaf.html

http://www.oecd-nea.org/jointproj/heaf.html

Test 23 Switchgear IP20 EnclosureOctober 16, 2015• Aluminum bus, low-voltage

pre-test• Immediate ignition of

internal combustible load• Melted inconel plate

thermocouples 3 ft away Aluminum byproducts coated test cell, shorted other electrical components

• No immediate safety concern (ML16064A250)

• Entered into Generic Issue Program (ML16126A091)

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Before After

Additional Slides

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NRC Organization

• Headquarters + 4 Regional Offices

• 5 Commissioners• ~3350 staff (FY 2016)• Annual budget ~$1B• Website: www.nrc.gov• Information Digest:

NUREG-1350 V27

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http://www.nrc.gov/

http://pbadupws.nrc.gov/docs/ML1525/ML15254A321.pdf

NRC Mission

“The U.S. Nuclear Regulatory Commission licenses and regulates the Nation’s civilian use of radioactive materials to protect public health and safety, promote the common defense and security, and protect the environment.”

- NUREG-1614 (NRC Strategic Plan)

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Risk-Informed Regulations• Backfitting (10 CFR 50.109)• Station blackout protection (10 CFR 50.63)• Maintenance management (10 CFR 50.65)• Combustible gas control (10 CFR 50.44)• Fire protection (10 CFR 50.48)• Reactor pressure vessel protection (10 CFR 50.61a)• Special treatment of structures, systems, and components

(10 CFR 50.69)• New reactor certification and licensing (10 CFR 52.47)

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Risk-Informed Licensing

• Changes in plant licensing basis• Environmental reviews• Application of risk-informed regulations

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Risk-Informed Oversight

• Reactor oversight process• Incident investigation• Enforcement discretion

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Risk-Informed Operational Experience

• Accident precursors• Emergent issues• Generic issues

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85

Data Sources

SPARModels

RADSDatabase

CCFDatabase

EPIX MSPIUAs LERs

Monthly OperatingReports

Fire Events

Integrated Data Collection and Coding System

Risk-Based Operating Experience Analyses

LERSearch

ASPDB

MitigatingSystems

PerformanceIndex

Signif icanceDetermination

Process

ASPProgram

OperatingExperience

Clearinghouse

InspectionProgram

IndustryTrends

Program

Public(External)

NRC Staf f(Internal)

Fire EventsInitiatingEvents

Comp. Studies(Parm. Est. + Eng.)

System Studies(SPAR and EPIX)

CCFParameters

SpecialStudies

Tool

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ases

Dat

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Pro

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s

Operating Experience Data

Blayais (12/27/1999)• Storm during high tide in Gironde River

estuary• Overtopping of protective dyke• Loss of

– Offsite power (Units 2 and 4) – wind– Essential service water (Unit 1, Train A), low head

safety injection and containment spray pumps (Units 1 and 2), site access – flooding

– Site accessibility

• Papers in 2005 IAEA workshop following Indian Ocean tsunami

• Presentation at 2010 USNRC Regulatory Information Conference

• Little notice in PSA community

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E. De Fraguier, “Lessons learned from 1999 Blayais flood: overview of EDF flood risk management plan,” U.S. NRC Regulatory Information Conference, March 11, 2010.

TVA File Photo

Browns Ferry (3/22/1975)• Candle ignited foam

penetration seal, initiated cable tray fire; water suppression delayed; complicated shutdown

• Second-most challenging event in U.S. nuclear power plant operating history

• Spurred changes in requirements and analysis

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8.5m 11.5m

3m

Adapted from NUREG-0050

Some Fire-Induced “Near Misses”

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Event Summary Description*Browns Ferry(BWR, 1975)

Multi-unit cable fire; multiple systems lost, spurious component and system operations; makeup from CRD pump

Greifswald(VVER, 1975)

Electrical cable fire; station blackout (SBO), loss of all normal core cooling for 5 hours, loss of coolant through valve; recovered through low pressure pumps and cross-tie with Unit 2

Beloyarsk (LWGR, 1978)

Turbine lube oil fire , collapsed turbine building roof, propagated into control building, main control room (MCR) damage, secondary fires; extinguished in 22 hours; damage to multiple safety systems and instrumentation.

Armenia(VVER, 1982)

Electrical cable fire (multiple locations), smoke spread to Unit 1 MCR, secondary explosions and fire; SBO (hose streams), loss of instrumentation and reactor control; temporary cable from emergency diesel generator to high pressure pump

Chernobyl (RBMK, 1991)

Turbine failure and fire, turbine building roof collapsed; loss of generators, loss of feedwater (direct and indirect causes); makeup from seal water supply

Narora(PHWR, 1993)

Turbine failure, explosion and fire, smoke forced abandonment of shared MCR; SBO, loss of instrumentation; shutdown cooling pump energized 17 hours later

*See NUREG/CR-6738 (2001), IAEA-TECDOC-1421 (2004)

Potential PRA Technology Challenges Revealed by Fukushima*• Extending PRA scope

– Multiple sources– Additional systems– Additional organizations– Post-accident risk

• Treating feedback loops• Reconsidering intentional

conservatism• Treating long-duration scenarios

– Severe accident management– Offsite resources– Aftershocks– Success criteria

• Improving human reliability analysis– Errors of commission– Severe accident management– Psychological effects– Recovery feasibility and time delays– Uncertainty in actual status– Cumulative effects over long-duration

scenarios– Crew-to-crew variability

• Uncertainty in phenomenological codes

• Increasing emphasis on “searching”

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*From Siu, N., et al., “PSA Technology Challenges Revealed by the Great East Japan Earthquake,” PSAM Topical Conference in Light of the Fukushima Dai-Ichi Accident, Tokyo, Japan, April 15-17, 2013. (ADAMS ML 13099A347 and ML13038A203)

Advanced Topic Projects

• Current– Methodological and Software Enhancements of Dynamic PRA

Platforms for Event Assessment Applications, UCLA– Study of the Implication of Multi-Unit Accidents in the Context of

NRC’s Quantitative Health Objectives, UMD– Severe Accident Management Guideline (SAMG) Validation

within the Context of Severe Accident Uncertainties, OSU• Recent

– Dynamic PRA Scoping Study, 2009 (ML092820446, ML092650237)

– ADS-IDAC/MELCOR coupling, demonstration problem, Sandia/UMD/OSU, 2012 (ML12305A351, ML100810206, ML120300281)

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