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Fire Risk Assessment for Nuclear Power Plants

Nathan SiuOffice of Nuclear Regulatory Research

Lecture: FPE 580R – Fire Risk Assessment and PolicyWorcester Polytechnic University

December 2, 2015

PreludeWhat are we talking about and why?

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How it started…

• Browns Ferry Nuclear Power Plant (3/22/75)

• Candle 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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Prelude

8.5m 11.5m

3m

Adapted from NUREG-0050

Browns Ferry Timeline

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Prelude

Current Issues – An Example

• High Energy Arc Faults (HEAF) in cabinets

• Operational events, e.g.,– Robinson (2010)– Onagawa (2011)

• Potentially important contributor to fire risk

• Multi-national experimental program

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From “Roadmap for Attaining Realism in Fire PRAs,” NEI, December 2010. (ML103430372)

Prelude

A more recent view…

OECD/NEA HEAF Project

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

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Prelude

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Outline of Talk

• Prelude• U.S. Nuclear Regulatory Commission (NRC)

Overview• Probabilistic Risk Assessment (PRA) at the

NRC• Fire PRA Methodology• Fire PRA History and Results• Current Challenges• Closing Thoughts

Prelude

Key Messages

• NRC uses PRA to support regulatory decision making (day-to-day and major decisions).

• Fire is a potentially important contributor to nuclear power plant risk.

• The general approach for performing fire PRA is well understood and well accepted.

• Details matter. Concerns with the realism of specific models affect confidence in overall results and the transition to risk-informed fire protection, and are spurring R&D.

• PRA is a tool, not an end. Fire PRA R&D is focused on improvements that will support practical risk management.

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Prelude

NRC OverviewWho we are and what we do

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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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NRC Overview

NRC Functions

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NRC Overview

NRC Organization

• Headquarters + 4 Regional Offices

• 5 Commissioners• ~4000 staff• Annual budget ~$1B• Website: www.nrc.gov• Information Digest:

NUREG-1350 V27

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NRC Overview

PRA at the NRCHow we define and estimate risk, and why

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On the Definition of “Risk”

• Triplet (vector) definition (Kaplan and Garrick, 1981): {si , Ci , pi }– What can go wrong?– What are the consequences?– How likely is it?

• Common definition (∑𝑖𝑖 𝑝𝑝𝑖𝑖 × 𝐶𝐶𝑖𝑖) does not capture difference between high-probability/low-consequence events and low-probability/high-consequence events

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PRA at the NRC

From Farmer, F.R., “Reactor safety and siting: a proposed risk criterion,” Nuclear Safety, 8, 539-548(1967).

Probabilistic Risk Assessment (PRA)

• A systems-oriented engineering analysis process that answers the risk triplet questions

• Unique/challenging analysis features– Sparse data– Explicit treatment of uncertainties– Cross-disciplinary scope

• Distinguishing features (nuclear power plant PRAs)– Plant operational mode– Hazards considered– Scenario endpoints

• Also called “Probabilistic Safety Assessment” (PSA)

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PRA at the NRC

PRA (cont.)

• Benefits (as compared with alternative analysis approaches)– Aimed at decision support– Engineering-oriented– Integrated– Systematic– Realistic– Supportive of “what-if” analyses– Open

• Typically involves event tree and fault tree analysis (but doesn’t have to)

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PRA at the NRC

Example Event Tree

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PRA at the NRC

Example Fault Tree

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PRA at the NRC

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

What: All NRC Functions

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PRA at the NRC

Risk Assessment

Who: NRC Staff, Contractors, and Others

• NRC Staff (HQ and Regions)– Analysts– Reviewers

• National Laboratories• Universities

– Contracts– Grants– Fellowships

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

R&D)– International (IAEA, OECD/NEA)

