mydocs.epri.com › docs › publicmeetingmaterials › 1111 › PZNPT4... EPRI/NRC-RES FIRE PRARES FIRE PRA …EPRI/NRC-RES FIRE PRARES FIRE PRA METHODOLOGY Module I-1: Fire Risk

EPRI/NRC RES FIRE PRAEPRI/NRC-RES FIRE PRA METHODOLOGYModule I-1: Fire Risk Requantification P j tProject

J.S. Hyslop – NRC/RESR.P. Kassawara – EPRI

Fire PRA Workshop 2011Fire PRA Workshop 2011San Diego CA and Jacksonville FLSan Diego CA and Jacksonville FL

A Collaboration of U.S. NRC Office of Nuclear Regulatory Research (RES) & Electric Power Research Institute (EPRI)

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General

• Based on MOU between NRC-RES and EPRI on fire risk

• Needed to provide more realistic methods for risk-informed, performance-based fire protection activities

• Scope is full power, CDF and LERF

• Course does not provide official NRC positions, but does represent the expertise of authors of NUREG/CR-6850 p p(EPRI 1011989)

Fire PRA Workshop 2011, San Diego & Jacksonville Fire PRA Workshop 2011, San Diego & Jacksonville Module Module II--1: Fire Risk 1: Fire Risk RequantificationRequantification ProjectProject

Slide Slide 22 A Collaboration of U.S. NRC Office of Nuclear Regulatory A Collaboration of U.S. NRC Office of Nuclear Regulatory Research (RES) & Electric Power Research Institute (EPRI)Research (RES) & Electric Power Research Institute (EPRI)

Quality Product

• Participation by industry and the public

• Formal process to resolve technical disputes within NUREG/CR-6850 (EPRI 1011989) team

• Improvements made in areas important to fire risk

• Fire PRA methodology still evolving

Solutions provided to fifteen fire PRA questions in NFPA 805– Solutions provided to fifteen fire PRA questions in NFPA 805 frequently asked questions (FAQ) process in NUREG/CR-6850 Supplement 1, EPRI 1019259



– “Unreviewed Analysis Methods” in process

Uses Of Methodology

• Support for 10CFR50.48(c) implementation– Plants using technology for fire PRA development/upgradePlants using technology for fire PRA development/upgrade

• ANS fire risk standard development– Typically defines state-of-art, although supports lesser capability

categories as wellcategories as well• Reactor Oversight Process analyses

– Refined phase 3 analyses– Development of phase 2 Fire Protection SDP (IMC 0609 Appendix F)Development of phase 2 Fire Protection SDP (IMC 0609, Appendix F)

• Other expected uses– Analyses under the current fire protection regulations (i.e.

exemptions/deviations or plant changes due to risk-informed technicalexemptions/deviations or plant changes due to risk informed technical specifications)



Fire Model Validation and Verification (V&V)

• Fire modeling is an integral part of fire PRA

• Fire model verification and validation (V&V) is required for NFPA 805 applications

• Most fire models are computational

• Some are based only on empirical correlationsSome are based only on empirical correlations – Address cases where computational fire models inadequate– Fill important gaps in fire PRA

• PRA Methodology document not a reference for fire models

• EPRI/RES V&V of fire models EPRI 1011999/NUREG 1824



• EPRI/RES V&V of fire models EPRI 1011999/NUREG-1824

Related Activities

• EPRI 1011989/NUREG/CR-6850 Publication 2005– Publication 2005

– General workshops 2005, 2006– Detailed training 2007-2010

• EPRI 1011999/NUREG-1824 2007• Fire PRA NFPA 805 FAQ Support 2010• Fire HRA Methodology Development Late 2011• Fire Modeling Application Guide Early 2012

Fi E t D t b 2012• Fire Events Database 2012• Fire Modeling Training Integrated

Low Power/Shutdown Fire PRA Methods NRCFire PRA Workshop 2011, San Diego & Jacksonville Fire PRA Workshop 2011, San Diego & Jacksonville Module Module II--1: Fire Risk 1: Fire Risk RequantificationRequantification ProjectProject


• Low Power/Shutdown Fire PRA Methods NRC

EPRI/NRC RES FIRE PRAEPRI/NRC-RES FIRE PRA METHODOLOGYIntroduction and Overview: the Fire PRA M h d l d C SMethodology and Course Structure

Steve Nowlen - Sandia National LaboratoriesSteve Nowlen - Sandia National LaboratoriesBijan Najafi - Science Applications International Corp.Joint RES/EPRI Fire PRA Training WorkshopSan Diego, CA, August 1-5, 2011 andJacksonville, FL, November 14-18, 2011


Jacksonville, FL, November 14 18, 2011

Overview of this morning’s presentations

I. A brief overview of the course structure, content and objectivesobjectives

II. User experience – industry and NRCIII. An overview of the ASME/ANS PRA StandardIII. An overview of the ASME/ANS PRA StandardIV. Ongoing projects and activities related to fire PRA

– Fire Event Database update activity– Supplemental methods development efforts and the NRC/industry

peer review panel– Ongoing efforts on fire-induced circuit failure mode and likelihood g g

analysis (recent testing, PIRT panel, PRA expert panel)

Fire PRA Training, 2011, San Diego CA and Jacksonville FLFire PRA Training, 2011, San Diego CA and Jacksonville FLIntroduction and OverviewIntroduction and Overview Slide Slide 22 A Collaboration of U.S. NRC Office of Nuclear Regulatory A Collaboration of U.S. NRC Office of Nuclear Regulatory

Research (RES) & Electric Power Research Institute (EPRI)Research (RES) & Electric Power Research Institute (EPRI)

PART IPART I

A brief Overview of the A brief Overview of the EPRI/NRCEPRI/NRC RES Fire PRA MethodologRES Fire PRA MethodologEPRI/NRCEPRI/NRC--RES Fire PRA MethodologyRES Fire PRA Methodology

&&

The Fire PRA CourseThe Fire PRA Course



Course Structure

• Five parallel modules:1. PRA/Systems Analysis2. Electrical Analysis3. Fire Analysis4. Fire PRA Human Reliability Analysis (HRA)5. Advanced Fire Modeling

• General structure for each module:– PowerPoint presentations covering key concepts and the general “how

” f h kto” of each task– Example problems to illustrate key elements

• Our intent is that you participate actively in the discussionsA k ti !– Ask questions!

– Instructors will both encourage and moderate discussions– They may also closure if discussion gets off-track (e.g., we cannot solve

your regulatory questions)



your regulatory questions)

OVERVIEW OF FIRE PRA PROCESS AND MODULE STRUCTURE

TASK 1: Plant Boundary & Partitioning

TASK 2: Fire PRA Component Selection

TASK 3: Fire PRA Cable Selection

TASK 4 Q lit ti S i TASK 5: Fire Induced RiskSUPPORT TASK A: Plant

W lk D TASK 4: Qualitative Screening

TASK 6: Fire Ignition Frequencies

TASK 5: Fire-Induced Risk Model

Walk Downs

TASK 7A: Quantitative Screening - I

TASK 8: Scoping Fire Modeling

SUPPORT TASK B: Fire PRA Database TASK 12A: Post-Fire HRA:

Screening

Fire Analysis Module

TASK 7B: Quantitative Screening - II

PRA/System Module

Circuits Module

HRA Module



B Fire Analysis and Fire Modeling Modules

OVERVIEW OF FIRE PRA PROCESS AND MODULE STRUCTURE (2)( )

Detailed Fire Scenario Analysis

B

TASK 11: Detailed Fire Modeling A. Single CompartmentB. Multi-Compartment C. Main Control Room

TASK 9: Detailed Circuit Failure Analysis

TASK 10: Circuit Failure Mode & Lik lih d A l iLikelihood Analysis

TASK 12B: Post fire HRA:TASK 14: Fire Risk Quantification

TASK 15: Uncertainty & Sensitivity Analyses

TASK 12B: Post fire HRA: Detailed & recoveryTASK 13: Seismic-Fire

Interactions

Fire Analysis Module

PRA/System ModuleSensitivity Analyses

TASK 16: Fire PRA Documentation

Circuits Module

HRA Module

Fire Analysis and Fire Modeling Modules



Module 1: PRA/Systems Analysis

• This module cover analysis elements related to plant systems accident• This module cover analysis elements related to plant systems accident response modeling, integration of human actions into the plant model, and risk quantification tasks

S ifi t k d• Specific tasks covered are:– Task 2: Equipment Selection– Task 4: Qualitative Screening– Task 5: Fire-Induced Risk Model– Task 7: Quantitative Screening– Task 15: Risk Quantification– Task 16: Uncertainty Analysis