• Standards Organizations• Public

– Industry– PRA/PSA community– General public

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PRA at the NRC

NRR

NRO

NSIR

NMSS

RES

Regions

How: Risk-Informed Decision Making

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PRA at the NRC

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The proposed change meets the current regulations unless

it is explicitly related to a requested exemption or rule

change

The proposed change is consistent with the defense-in-

depth philosophy The proposed change maintains sufficient safety

margins

When proposed changes result in an increase in core damage frequency and/or risk, the

increases should be small and consistent with the intent of the

Commission’s Safety Goal Policy Statement

The impact of the proposed change should be monitored

using performance measurement strategies

Integrated Decision Making

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

Risk Management• NRC’s risk-informed decisions can be industry-wide or

licensee-specific– Industry-wide decisions (e.g., new regulations) consider effect on

operating fleet– Licensee applications are voluntary (PRAs are not required); NRC

reviews and approves applications• NRC is currently exploring means to increase use of risk

information in decision making

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PRA at the NRC

From U.S. Nuclear Regulatory Commission “A Proposed Risk Management Regulatory Framework,” NUREG-2150, 2012.

Fire PRA MethodologyTailoring the approach to meet analysis needs

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Before tripAfter trip1 hour1 day1 week

3300 MWt260 MWt

50 MWt15 MWt

7 MWt

Nuclear Design 101: How Things Work

• Risk = {si, Ci, pi}• Nuclear fission →

heat → steam → electricity

• Chain reaction controlled/stopped by control rods

• Heat generation continues after chain reaction is stopped (“decay heat”)

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Fire PRA

Nuclear Power Plant Design Features

• General Design Criteria (10 CFR Part 50, Appendix A)

http://www.ecfr.gov/cgi-bin/text-idx?SID=5aa0f7b9ce8da0f9bd8aa303f964c67a&mc=true&node=ap10.1.50_1150.a&rgn=div9

• Key safety principles– Defense-in-depth– Single failure criterion and

redundancy– Diversity

• Robust structures, separation

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Fire PRA

Why Pay Attention to Fire?

• Actual events + study results => Potentially important contributor (Completeness)

• Single fire event might affect multiple systems, structures, and components (Dependencies)– P{A and B} ≠ P{A} x P{B} – Common enclosures– Defeat separation– Effects on plant operators

• Nature of scenario affects fixes (Risk Management)

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Fire PRA Methodology

Fire PRA Methodological Framework

• Performed as part of plant PRA

• Elements mirror NPP fire protection defense-in-depth

• Basic methodology developed and applied in early 1980s

• Refinements added over time (NUREG/CR-6850)

• Analysis is iterative• Current work focused on

improving data and specific models

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Fire PRA Methodology

Fire Frequency Analysis

• Objectives– Identify and characterize

potentially significant fire scenarios

– Estimate scenario frequencies

• Data: historical fire events• Estimation

– Generic– Plant-specific

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Fire PRA Methodology

Equipment Damage Analysis

• Objectives– Identify potentially significant

combinations of equipment that can be damaged by a fire scenario

– Estimate conditional probabilities of equipment failure modes, given a fire scenario

• Underlying model: competition between damage and suppression processes

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Fire PRA Methodology

Damage occurs if tdamage < tsuppression

Equipment Damage Analysis Elements

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Fire PRA Methodology

Equipment Damage Analysis (cont.)

• Prediction of fire environment– Correlations– Zone models– CFD models

• Equipment response/component fragility– Temperature and/or heat flux thresholds– Empirical data and probabilistic models for specific failure

modes (e.g., spurious operation, high-energy arc faults)

• Fire suppression– Historical data– Fire brigade drills

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Fire PRA Methodology

Plant Response Analysis

• Objectives– Identify potentially significant

fire-induced accident scenarios– Estimate fire-induced core

damage frequency (CDF)• General approach: propagate

fire-induced losses through event tree/fault tree model– Start with internal events model– Modify to include effects on

equipment availability and operator actions

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Fire PRA Methodology

Fire PRA History and ResultsLessons and applications over the years

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Fire PRA in the U.S.

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1975 1980 19901985 1995 2000 20102005 2015

Bro

wns

Fer

ry fi

re(W

ASH

-140

0 an

alys

is)

Fire PRA R&D

IPEEEsIndustry

Full-Scope PRAs

NUREG-1150/RMIEP

NFPA 805 LARs

NFPA 805, 10 CFR 50.48(c),RG 1.205, NEI 04-02,

EPRI 1011989/NUREG/CR-6850, …

Fire PRA History and Results

Some Decisions Supported by Fire PRA

• Indian Point continued operation• Risk-informed plant licensing

basis changes• Reactor oversight

– Inspection prioritization– Finding significance

• Plant license renewals (cost-beneficial severe accident mitigation alternative analyses)

• Fire protection program transitions (per NFPA 805)

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Fire PRA History and Results

Fire PRAs – Early Summary Results

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Fire PRA History and Results

Fire PRAs – More Recent Results

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Fire PRA History and Results

Fire PRAs – Risk Contributors

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Fire PRA History and Results

From Canavan, K., R. et al., “Roadmap for Attaining Realism in Fire PRAs,” Nuclear Energy Institute, 2010.

Then…Now…

Current ChallengesTransitioning to risk-informed, performance-based fire protection

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Towards Risk-Informed Fire Protection• Post-Browns Ferry deterministic

protection of redundant safe shutdown equipment (10 CFR Part 50, App R)– 3-hour fire barrier, OR– 20 feet separation with detectors and

auto suppression, OR– 1-hour fire barrier with detectors and auto

suppression• 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 AHJ– Ensure risk is “acceptable” to AHJ

• As of 3/2015, license amendment requests from 27 of 61 sites

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Current Challenges

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

A “Heated Debate”

• Is fire PRA mature? Are the results overly conservative?

• Industry concerns– Expense of detailed analyses– Realism of specific sub-models– Flexibility in making plant

changes– Implications for other risk-

informed applications• NRC concerns

– Technical basis for alternative models

– Implications for other risk-informed applications

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Current Challenges

Adapted from 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.

Common View: The Need for Realism

• Excessive conservatism or optimism can– Inappropriately focus decision maker attention– Lead to wasteful or even harmful “solutions”– Miss opportunities for other improvements– Damage stakeholder confidence

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?

Current Challenges

Fire R&D: The “Laundry List” (c. 1998)

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Current Challenges

Fire R&D at NRC• Fire is one of many contributors to risk; resources for R&D

and for performing analyses are limited.• NRC R&D activities

– primarily aimed at supporting practical regulatory office needs (review/acceptance of new technologies and methods, understanding of new phenomena)

– support current fire PRA framework (“evolution”)• Examples

– Guidance development/updating (cooperative with industry)– Experiments to provide basic data for complex phenomena, expert

panels to interpret data– International cooperation (sharing lessons from operational

experience, experiments)

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Current Challenges

Fire R&D at NRC (cont.)• Example Topics/Projects

– Cable Response to Live Fire (CAROLFIRE)– Cable Heat Release Ignition, and Spread in Tray Installations

During FIRE (CHRISTIFIRE) – Direct Current Electrical Shorting in Response to Exposure Fire

(DESIREE-FIRE)– Refining and Characterizing Heat Release Rates from Electrical

Enclosures During Fire (RACHELLE-FIRE)– Fire Events Database– High Energy Arc Fault (HEAF)

• Partners– Electric Power Research Institute (EPRI)– National Institute of Standards and Technology (NIST)– Department of Energy (DOE) National Laboratories– International Partners (OECD/NEA)– Universities

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Current Challenges

Closing ThoughtsKey messages, Fukushima implications, references

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PRA Implications of Fukushima• PRA remains a useful decision support tool

– PRAs identify and quantify possibilities; they do not “predict”

– PRAs look beyond the design basis and past operational experience; the ideal is to search for failures using all relevant information

– PRAs recognize and treat uncertainties– Global statistical estimates and “worst case”

analyses have their own modeling assumptions– Changing analytical approaches will not overcome

fundamental weaknesses in available information

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Closing Thoughts

• Fukushima: indicates the importance of all severe spatial hazards (not just earthquakes and flooding) highlights several areas where the PRA state-of-practice and the PRA state-of-the-art need

improvement• A broader lesson: lessons from past events (e.g., Blayais, 1999) need to be better

disseminated and institutionalized

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

Key Messages

• NRC uses PRA to support regulatory decision making (day-to-day and major decisions).

• Fire is a potentially important contributor to nuclear power plant risk.

• The general approach for performing fire PRA is well understood and well accepted.

• Details matter. Concerns with the realism of specific models affect confidence in overall results and the transition to risk-informed fire protection, and are spurring R&D.

• PRA is a tool, not an end. Fire PRA R&D is focused on improvements that will support practical risk management.