Module 2: Electrical Analysis

• This module covers analysis elements related to the identification andThis module covers analysis elements related to the identification and tracing of cables, and the analysis of electrical circuit failure modes and likelihood

• Tasks covered are:– Task 3: Cable Selection (and Routing)– Task 9: Detailed Circuit Analysisy– Task 10: Failure Mode Likelihood Analysis– Support Task B: FPRA Database



Module 3: Fire Analysis

• This module covers analysis elements related to the identification, characterization and analysis of fires scenarios– fire sources, fire frequency, fire damage, and fire protection systems and

features,

• Tasks covered are:T k 1 Pl t P titi i– Task 1: Plant Partitioning

– Task 6: Fire Ignition Frequency– Task 8: Scoping Fire Modeling– Task 11: Detailed Fire Scenario Analysis– Task 13: Seismic/Fire Interactions (briefly)– Support Task A: Plant Walkdowns



Module 4: Fire PRA Human Reliability Analysis

• Second year for fire HRA module– Updated based on last year’s roll-out experience and refinements to the reference p y p

document (NUREG-1921, TR1019196, draft for comment)

• Covers fire PRA analysis elements related to the identification of human failure events (HFEs) and quantification of human error probabilities (HEPs)

• Tasks covered:– Task 12a: screening level HRAg– Task 12b: detailed HRA quantification

• Note: EPRI/RES guidance includes an intermediate “scoping” approachApproach more realistic than screening yet lacks detail and realism of– Approach more realistic than screening, yet lacks detail and realism of detailed HRA quantification

– Instructors will explain how this fits into overall analysis framework



Module 5: Advanced Fire Modeling (New!)

• Covers the use of computational fire models in support of fire scenario screening and analysisscreening and analysis

• Reference: RES/EPRI NPP Fire Modeling Application Guide (NUREG-1934, TR1023259, draft for comment)

• From handbook correlations to computational fluid dynamicsFrom handbook correlations to computational fluid dynamics…

• Fire modeling tools support:– Task 8: Scoping Fire Modeling

Task 11: Detailed Fire Scenario Analysis– Task 11: Detailed Fire Scenario Analysis

• Intended as a complement to the fire analysis module 3:– Module 3 focuses on the overall structure and content of the fire analysis elements

M d l 5 f th t l f t ti l fi d l– Module 5 focuses on the actual use of computational fire models



Another element that you will see in training is citations to the PRA Quality Standardy

• ASME/ANS RA-Sa–2009:– Addenda to ASME/ANS RA-S-2008: Standard for Level 1/Large EarlyAddenda to ASME/ANS RA S 2008: Standard for Level 1/Large Early

Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications

• The training slides provide links from elements of the methodology to g p gyrequirements of the standard– Not intended to “teach” the standard– Intended to provide a road map between the methods and the standard’s p p

Supporting Requirements• A separate presentation this morning will provide a high-level

introduction to the standard, it’s structure, and it’s placep



Any questions before we move on?Any questions before we move on?Any questions before we move on?Any questions before we move on?



EPRI/NRC-RES FIRE PRA METHODOLOGY AND ITS RELATIONSHIP TO NRC’s REGULATORY STRUCTURE

Mary DrouinU.S. Nuclear Regulatory Commission

Fire PRA Workshop 2011Fire PRA Workshop 2011San Diego CA and Jacksonville FLSan Diego CA and Jacksonville FL


How Does NUREG/CR-6850 Fit into the NRC Regulatory Structure?g y

• The objective here is to provide an understanding, j p g,from a regulatory perspective, the need for a fire probabilistic risk assessment (PRA) methodology d t d th f it l i th l tdocument, and therefore, its role in the regulatory structure.

• A major aspect of this objective is understanding what is meant by regulatory structure.what is meant by regulatory structure.

Fire PRA Workshop 2011, San Diego & Jacksonville Fire PRA Workshop 2011, San Diego & Jacksonville Fire PRA Methodology and its Relationship to NRC’S Regulatory Structure


NRC Regulatory Structure

• Congressional MandateAtomic Energy Act indicates that the mission of the NRC is to– Atomic Energy Act indicates that the mission of the NRC is to ensure that commercial nuclear power plants are operated in a manner that provides adequate protection of public health and safety and is consistent with the common defense and securitysafety and is consistent with the common defense and security.

• NRC provides for public health and safety via a licensing, oversight and enforcement process.

• Licensing, oversight and enforcement all involve establishing regulations and developing the necessary supporting structure (e g regulatory guides)supporting structure (e.g., regulatory guides).



What is the Relationship Between a Regulation and a Methodology Document?gy

REGULATIONS Defines technical requirements for the design and operation of the(e.g., 10 CFR Part 50) the design and operation of the

nuclear power plant

REGULATORY GUIDES

Provides one technical approach that is acceptable for meeting the

regulation

SUPPORTING TECHNICAL

Technical documents provide detailed guidance on issues

i t d ith ti thTECHNICAL GUIDANCE

associated with meeting the approach established in the

regulatory guide (or regulation)



What is the Relationship in the Context of a Fire PRA?

• Example relevant regulations:10 CFR §50 48(c) Fire Protection National Fire Protection Association– 10 CFR §50.48(c), Fire Protection, National Fire Protection Association Standard NFPA 805

– 10 CFR §50.69, Risk-informed categorization and treatment of structures, systems and components for nuclear power reactorssystems and components for nuclear power reactors

– 10 CFR §50.90, Application for amendment of license, construction permit, or early site permit

– 10 CFR §50.36, Technical Specifications10 CFR §50.36, Technical Specifications

• What is the common element among these regulations?– The use of risk information, and therefore, the need to have confidence

in the risk analyses (or PRAs) being used to generate the informationin the risk analyses (or PRAs) being used to generate the information– Risk contributors to be addressed include internal fires.



How is This Confidence Achieved?

REGULATIONS10 CFR §50 48(c) 10 CFR §50 69 10 CFR §50 90 10 CFR §50 36 etc10 CFR §50.48(c) 10 CFR §50.69 10 CFR §50.90 10 CFR §50.36 etc.

RG 1.205 RG 1.201 RG 1.174 RG 1.177

Guidance/staff position includes use of risk/PRA information

Reference to RG 1.200 for guidance on determining PRA technical acceptability for both internal and external hazards (e.g., internal fire)acceptability for both internal and external hazards (e.g., internal fire)

RG 1.200

“Describes one acceptable approach for determining whether the technical adequacy of the PRA, in total or the parts that are used to support an application, is sufficient to provide confidence in the results, such that the PRA can be used in regulatory decision-making for light-



water reactors.”

How is This Confidence Achieved (cont’d)?

• The approach provided in RG 1.200 defines the attributes and characteristics of a technically acceptable PRA.– The defined attributes and characteristics are very high level.

• For example, characteristics and attributes provided in RG 1.200 for Fire Ignition Frequencies:– Frequencies are established for ignition sources and– Frequencies are established for ignition sources and

consequently for physical analysis units.– Transient fires should be postulated for all physical analysis units

regardless of administrative controls.g– Appropriate justification must be provided to use nonnuclear

experience to determine fire ignition frequency.




• RG 1.200 allows the use of a consensus standard (as endorsed by the NRC) with a peer review to demonstrate conformance with the defined attributes and characteristics.– RG 1.200 endorses and provides a position on the ASME/ANS PRA p p

Standard (ASME/ANS RA-Sa-2009).– Part 4 of the ASME/ANS standard provides the requirements for

fires at-power PRA.

• The PRA Standard, however, only defines what is required for a technically acceptable PRA and an acceptable peer reviewreview.




• Guidance is needed for how to accomplish the requirements in the standard and guidance isrequirements in the standard and guidance is needed for the peer review in determining whether the intent of requirement is met.

• This guidance is particularly needed for those aspects in the PRA where the model is not well knownknown.

• One major objective of NUREG/CR-6850 is to provide the detailed guidance for how to p gaccomplish meeting the requirements for Fire PRA.– As such, NUREG/CR-6850 supports both the PRA

standard and the PRA peer review guidance



standard and the PRA peer review guidance

Overall Relationship

REGULATIONS( 10 CFR P t 50)

§50.48(c) §50.69 §50.90 §50.36 etc. (e.g., 10 CFR Part 50)

REGULATORY RG 1.205 RG 1.201 RG 1.174 RG 1.177 etc.