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Closing Thoughts

For Further Reading*• Electric Power Research Institute and U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research,

“EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities,” EPRI 1011989 and NUREG/CR-6850, 2005.• Haskin, F.E., et al., “Perspectives on Reactor Safety,” NUREG/CR-6042, Rev. 2, 2002.• Kaplan, S. and B.J. Garrick, “On the quantitative definition of risk,” Risk Analysis, 1, 11-37(1981).• Nowlen S.P., M. Kazarians, and F. Wyant, “Risk Methods Insights Gained from Fire Incidents,” NUREG/CR-6738, 2001.• Siu, N., N. Melly, S.P. Nowlen, and M. Kazarians, “Fire Risk Analysis for Nuclear Power Plants,” to be published in the

next Society for Fire Protection Engineers’ Handbook of Fire Protection Engineering.• Siu, N., K. Coyne, and N. Melly, “Fire PRA Maturity and Realism: A Technical Evaluation,” white paper in preparation.• Siu, N., et al., “Probabilistic Risk Assessment and Regulatory Decision Making: Some Frequently Asked Questions,”

report in preparation.• U.S. Nuclear Regulatory Commission, “Use of Probabilistic Risk Assessment Methods in Nuclear Activities: Final Policy

Statement,” Federal Register, Vol. 60, p. 42622 (60 FR 42622), August 16, 1995.• 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.• U.S. Nuclear Regulatory Commission, “A Proposed Risk Management Regulatory Framework,” NUREG-2150, 2012.• U.S. Nuclear Regulatory Commission, “The Browns Ferry Nuclear Plant Fire of 1975 Knowledge Management Digest,”

NUREG/KM-0002, 2013.• U.S. Nuclear Regulatory Commission, “Fire Protection and Fire Research Knowledge Management Digest, 2013”

NUREG/KM-0003, 2014.• U.S. Nuclear Regulatory Commission, “No Undue Risk: Regulating the Safety of Operating Nuclear Power Plants,”

NUREG/BR-0518, 2014.

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Closing Thoughts

*Most of these references can be found at www.nrc.gov

Additional Slides

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Some Acronyms• AB – Auxiliary Building• AC – Alternating Current• AEC – U.S. Atomic Energy Commission• ACRS – Advisory Committee of Reactor Safeguards• AHJ – Authority Having Jurisdiction• ANS – American Nuclear Society• ASME – American Society of Mechanical Engineers• ASP – Accident Sequence Precursor• BWR – Boiling Water Reactor• CCDP – Conditional Core Damage Probability• CDF – Core Damage Frequency• CFD – Computational Fluid Dynamics• CFR – Code of Federal Regulations• CRD – Control Rod Drive• CSR – Cable Spreading Room• DC – Direct Current• DOE – U.S. Department of Energy• ECCS – Emergency Core Cooling System• EPRI – Electric Pow er Research Institute• GI – Generic Issue• GW - Gigaw att• HEAF – High Energy Arc Fault• HPCI – High Pressure Coolant Injection• HRA – Human Reliability Analysis• IAEA – International Atomic Energy Agency• IPE – Individual Plant Examination• IPEEE – Individual Plant Examination of External Events• LER – Licensee Event Report• LERF – Large Early Release Frequency• LOOP – Loss of Offsite Pow er• LWGR – Light Water Graphite Reactor• MCR – Main Control Room• MW – Megaw att• NEA – Nuclear Energy Agency

• NEI – Nuclear Energy Institute• NFPA – National Fire Protection Association• NIST – National Institute of Standards and Technology• NMSS – NRC Office of Nuclear Material Safety and Safeguards• NPP – Nuclear Pow er Plant• NRC – U.S. Nuclear Regulatory Commission• NRO – NRC Office of New Reactors• NRR – NRC Office of Nuclear Reactor Regulation• NSIR – NRC Office of Nuclear Security and Incident Response• NUREG – NRC report designator• OECD – Organization for Economic Cooperation and Development• PHWR – Pressurized Heavy Water Reactor• PRA – Probabilistic Risk Assessment• PSA – Probabilistic Safety Assessment• PWR – Pressurized Water Reactor• RBMK – Reaktor Bolshoy Moshchnosti Kanalnyy• RCIC – Reactor Core Isolation Cooling• RES – NRC Office of Nuclear Regulatory Research• RG – Regulatory Guide• RIDM – Risk-Informed Decision Making• RMIEP – Risk Methods Integration and Evaluation Program• ROP – Reactor Oversight Program• SAMA – Severe Accident Mitigation Alternative• SAMDA – Severe Accident Mitigation Design Alternative• SDP – Signif icance Determination Process• SBO – Station Blackout• SECY – NRC Office of Secretary (also designator for staff papers)• SPAR – Standardized Plant Analysis Risk• SRP – Standard Review Plan• SRV – Safety Relief Valve• SSC – Systems, Structures, and Components• TMI – Three Mile Island• VVER – Vodo-Vodyanoi Energetichesky Reaktor• WASH – AEC report designator