GUIDESRG 1.200

PRA Standard and Peer Review(ASME/ANS RA-Sa-2009 and NEI-07-12)

“WHAT TO” SUPPORTING TECHNICAL GUIDANCE

PRA Technical

Acceptability

DETAILED “HOW TO” SUPPORTING TECHNICAL

GUIDANCE

NUREG/CR-6850NUREG-1855NUREG-1842NUREG/CR-6823etc.



etc.

Structure of PRA Standard

Hazard Type • Two different hazard types are defined in the Standard which include both internal and external hazards.

Hazard groups• For each hazard type, the various groups of hazards are

defined, for example for internal hazards: internal events, internal flood and internal fires.

Technical Elements

Objectives

• For each hazard group, the technical elements are defined for developing the hazard-specific PRA model.

• The objective of each technical element is definedObjectives

High Level Requirements

(HLRs)

• The objective of each technical element is defined.

• High level requirements are defined which specify what is needed to accomplish the defined objective of the technical element(HLRs)

Supporting Requirements

technical element.

• Supporting requirements are defined which, in turn, specify what is needed to accomplish the defined high l l i t



(SRs) level requirement.

Structure of PRA Standard (cont’d)

• At the supporting requirement level, it was recognized that the extent or to which the level of detail, the level of plant specificity and the level of realism of the PRA model can vary

• Consequently, the minimum defined by the requirement can alsoConsequently, the minimum defined by the requirement can also vary

• Consequently, three “categories of capability” were defined for the supporting requirementssupporting requirements.

• Capability Category I: the very basic needed.

• Capability Category II: the minimum to support majority of applications

• Capability Category III: the minimum to support application with high reliance on the fidelity of the PRA results.



Standard Capability Categories

Capability Category I • Level of p y g yminimum addressed by the

technical requirementdetail of plant model

Pl t d i

Supporting Requirement

Capability Category II minimum to achieve

“current good practice” for the technical requirement

• Plant design and operational specificity

Specifies the minimum with regard to level of detail,

plant design and q

Capability Category III minimum to achieve the

• Realism of plant response

operational specificity, and realism of plant

response

state‐of‐the‐art for the technical requirement

p



Standard Capability Categories (cont’d)

PRA Attribute Capability Category I Capability Category II Capability Category III

Scope and Level of Detail

Sufficient to identify relative importance at the system

or train level

Sufficient to identify relative importance of significant

component level contributors

Sufficient to identify relative importance of all component level

contributors

G /Plant-Specificity

Generic data/models except for unique design and operational features

Plant-specific data/models for the significant contributors

Plant-specific data/models for all contributors

Realism Departures from realism ill h d t i t*

Departures from realism will h ll i t*

Departures for realism will h li ibl i t*will have moderate impact* have small impact* have negligible impact*

*Differentiation from moderate, to small, to negligible is determined by the degree that the conclusions and risk insights could affect a g gdecision under consideration.

• Moderate: likely that a decision could be affected• Small: unlikely a decision could be affected



• Negligible: a decision would not be affected.

Illustration of Fire PRA Standard Structure and NUREG/CR-6850

Hazard Type Internal Hazards External Hazards

Hazard groupsSeismic EventsHi h Wi d

Internal EventsHazard groups High Winds

External FloodsOther External Hazards

Technical

Internal FloodsInternal Fires

Plant Partitioning Fire Ignition FrequencyEquipment Selection Quantitative ScreeningCable Selection Circuit Failure

Fire PRA Elements

Objectives The objectives of the equipment selection elements is to select the plantequipment that will be included/credited in Fire PRA plant response model.

Cable Selection Circuit FailureQualitative Screening Human Reliability AnalysisPlant Response Model Risk QuantificationFire Scenario Selection Seismic/Fire Interactions

Standard

High Level Requirements

(HLRs)

HLR‐ES‐A The Fire PRA shall identify equipment whose failure . . . .HLR‐ES‐B The Fire PRA shall identify equipment whose failure including spurious operation

would adversely affect the operability/functionality of that portion of the plantdesign to be credited in the Fire PRA.

HLR‐ES‐C the Fire PRA shall identify instrumentation . . . .HLR‐ES‐D The Fire PRA shall document the Fire PRA equipment selection

Supporting Requirements

HLR‐ES‐D The Fire PRA shall document the Fire PRA equipment selection . . . .

ES‐B1 Capability Category I Capability Category II Capability Category III

IDENTIFY Fire Safe Shutdown/ IDENTIFY Fire Safe Shutdown/ IDENTIFY Fire Safe Shutdown/Appendix R equipment to be Appendix R equipment to be Appendix R equipment to becredited in the Fire PRA credited in the Fire PRA and credited in the Fire PRA and

NUREG/CR 6850



q(SRs)

credited in the Fire PRA. credited in the Fire PRA and credited in the Fire PRA andINCLUDE risk‐significant INCLUDE all equipment fromequipment from the IE PRA the IE PRA

CR-6850

Fire PRA Standard and NUREG/CR-6850: Illustration of the Mapping of HLRs & SRs to 6850

Technical element

HLR SR 6850/1011989 sections that

cover SR

Comments

ES A The Fire PRA shall identify equipment whose failure caused by an initiating fire including spurious y q p y g g poperation will contribute to or otherwise cause an initiating event.

1 2.5.3 2 3.5.3 Covered in “Cable Selection” chapter 3 2.5.3 4 2.5.1, 2.5.4

5 2 5 4 5 2.5.4 6 2.5.6 B The Fire PRA shall identify equipment whose failure including spurious operation would

adversely affect the operability/functionality of that portion of the plant design to be credited in the Fire PRA.

1 2.5.2 2 2.5.4 3 5.5.1 Covered in “Fire-Induced Risk Model” chapter 4 3.5.3 Covered in “Cable Selection” chapter 5 n/a Exclusion based on probability is not covered in 6850/1011989 C The Fire PRA shall identify instrumentation whose failure including spurious operation would

impact the reliability of operator actions associated with that portion of the plant design to be p y p p p gcredited in the Fire PRA.

1 2.5.5 2 2.5.5 D The Fire PRA shall document the Fire PRA equipment selection, including that information about

the equipment necessary to support the other Fire PRA tasks (e.g., equipment identification; equipment type; normal desired failed states of equipment; etc ) in a manner that facilitates Fire



equipment type; normal, desired, failed states of equipment; etc.) in a manner that facilitates Fire PRA applications, upgrades, and peer review.

1 n/a Documentation not covered in 6850/1011989

Overall Process

When implementing RG 1.200, PRA technical acceptability

involves:

ASME/ANS Developing a PRA in accordance to

the technical

Technical requirements in the PRA Standard

describe what needs to be in the PRA, but do not describe how to

RG

RA-Sa-2009

PRA

the technical requirements of the Standard (as

endorsed by NRC)

build the PRA model

NUREG/CR 6850

Provides guidance for one acceptable

methodology that canRG 1.200

PRA Model

Performing a peer review (as endorsed A i i i d t

NUREG/CR-6850 methodology that can be used for how to

develop PRA and meet the PRA Standard

NEI 07-12review (as endorsed by NRC) to assess

whether the technical requirements were

actually met

A peer review is required to evaluate how the PRA

model was built to determine whether it meets the intent of the technical requirements in

the PRA Standard



the PRA Standard

Summary/Conclusion

G/C• NUREG/CR-6850 is a methodology document and, while not required to be

fmet, plays a major role in defining a technically acceptable Fire PRA to support NRC i i i h Fi PRA d l iNRC activities where a Fire PRA model is needed and the results of the Fire PRA

d l d l imodel are used to meet a regulation.



FIRE PRA METHODS AND RESEARCH UPDATE 2011RESEARCH UPDATE - 2011

Rick WachowiakSr Project Manager

EPRIFire Fire PRA Workshop 2011PRA Workshop 2011

San Diego CA and Jacksonville FLSan Diego CA and Jacksonville FL

Fire PRA Workshop 2011, San Diego & Jacksonville Fire PRA Workshop 2011, San Diego & Jacksonville Fire PRA New Methods UpdateFire PRA New Methods Update


Contents

• Fire Events Database

• Heat Release Rates for Electrical Cabinets

Fi M th d E t R i P l• Fire Methods Expert Review Panel

• Fire Modeling Application Guideg pp



FEDB Purpose and Objectives

The Electric Power Research Institute’s Fire Events Data Base is the i i l f fi i id t ti l d t f i fi PRA

This project will improve the FEDB by:

principal source of fire incident operational data for use in fire PRAs

s p oject p o e t e by• Including expanded and improved data fields• Improving consistency and quality of information

I d fi t it l ifi ti• Improved fire event severity classificationThese improvements support fire PRA:• Updated, improved fire frequencies• Treatment of detection & suppression effectiveness• Estimates of damaging fire frequencies and their characteristics



Scope and Cooperation with NRC under MOU

• Update data through 2009, and beyond as availableP i i l d t• Principal data sources:

– Plant records, e.g. condition or corrective action reports supplemented with available

– LERs, ENs– NEIL, EPIX reports – Plant specific data collected for FPRAs

• Cooperative effort lead by EPRI with NRC under MOU– Database software upgrades– Data field content and definitionsData field content and definitions– Fire severity criteria– Beta testing

QA audit



– QA audit

Key Fire Event Data Base Features

• Event summary description

• Location and source characteristicsLocation and source characteristics

• Fire duration, growth, and damage descriptive details

• Detection– Time(s)– Systems & equipment– Fire brigade and other personnel role– Fire brigade and other personnel role

• Suppression– Time(s)– Systems & equipment– Fire brigade and other personnel role

• Fire severity classification graded dependent on magnitude and



• Fire severity classification graded, dependent on magnitude and impact details

Fire Event Severity Classifications

• The severity levels used in the updated FEDB are “challenging”, “potentially challenging” and “not challenging”potentially challenging , and not challenging

• The new “challenging” and “potentially challenging” classifications combined are essentially equivalent to the NUREG/CR-6850 y q“potentially challenging” classification

• A distinction made between “challenging” fires that did (or could) d iti l t b d th i iti d thdamage a critical component beyond the ignition source and those “potentially challenging” fires that might evolve into challenging fires in fire PRA model

• Initial classification using fire severity algorithm



Fire Event Severity Determination

• Applied to new and existing fire event data (1990 forward)

• Fire event severity classification algorithm:

– Logic model using FEDB data field entries to indicate initial fire severity classificationy

– Challenging → Potentially Challenging → Not Challenging

– Identifies key missing information needed to make classification on “undetermined” eventson “undetermined” events

• Fire event severity classification review & justification for exceptions requiredrequired

• Resolution of “undetermined” severity classifications incorporated in event data collection and coding process to extent practical



Event Derived/Inferred: Fire Severity

Event Classification Event Sub-Classification Criteria

CHALLENGING Damage to or ignition of an adjacent object occurred. This includes ignition

One of the following:

g g j j gof secondary combustibles.Damage to or ignition of an adjacent object could have occurred if the fire were in a different location.Damage to or ignition of an adjacent object or component could haveDamage to or ignition of an adjacent object or component could have occurred if significant suppression actions had not been taken.

POTENTIALLYCHALLENGING

Damage to or ignition of an adjacent object could have occurred if minor suppression actions were not taken in a timely manner.

Not “challenging” andone of the following:

Damage to or ignition of an adjacent object could have occurred if the fire were in a different location and if minor suppression actions were not taken in a timely manner.

NOT CHALLENGING Overheat condition only; no smoldering or flaming combustion.

Not “potentially challenging” and one of the following:

Smoldering fire self-extinguishes without any active intervention.

Fire involves an ignition source in a location that has no relevance to plant operations or safety.

UNDETERMINED A fi t th t d t h ffi i t i f ti t l if



UNDETERMINED Any fire event that does not have sufficient information to classify as challenging, potentially challenging, or not challenging.

Fire Events Database Progress

Request CR “List” for Key Words (2 000 4 000 / plant)

1n = input from utilities required

(2,000 - 4,000 / plant)

Owner’s Groups Perform Screening Review

Request CRs for Probable Pl t di t t

m = Audit (NRC)

1

I t t

Request CRs for Probable Fires

(75 - 125 / plant)

EPRI Screen for Potential Significance

2 Plants responding to requests

Expect completion in 2011

2Important Existing FEDB

Fire Events(0 – 5 / plant)

Site Follow-up for “Real Fires”

(10 – 15 / plant)

3

2

Events in New FEDB

Potentially Challenging and Challenging Fires

EPRI CategorizeImportant

Fires(0-3/plant)

33rd audit in Nov 2011



Challenging Fires

Fire Ignition Frequency (Re)Analysis:Purpose, Objective, Scopep , j , p

• Re-examine and update fire event trends and fire ignition frequencies developed in EPRI interim Report 1016735 FAQ 08-0048developed in EPRI interim Report 1016735, FAQ 08-0048

• Use data from the FEDB updated through 2009

• Improved methods to address statistical issues/concerns for low density fire ignition bins

R lt t b it bl f i l t ifi fi PRA li ti• Results to be suitable for use in plant specific fire PRA applications

• Includes coordination with the NRC, industry technical review

• Preparation of EPRI technical reports



Fire Ignition Frequency (Re)Analysis: Technical Issues and Approach

• Based on data from 1990-2000, between plant variability detected– Strong indication for aggregated data, high density fire ignition binsStrong indication for aggregated data, high density fire ignition bins – Low density fire ignition bins indeterminate – Approach assumes between plant variability

• Plan to apply hierarchical Bayes treatment for all fire ignition bins; draft methodology report prepared and in review

• Uncertainty bounds & plant specific sensitivity will generally increase while maintaining generic “results” for mean fire ignition frequencies

• Industry wide trends and any implications unknown until new data is available for analysis

• Use of pre 1990 data under discussion between RES and EPRI



• Use of pre-1990 data under discussion between RES and EPRI

Fire Ignition Frequency (Re)Analysis: Status and Schedule

• Methodology report for technical review with NRC under MOU: Comments received 2nd Qtr 2011Comments received 2 Qtr 2011

• Draft EPRI Interim Technical Report on methodology for industry technical review: 4th Qtr 2011

• Draft EPRI Technical Report on updated FEDB fire ignition frequency (re)analysis: 1st Qtr 2012

• Final EPRI Technical Report (methodology and data analysis): Mid 2012



Vertical Electrical Cabinet Heat Release RateProposed Methodology

• Purpose of study: Re-evaluate the heat release rates (HRRs) of cabinet fires recommended for use in NUREG/CR 6850 (Table G 1)cabinet fires recommended for use in NUREG/CR-6850 (Table G-1)

• Scope limited to:– Vertical electrical cabinets– No quantification of external influences– No quantification of fire duration

• Model developed to characterize• Model developed to characterizepotential HRR for cabinet fires

– Cabinet size and combustible load– Ventilation characteristics– Types of combustibles



Vertical Electrical Cabinet Heat Release Rate ModelProposed Methodology

• Model maintains distinction between qualified cable and non-qualified cable fires

• Establish probability that fire in a cabinet with only qualified cables may remain localized

• Model adds refinements and thus more inputs needed than in NUREG/CR-6850:6850:– If not ventilation-limited: HRR dependent on combustible loading.

Combustible loading estimated based on:• Cabinet volume• Estimate of cabinet fullness (low/high loading).

– If ventilation-limited: HRR controlled by amount of air available for combustion.• Determination on whether cabinet is cooled by forced or naturalDetermination on whether cabinet is cooled by forced or natural

ventilation• Assessment of robustness regarding gap potential• Need to know vent area and position of vents (top, bottom, or vents at

i t di t l l)



intermediate level)

Vertical Electrical Cabinet Heat Release RateProposed Methodology

Current StatusI d d t t i l• Independent expert review panel– Fire modeling and fire PRA practitioners– Mix of contractors and utilityy– Verify that the method is correct, has proper bases, and is

useable

• Comments received and are being resolved• Comments received and are being resolved• Expected to be published in 2011• NRC requested formal review by NRC-RESNRC requested formal review by NRC RES



FPRA Unreviewed Analysis Methods Review Panel

• NEI 07-12 allows Fire PRA peer reviews to identify “unreviewed analysis methods”unreviewed analysis methods

• Panel process established to provide a review for any new methods– 6 to 10 members – Utilities, EPRI, Owners’ Groups, NRC– Independent from developer and user(s) of method(s)

Review proposed method and provide comments– Review proposed method and provide comments– Determine the applicability / limitations of the method– Identify interface with the Fire PRA Standard

• EPRI to “publish” approved methods via letter to NEI– Eventually provided as a collection in an EPRI report



Panel Status

• ERIN Engineering and Research, Inc. submitted four methods –Alignment factors for:g– Electrical cabinet fires– Pump oil fires– Transient firesTransient fires– Hot work fires

• Each method attempts to add realism to a fire PRAProposition is that fires included in the generic frequencies do not– Proposition is that fires included in the generic frequencies do not generally behave as aggressively as the Fire PRA treats them

• Review of methods as submitted is complete and there are significant commentscomments