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U.S. Nuclear Power Plants

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• 99 plants (61 sites)• ~99,000 MWe, ~789,000 MW-hr (2013) = 19% U.S. total• Worldwide: 435 plants, 372 GWe capacity

More Electricty Fun Facts

• Generation– Modern NPP ~1000 MW (1 unit)– Saint-Gobain (coal) 5.6 MW– Brayton (coal) ~1400 MW– Burriville (natural gas) ~900 MW– Block Island (wind) ~30 MW– Cape Wind (wind) ~470 MW– Hoover Dam (hydro) ~2000 MW– Robert-Bourassa (hydro) ~5600 MW

• Consumption– 1000 MW: 1M homes– 17000 MW: U.S. data centers (2013)

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USNRC (10/15/2015)

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Regulatory Documents

• Regulations - http://www.nrc.gov/reading-rm/doc-collections/cfr/

• Regulatory Guide (RG) - http://www.nrc.gov/reading-rm/doc-collections/reg-guides/

• Standard Review Plan (SRP) -http://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr0800/

• NUREG Series Reports - http://www.nrc.gov/reading-rm/doc-collections/nuregs/

• Policy Statements - http://www.nrc.gov/reading-rm/doc-collections/commission/policy/

• Inspection Manual - http://www.nrc.gov/reading-rm/doc-collections/insp-manual/

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Regulatory Documents - Examples

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Regulation

• 10 CFR 50, Appendix A, Criterion 2• Structures, systems, and components important to safety shall be

designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seicheswithout loss of capability to perform their safety functions.

RG

• RG 1.76, “Design-Basis Tornado and Tornado Missiles for Nuclear Power Plants”• NUREG/CR-4461, “Tornado Climatology of the Contiguous United States,”

• RG1.221, “Design-Basis Hurricane and Hurricane Missiles for Nuclear Power Plants”• NUREG/CR-7004 Technical Basis for Regulatory Guidance on Design-Basis

Hurricane-Borne Missile Speeds for Nuclear Power Plants • NUREG/CR-7005 Technical Basis for Regulatory Guidance on Design-Basis

Hurricane Wind Speeds for Nuclear Power Plants

SRP

• Standard Review Plan Chapter 3.3.1, “Wind Loading”• Standard Review Plan Chapter 3.5.1.4, “Missiles Generated By

Tornadoes And Extreme Winds”

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General Design Criterion 35

Emergency core cooling. A system to provide abundant emergency core cooling shall be provided. The system safety function shall be to transfer heat from the reactor core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could interfere with continued effective core cooling is prevented and (2) clad metal-water reaction is limited to negligible amounts.

Suitable redundancy in components and features, and suitable interconnections, leak detection, isolation, and containment capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.

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Browns Ferry (March 22, 1975)

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Some “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)

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Risk Assessment vs. Risk Management

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From National Research Council, “Understanding Risk: Informing Decisions in a Democratic Society,” National Academy Press, 1996.

Uncertainties in PRA Results

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Core Damage Frequency – CDF (/ry)

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)

NRC Information

• Website: www.nrc.gov• Agencywide Document Access and Management

System (ADAMS): http://adams.nrc.gov/wba/• Jobs (USAJOBS): http://www.nrc.gov/about-

nrc/employment/apply.html• Status of Risk-Informed Activities: SECY-15-0135

(“Annual Update of the Risk-Informed Activities Public Web Site,” ADAMS ML15267A387, October 27, 2015)

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