• ERIN is resolving comments• Each of the methods will need to be re-reviewed following update of

h d i



the documentation

Insights From the Panel Process

• Methods that treat all scenarios the same are more difficult to justifymore difficult to justify– Tie method to scenario specific

information– Specify information needed to justify m

in M

etho

d

Specify information needed to justify applicability

• Methods relying on expert judgment may put the panel into a different role on

serv

atis

m

may put the panel into a different role– Expert panel vs. independent review– Different process may be needed when in this role

S i ifi l d d

Co

Generic Applicability

• Scenario specific examples needed• Involve the author earlier in the process• Review one method at a time



Review one method at a time

Fire Modeling Application Guide

• Joint EPRI / NRC RES publication under MOUD ib th f d ti fi d li• Describes the process of conducting a fire modeling analysis for commercial nuclear power plant applications

• Provides guidance on incorporating the uncertaintyProvides guidance on incorporating the uncertainty associated with a fire modeling analysis in a Fire PRA

• Five fire models discussed– Two empirical, two zone, one CFD

• Eight “real world” examples providedApplicability of models and data ranges– Applicability of models and data ranges

• Draft is out for public comment– This forms the bases for Module 5 – Advanced Fire Modeling



g

E h d Fi E t D t b tEnhanced Fire Events Database to Support Fire PRA(A joint NRC-RES/EPRI project)( j p j )

Given by J.S. Hyslop, NRC/RES

Joint RES/EPRI Fire PRA WorkshopAugust and November 2011S Di CA


San Diego, CA

Motivation

• Need for improved database identified by NFPA 805 FAQ programprogram – FAQ 48 required an update of current fire events database to

include 2001-2009 data address fire ignition frequency

• Updated, improved database will also serve more general fire PRA needs of both NRC and industry

Fire PRA Workshop, Fire PRA Workshop, 2010, San Diego & Jacksonville2010, San Diego & JacksonvilleEnhanced Fire Events DatabaseEnhanced Fire Events Database


Participants and Roles

• Joint project by RES and EPRIRES and EPRI collaborated on data fields and criteria for severity– RES and EPRI collaborated on data fields and criteria for severity classification

– EPRI taking the lead in collecting data and populating database, with RES in an audit rolewith RES in an audit role

• SNL and INL under contract to RES; ERIN lead contractor to EPRI

• Plan for a joint RES/EPRI report to be issued on project in 2012



Improvements

• Expanded and improved data fieldsImproved Fire Ignition Source Details– Improved Fire Ignition Source Details

– Improved Fire Detection and Suppression Response Details– Improved Fire Event Severity Classification

• More complete data– Update expected to collect more data than currently exists for total

1968-2000 period of existing database1968-2000 period of existing database

• Expanded fields will capture important information relevant to analyses unavailable in existing version.

• Each individual event description will also be more completeR d d t f “ d t i d” d t



• Reduced percentage of “undetermined” data

Improvements (cont.)

• Severity classification for applying database to frequency expanded to be more transparent and reproducibleexpanded to be more transparent and reproducible, although some judgment remains– Generic rules to eliminate “false positives” for events under the

it l ifi ti t dseverity classification generated

• Improving Consistency and Quality Assurance• Allowing for Reference Data Source Traceability• Allowing for Reference Data Source Traceability



Event Derived/Inferred :Fire Severityy

• Four severity classificationsChallenging– Challenging

– Potentially Challenging– Not Challenging– Undetermined

• For challenging, the fire is further developed than for potentially challengingpotentially challenging– Potentially challenging fires evolve into challenging fires in fire

PRA model

• Thus, no PRA applications distinguish between potentially challenging and challenging fires



Applications

• Initial application will be fire ignition frequencyA ti l i li ti ill b i• An essential companion application will be suppression reliability– Some very minor fires kept in earlier database will be removed in y p

update

• Expect true component frequencies will be developed once component population data collected by Owners’once component population data collected by Owners Groups (later than 2012 schedule cited earlier)

• Further qualitative insights on fires to be gained g g



EPRI/NRC RES FIRE PRAEPRI/NRC-RES FIRE PRAMETHODOLOGY

dc Circuits Project and Related Activities

Gabriel Taylor – U.S. NRC U. S. NRC Office of Nuclear Regulatory Research

J i t RES/EPRI Fi PRA T i i W k hJoint RES/EPRI Fire PRA Training WorkshopSan Diego, CA, August 2011 andJacksonville, FL, November 2011


History of Fire & Circuit ProblemsFailure, Protection, Testing

• RG 1.75, “Physical Independence of Electrical Systems”B F Fi 1975• Browns Ferry Fire 1975

• IEEE 383, “Qualifying Class 1E Electrical Cables”• IN 99 17 “Post Fire Safe Shutdown Circuit Analysis Problems”• IN 99-17, Post-Fire Safe-Shutdown Circuit Analysis Problems

• NEI Cable Testing 2001 & Expert Elicitation 2002• RIS 2004-03 “Risk-Informed Approach for Post-Fire Safe-RIS 2004 03, Risk Informed Approach for Post Fire Safe

Shutdown Circuit Inspections”• NRC CAROLFIRE – RIS 2004-03 Bin 2 & Fire Modeling• DUKE Armored Cable Testing• NRC DESIREE-FIRE – direct current circuits

Fire PRA Workshop, 2011Fire PRA Workshop, 2011Fire PRA Circuit Analysis Fire PRA Circuit Analysis ActivitiesActivities


• KATE-FIRE – Kerite Thermal Fragility Evaluation

NEI / EPRI Testing 2001

• Square Steel Room• Diffusion Burner• Diffusion Burner• Cable Tray &

Conduit Bend• Sprinkler Head• Hot Gas & Plume

Exposures• 18 Tests using MOV

CircuitCircuit



CAROLFIRE Cable Response to Live Fire

• 78 Small- and 18 Intermediate-scale tests• ac MOV and Insulation Resistance Measurement

Systems used for failure mode diagnosticRIS 2004 03 & Fire Model Improvement• RIS 2004-03 & Fire Model Improvement

Penlight Shroud

Cable Tray & Test Cables

A

A

A-A

Cable Tray Supports

A A-A0.81 m (2'-8") 0.51 m (1'-8 1/4")



DESIREE-FIREDirect Current Electrical Shorting In Response to Exposure Fire

• Similar thermal exposure methods as CAROLFIRE• 59 small- and 17 intermediate scale tests

– 225 individual circuit trials

Ins lation resistance meas rement as not sed• Insulation resistance measurement was not used• Several different dc control circuits simulated NPP circuits

and monitored cable failure modes.and monitored cable failure modes.– dc MOV– Medium Voltage Switchgear (15kV / 4.16kV)

Pil SOV– Pilot SOV– 1” SOV assembly– Large coil



g

Diagnostic Units

PA

DCCCS CABLE UNDER TESTFROM DISTRIBUTION CIRCUIT BREAKER

FUSE BLOCK

G

PA

V

V

1750

NCA

C

YC1

R

A

V

VNC

+

VDC

NO

C

A

1750

10A

A

B

YO1

NA

A

VNC-

125

V

O

10A

A

SP

V

V



DESIREE-FIRES



CAROLFIRE ac Results – RIS 2004-03

• Inter-cable shorting less likely than Intra-• Inter-cable between thermoset and thermoplastic is very small compared to either common type

•No basis on number of failures to consider•No basis for taking credit of CPT•Very low probability of ac hot shorts >20 minutes in duration

•THIEF fire model



Cable Damage Prediction Model – THIEF

• Thermally-Induced Electrical Failure (THIEF)• Provides a numerical solution of the one dimensional heat• Provides a numerical solution of the one-dimensional heat

conduction equations within a homogenous cylinder with fixed temperature-independent properties

• Incorporated in CFAST and FDS fire models• User specifies cable failure temperature



Electrical PIRT

• Phenomena Identification and Ranking Table• Structured approach to an expert elicitation• Structured approach to an expert elicitation• Facilitated by Brookhaven National Laboratories• Group of 8 electrical experts• Group of 8 electrical experts

– 4 NRC, National Laboratories– 4 EPRI, Utilities, Contractors



PIRT Objectives

• Review test data, identify important phenomena that ff t bl / i it f il d d th k thaffects cable/circuit failure mode, and then rank the

phenomena and important influencing parameters.• Consensus on longstanding post-fire safe-shutdownConsensus on longstanding post fire safe shutdown

circuit issues.– MHIF, 3 phase proper polarity, CTs, etc.



PIRT Work

• PIRT panel identified lack of readily usable information from multiple testing reportsfrom multiple testing reports

• NRC has initiative to provide a compilation NUREG document of previous test results

• Focus on evaluating the data based on the key phenomena and parameters identified by PIRT panel.



Data Report Example – Failure Mode

80

Global Approach - Insultion Type

100%

60

52%

62%

100%40ount Fuse Clear

Spurious Actuation

44%

38%

51%56%

20

Co Spurious Actuation

Hot Short

HS/SA Possible

44%

0

20



0TP TS

Data Report Example – Raceway Routing Duration

Duration - Raceway Routing

300

400

MAX = 1345 MAX = 456

200

Seco

nds

q1

min

median

100

max

q3

0AIR CONDUIT TRAY AIR CONDUIT TRAY

Hot Short Spurious Actuation



PRA Expert Panel

• Electrical PIRT Panel results will be used by a follow-on group of fire PRA expertsgroup of fire PRA experts

• PRA panel will revise EPRI best-estimate conditional probabilities for circuit hot short/spurious actuations used p pin fire PRA methods, along with developing new probabilities.



dc Circuits Project and Related Activities Questions

Questions ?Questions ?



EPRI/NRC-RES Fire PRA Course

Revision Date: June 21, 2011

Electric Power Research Institute (EPRI) Division of Risk Analysis 3412 Hillview Avenue Office of Nuclear Regulatory Research (RES) Palo Alto, CA 94303 U.S. Nuclear Regulatory Commission Washington, DC 20555

2

iii

PREPARERS

TECHNICAL TEAM LEADS:

Bijan Najafi

Science Applications International Corp. 1671 Dell Ave, Suite 100 Campbell, CA 95008

Steven P. Nowlen

Sandia National Laboratories (SNL) PO Box 5800 Albuquerque, NM 87185-0748

PROJECT MANAGERS:

Richard Wachowiak

Electric Power Research Institute

J. S. Hyslop

U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research Fire Research Branch

v

CONTENTS

1 INTRODUCTION .................................................................................................................... 1-1

1.1 Background .................................................................................................................. 1-1

1.2 How to Use this Package ............................................................................................. 1-4

1.3 References .................................................................................................................. 1-4

2 EXAMPLE CASE PLANT - GENERAL INFORMATION ......................................................... 2-1

2.1 Overall Plant Description ............................................................................................. 2-1

2.2 Systems Description .................................................................................................... 2-1

2.2.1 Primary Coolant System .......................................................................................... 2-1

2.2.2 Chemical Volume Control and High Pressure Injection Systems ............................ 2-2

2.2.4 Residual Heat Removal System .............................................................................. 2-3

2.2.5 Auxiliary Feedwater System .................................................................................... 2-3

2.2.6 Electrical System ..................................................................................................... 2-5

2.2.7 Other Systems ......................................................................................................... 2-5

2.3 Plant Layout ................................................................................................................. 2-6

2.4 SNPP Drawings.............................................................................................................. 2-6

3 MODULE 1: PRA/SYSTEMS .................................................................................................. 3-1

4 MODULE 2: ELECTRICAL ANALYSIS ................................................................................... 4-1

5 MODULE 3: FIRE ANALYSIS ................................................................................................. 5-1

6 MODULE 4: FIRE PRA HUMAN RELIABILITY ANALYSIS .................................................... 6-1

7 MODULE 5: ADVANCED FIRE MODELING .......................................................................... 7-2

vii

LIST OF ACRONYMS

AFW Auxiliary Feedwater ATWS Anticipated Transient Without Scram BWR Boiling Water Reactor CCDP Conditional Core Damage Probability CF Cable (Configuration) Factors CCW Component Cooling Water CDF Core Damage Frequency CFD Computational Fluid Dynamics CFR Code of Federal Regulations CLERP Conditional Large Early Release Probability CM Corrective Maintenance CRS Cable and Raceway (Database) System CVCS Chemical and Volume Control System EDG Emergency Diesel Generator EF Error Factor EOP Emergency Operating Procedure EPR Ethylene-Propylene Rubber EPRI Electric Power Research Institute FEDB Fire Events Database FEP Fire Emergency Procedure FHA Fire Hazards Analysis FIVE Fire-Induced Vulnerability Evaluation (EPRI TR 100370) FMRC Factory Mutual Research Corporation FPRAIG Fire PRA Implementation Guide (EPRI TR 105928) FRSS Fire Risk Scoping Study (NUREG/CR-5088) FSAR Final Safety Analysis Report HEAF High Energy Arcing Fault HEP Human Error Probability HFE Human Failure Event HPI High Pressure Injection HPCI High Pressure Coolant Injection HRA Human Reliability Analysis HRR Heat Release Rate HVAC Heating, Ventilation, and Air Conditioning ICDP Incremental Core Damage Probability ILERP Incremental Large Early Release Probability

viii

IPE Individual Plant Examination IPEEE Individual Plant Examination of External Events IS Ignition Source ISLOCA Interfacing Systems Loss of Coolant Accident KS Key Switch LERF Large Early Release Frequency LFL Lower Flammability Limit LOC Loss of Control LOCA Loss of Coolant Accident MCC Motor Control Center MCR Main Control Room MG Motor-Generator MOV Motor Operated Valve MQH McCaffrey, Quintiere and Harkleroad’s Method MS Main Steam NC No Consequence NEI Nuclear Energy Institute NEIL Nuclear Electric Insurance Limited NFPA National Fire Protection Association NPP Nuclear Power Plant NPSH Net Positive Suction Head NQ cable Non-Qualified (IEEE-383) cable NRC Nuclear Regulatory Commission P&ID Piping and Instrumentation Diagram PE Polyethylene PM Preventive Maintenance PMMA Polymethyl Methacrylate PORV Power Operated Relief Valve PRA Probabilistic Risk Assessment PSF Performance Shaping Factor PVC Polyvinyl Chloride PWR Pressurized Water Reactor Q cable Qualified (IEEE-383) cable RCP Reactor Coolant Pump RCS Reactor Coolant System RDAT Computer program for Bayesian analysis RES The Office of Nuclear Regulatory Research (at NRC) RHR Residual Heat Removal RPS Reactor Protection System RWST Refueling Water Storage Tank SDP Significance Determination Process SGTR Steam Generator Tube Rupture SI Safety Injection SO Spurious Operation SOV Solenoid Operated Valve SRV Safety Relief Valve

ix

SSD Safe Shutdown SSEL Safe Shutdown Equipment List SUT Start-up Transformer T/G Turbine/Generator TGB Turbine-Generator Building TSP Transfer Switch Panel UAT Unit Auxiliary Transformer VCT Volume Control Tank VTT Valtion Teknillinen Tutkimuskeskus (Technical Research Centre of Finland) XLPE Cross-Linked Polyethylene ZOI Zone of Influence

1-1

1 INTRODUCTION

1.1 Background

The U.S. Nuclear Regulatory Commission and Electric Power Research Institute under a Memorandum of Understanding (MOU) on Cooperative Nuclear Safety Research have been developing state of the art methods for conduct of fire PRA. In September 2005, this work produced the “EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities,” EPRI 1011989, and NUREG/CR-6850 [1].

A Fire PRA Course has been put together to train interested parties in the application of this methodology. The Course/Seminar is provided in five parallel modules. The first three modules are based directly on Reference [1]. However, that document did not cover fire human reliability analysis (HRA) methods in detail. For 2010, the training materials were enhanced to include a fourth module based on a more recent EPRI/RES collaboration and a draft guidance document, EPRI 1019196, NUREG-1921 [2] published in late 2009. The training materials are based on this draft document including the consideration of public comments received on the draft report and the team’s responses to those comments. For 2011 a fifth training module on Advanced Fire Modeling techniques and concepts has been added to the course. This module is based on the another joint RES/EPRI collaboration and a draft guidance published in January 2010, EPRI 1019195, NUREG-1934 [3].

The four training modules are:

Module 1: PRA/Systems Analysis - This module covers the technical tasks for development of the system response to a fire including human failure events. Specifically, this module covers Tasks/Sections 2, 4, 5, 7, 14, and 15 of Reference [1].

Module 2: Electrical Analysis – This module covers the technical tasks for analysis of electrical failures as the result of a fire. Specifically, this module covers Tasks/Sections 3, 9, and 10 of Reference [1].

Module 3: Fire Analysis – This module covers technical tasks involved in development of fire scenarios from initiation to target (e.g., cable) impact. Specifically, this module covers Tasks/Sections 1, 6, 8, 11, and 13 of Reference [1].

Module 4: Fire Human Reliability Analysis: This module covers the technical tasks associated with identifying and analyzing operator actions and performance during a postulated fire scenario. Specifically, this module covers Task 12 as outlined in Reference [1] based on the application of the approaches documented in Reference [2].

1-2

Module 5: Advanced Fire Modeling: This module is new for the 2011 training course and covers the fundamentals of fire science and provided practical implementation guidance for the application of fire modeling in support of a fire PRA. Module 5 covers fire modeling applications for Tasks 8 and 11 as outlined in Reference [1] based on the material presented in Reference [3].

Integral to Modules 1, 2 and 3 is a set of hands-on problems based on a fictitious, simplified nuclear power plant. The same power plant is used in all three modules. This document provides the background information for the problem sets of each module. Clearly, the power plant defined in this package is an extremely simplified one that in many cases does not meet any regulatory requirements or good engineering practices. Design features presented are focused on bringing forward the various aspects of the Fire PRA methodology. This package includes a general description of the power plant and the internal events PRA needed as input to the Fire PRA.

For Module 4 and 5, independent sets of examples are used to illustrate key points of the analysis procedures. The examples for these two modules are not tied to the simplified plant. Module 4 uses examples that were derived based largely on pilot applications of the proposed fire HRA methods and on independent work of the EPRI and RES HRA teams. The examples for Module 5 were taken directly from Reference [3] and represent a range of typical NPP fire scenarios across a range of complexity and that highlight some of the computation challenges associated with the NPP fire PRA fire modeling applications.

The instruction package for specific technical tasks is provided in Sections 3, 4, 5 and 6 which are organized by Modules (see above). A short description of the Fire PRA technical tasks is provided below. For further details, refer to the individual task descriptions in EPRI 1011989, NUREG/CR-6850, Volume 2. The figure presented at the end of this chapter provides a simplified flow chart for the analysis process and indicates which training module covers each of the analysis tasks.

Plant Boundary Definition and Partitioning (Task 1). The first step in a Fire PRA is to define the physical boundary of the analysis, and to divide the area within that boundary into analysis compartments.

Fire PRA Component Selection (Task 2). The selection of components that are to be credited for plant shutdown following a fire is a critical step in any Fire PRA. Components selected would generally include many, but not necessarily all components credited in the 10 CFR 50 Appendix R post-fire SSD analysis. Additional components will likely be selected, potentially including most but not all components credited in the plant’s internal events PRA. Also, the proposed methodology would likely introduce components beyond either the 10 CFR 50 Appendix R list or the internal events PRA model. Such components are often of interest due to considerations of multiple spurious actuations that may threaten the credited functions and components; as well as due to concerns about fire effects on instrumentation used by the plant crew to respond to the event.

Fire PRA Cable Selection (Task 3). This task provides instructions and technical considerations associated with identifying cables supporting those components selected in Task 2. In previous Fire PRA methods (such as EPRI FIVE and Fire PRA Implementation

1-3

Guide) this task was relegated to the SSD analysis and its associated databases. This document offers a more structured set of rules for selection of cables.

Qualitative Screening (Task 4). This task identifies fire analysis compartments that can be shown to have little or no risk significance without quantitative analysis. Fire compartments may be screened out if they contain no components or cables identified in Tasks 2 and 3, and if they cannot lead to a plant trip due to either plant procedures, an automatic trip signal, or technical specification requirements.

Plant Fire-Induced Risk Model (Task 5). This task discusses steps for the development of a logic model that reflects plant response following a fire. Specific instructions have been provided for treatment of fire-specific procedures or preplans. These procedures may impact availability of functions and components, or include fire-specific operator actions (e.g., self-induced-station-blackout).

Fire Ignition Frequency (Task 6). This task describes the approach to develop frequency estimates for fire compartments and scenarios. Significant changes from the EPRI FIVE method have been made in this task. The changes generally relate to use of challenging events, considerations associated with data quality, and increased use of a fully component-based ignition frequency model (as opposed to the location/component-based model used, for example, in FIVE).

Quantitative Screening (Task 7). A Fire PRA allows the screening of fire compartments and scenarios based on their contribution to fire risk. This approach considers the cumulative risk associated with the screened compartments (i.e., the ones not retained for detailed analysis) to ensure that a true estimate of fire risk profile (as opposed to vulnerability) is obtained.

Scoping Fire Modeling (Task 8). This step provides simple rules to define and screen fire ignition sources (and therefore fire scenarios) in an unscreened fire compartment.

Detailed Circuit Failure Analysis (Task 9). This task provides an approach and technical considerations for identifying how the failure of specific cables will impact the components included in the Fire PRA SSD plant response model.

Circuit Failure Mode Likelihood Analysis (Task 10). This task considers the relative likelihood of various circuit failure modes. This added level of resolution may be a desired option for those fire scenarios that are significant contributors to the risk. The methodology provided in this document benefits from the knowledge gained from the tests performed in response to the circuit failure issue.

Detailed Fire Modeling (Task 11). This task describes the method to examine the consequences of a fire. This includes consideration of scenarios involving single compartments, multiple fire compartments, and the main control room. Factors considered include initial fire characteristics, fire growth in a fire compartment or across fire compartments, detection and suppression, electrical raceway fire barrier systems, and damage from heat and smoke. Special consideration is given to turbine generator (T/G) fires, hydrogen fires, high-energy arcing faults, cable fires, and main control board (MCB) fires. There are considerable improvements in the method for this task over the EPRI FIVE and Fire PRA Implementation Guide in nearly all technical areas.

1-4

Post-Fire Human Reliability Analysis (Task 12). This task considers operator actions for manipulation of plant components. The analysis task procedure provides structured instructions for identification and inclusion of these actions in the Fire PRA. The procedure also provides instructions for incorporating human error probabilities (HEPs) into the fire PRA analysis. (Note that NUREG/CR-6850, EPRI 1011989 did not develop a detailed fire HRA methodology. Fire-specific HRA guidance can be found in NUREG-1921, EPRI 1019196, EPRI/NRC-RES Fire Human Reliability Analysis Guidelines – Draft Report for Comment, November 2009. Publication of the final Fire HRA report remains pending.)

Seismic Fire Interactions (Task 13). This task is a qualitative approach to help identify the risk from any potential interactions between an earthquake and fire.

Fire Risk Quantification (Task 14). The task summarizes what is to be done for quantification of the fire risk results.

Uncertainty and Sensitivity Analyses (Task 15). This task describes the approach to follow for identifying and treating uncertainties throughout the Fire PRA process. The treatment may vary from quantitative estimation and propagation of uncertainties where possible (e.g., in fire frequency and non-suppression probability) to identification of sources without quantitative estimation. The treatment may also include one-at-a-time variation of individual parameter values or modeling approaches to determine the effect on the overall fire risk (sensitivity analysis).

1.2 How to Use this Package

This package is intended to provide the background information necessary to perform some of the problem sets of the Course/Seminar. Please note:

1. All Course/Seminar attendees are expected to review Section 2 of this document and become familiar with the power plant defined in that section.

2. The instructors of each module will provide questions or case study problem sets and will guide the attendees to sections relevant to each specific problem set. Attendees will be expected to review those relevant sections and use the information or examples provided in those sections to complete the assigned problem set.

3. Do not make any additional assumptions in terms of equipment, systems, or plant layout other than those presented in the problem package without consulting the instructor.

1.3 References 1. EPRI 1011989, NUREG/CR-6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear

Power Facilities, September 2005. 2. EPRI 1019196, NUREG-1921, EPRI/NRC-RES Fire Human Reliability Analysis Guidelines

– Draft Report for Comment, Technical Update, November 2009.

3. EPRI 1019195, NUREG-1934, Nuclear Power Plant Fire Modeling Application Guide – Draft Report for Comment, January 2010.

1-5

1-6

2-1

2 EXAMPLE CASE PLANT - GENERAL INFORMATION

2.1 Overall Plant Description

This chapter provides background information about the fictitious plant used in the hands-on problem sets of Modules 1, 2 and 3. Note that the examples used in Module 4 (HRA) are not based on the example case plant.

The following notes generally describe the example case plant, including its layout:

1. The plant is a Pressurized Water Reactor (PWR) consisting of one Primary Coolant Loop, which consists of one Steam Generator, one Reactor Coolant Pump and the Pressurizer. A Chemical Volume Control System and multiple train High Pressure Injection system, as well as a single train Residual Heat Removal system interface with the primary system

2. The secondary side of the plant contains a Main Steam and Feedwater loop associated with the single Steam Generator, and a multiple train Auxiliary Feedwater System to provide decay heat removal.

3. The operating conditions and parameters of this plant are similar to that of a typical PWR. For example, the primary side runs at about 2,200 psi pressure. The steam generator can reject the decay heat after a reactor trip. There is a possibility for feed and bleed.

4. It is assumed that the reactor is initially at 100% power. 5. The plant is laid out in accordance with Figures 1 through 9. The plant consists of a

Containment Building, Auxiliary Building, Turbine Building, Diesel Generator Building and the Yard. All other buildings and plant areas are shown but no details are provided.

2.2 Systems Description

This section provides a more detailed description of the various systems within the plant and addressed in the case studies. Each system is described separately.

2.2.1 Primary Coolant System

The following notes and Figure 10 define the Primary Coolant System:

1. The Primary Coolant Loop consists of the Reactor Vessel, one Reactor Coolant Pump, and one Steam Generator and the Pressurizer, along with associated piping.

2-2

2. The Pressurizer is equipped with a normally closed Power Operated Relief Valve (PORV), which is an air operated valve (AOV-1) with its pilot solenoid operated valve (SOV-1). There is also a normally open motor operated block valve (MOV-13) upstream of the PORV.

3. The Pressure Transmitter (PT-1) on the pressurizer provides the pressure reading for the Primary Coolant Loop and is used to signal a switch from Chemical Volume Control System (CVCS) to High Pressure Injection (HPI) configuration. That is, PT-1 provides the automatic signal for high pressure injection on low RCS pressure. It also provides the automatic signal to open the PORV on high RCS pressure.

4. A nitrogen bottle provides the necessary pressurized gas to operate the PORV in case of loss of plant air but does not have sufficient capacity to support long-term operation.

2.2.2 Chemical Volume Control and High Pressure Injection Systems

The following notes and Figure 10 define the shared CVCS and HPI System:

1. The CVCS normally operates during power generation. 2. Valve type and position information include:

Valve Type Status on Loss of Power (or Air as applicable)

Position During Normal Operation

Motor Power (hp)

AOV-2 Air Operated Valve Fail Closed Open N/A

AOV-3 Air Operated Valve Fail Open Open N/A

MOV-1 Motor Operated Valve

Fail As Is Closed >5


Fail As Is Open <5


Fail As Is Closed <5









3. One of the two HPI pumps runs when the CVCS is operating. 4. One of the two HPI pumps is sufficient to provide all injection needs after a reactor trip and

all postulated accident conditions. 5. HPI and CVCS use the same set of pumps.

2-3

6. On a need for safety injection, the following lineup takes place automatically: AOV-3 closes MOV-5 and MOV-6 open MOV-2 closes. Both HPI pumps receive start signal, the stand-by pump starts and the operating pump

continues operating. MOV-1 and MOV-9 open.

7. HPI is used for re-circulating sump water after a Loss of Coolant Accident (LOCA). For recirculation, upon proper indication of low RWST level and sufficient sump level, the operator manually opens MOV-3 and MOV-4, closes MOV-5 and MOV-6, starts the RHR pump, and aligns CCW to the RHR heat exchanger.

8. RWST provides the necessary cooling water for the HPI pumps during injection. During the recirculation mode, HPI pump cooling is provided by the recirculation water.

9. There are level indications of the RWST and containment sump levels that are used by the operator to know when to switch from high pressure injection to recirculation cooling mode.

10. The Air Compressor provides the motive power for operating the Air Operated Valves but the detailed connections to the various valves are not shown.

2.2.4 Residual Heat Removal System

The following notes and Figure 10 define the Residual Heat Removal (RHR) System:

1. The design pressure of the RHR system downstream of MOV-8 is low. 2. Valve type and position information include:

Valve Type Status on Loss of Power


Motor Power (hp)


Fail As Is Closed (breaker racked out)

>5




Fails As Is Closed >5

3. Operators have to align the system for shutdown cooling, after reactor vessel de-pressurization from the control room by opening MOV-7 and MOV-8, turn the RHR pump on and establish cooling in the RHR Heat Exchanger.

2.2.5 Auxiliary Feedwater System

The following notes and Figure 11 define the Auxiliary Feedwater (AFW) System:

2-4

1. One of three pumps of the AFW system can provide the necessary secondary side cooling for reactor heat removal after a reactor trip.

2. Pump AFW-A is motor-driven, AFW-B is steam turbine-driven, and AFW-C is diesel-driven.

3. Valve type and position information include:


Position During Normal

Operation Motor

Power (hp)

















4. Upon a plant trip, Main Feedwater isolates and AFW automatically initiates by starting AFW-A and AFW-C pumps, opening the steam valves MOV-14 and MOV-15 to operate the AFW-B steam-driven pump, and opening valves MOV-10, MOV-11, and MOV-18.

5. The CST has sufficient capacity to provide core cooling until cold shutdown is achieved. 6. The test return paths through MOVs-16, 17, and 19 are low flow lines and do not represent

significant diversions of AFW flow even if the valves are open 7. There is a high motor temperature alarm on AFW pump A. Upon indication in the control

room, the operator is to stop the pump immediately and have the condition subsequently checked by dispatching a local operator.

8. The atmospheric relief valve opens, as needed, automatically to remove decay heat if/should the main condenser path be unavailable.

9. The connections to the Main Turbine and Main Feedwater are shown in terms of one Main Steam Isolation Valve (MSIV) and a check valve. Portions of the plant beyond these interfacing components will not be addressed in the course.

2-5

10. Atmospheric dump valve AOV-4 is used to depressurize the steam generator in case of a tube rupture.

2.2.6 Electrical System

Figure 12 is a one-line diagram of the Electrical Distribution System (EDS). Safety related buses are identified by the use of alphabetic letters (e.g., SWGR-A, MCC-B1, etc.) while the non-safety buses use numbers as part of their designations (e.g., SWGR-1 and MCC-2).

The safety-related portions of the EDS include 4160 volt switchgear buses SWGR-A and SWGR-B, which are normally powered from the startup transformer SUT-1. In the event that off-site power is lost, these switchgear receive power from emergency diesel generators EDG-A and EDG-B. The 480 volt safety-related load centers (LC-A and LC-B) receive power from the switchgear buses via station service transformers SST-A and SST-B. The motor control centers (MCC-A1 and MCC-B1) are powered directly from the load centers. The MCCs provide motive power to several safety-related motor operated valves (MOVs) and to DC buses DC BUS-A and DC BUS-B via Battery Chargers BC-A and BC-B. The two 125 VDC batteries, BAT-A and BAT-B, supply power to the DC buses in the event that all AC power is lost. DC control power for the 4160 safety-related switchgear is provided through distribution panels PNL-A and PNL-B. The 120 VAC vital loads are powered from buses VITAL-A and VITAL-B, which in turn receive their power from the DC buses through inverters INV-A and INV-B.

The non-safety portions of the EDS reflect a similar hierarchy of power flow. There are important differences however. For example, 4160 volt SWGR-1 and SWGR-2 are normally energized from the unit auxiliary transformer (UAT-1) with backup power available from SUT-1. A cross-tie breaker allows one non-safety switchgear bus to provide power to the other. Non-safety load centers LC-1 and LC-2 are powered at 480 volts from the 4160 volt switchgear via SST-1 and SST-2. These load centers provide power directly to the non-safety MCCs. The non-vital DC bus (DC BUS-1) can be powered from either MCC via an automatic transfer switch (ATS-1) and battery charger BC-1 or directly from the 125 volt DC battery, BAT-1.

2.2.7 Other Systems

The following systems and equipment are mentioned in the plant description but not explicitly included in the fire PRA:

Component Cooling Water (CCW) – provides cooling to Letdown Heat Exchanger and the RHR Heat Exchanger– assumed to be available at all times.

It is assumed that the control rods can successfully insert and shutdown the reactor under all conditions.

It is assumed that the ECCS and other AFW related instrumentation and control circuits (other than those specifically noted in the diagrams) exist and are perfect such that in all cases, they would sense the presence of a LOCA or otherwise a need to trip the plant and provide safety injection and auxiliary feedwater by sending the proper signals to the affected components (i.e., close valves and start pumps, insert control rods, etc.).

2-6

Instrument air is required for operation of AOV-1, AOV-2, AOV-3, and AOV-4.

2.3 Plant Layout

The following notes augment the information provided in Figures 1 through 9 (Drawings A-01 through A09):

The main structures of the plant are as follows: - Containment - Auxiliary Building - Turbine Building - Diesel Generator Building - Intake Structure - Security Building

In Figure 1 (Drawing A-01), the dashed lines represent the fence that separates two major parts: the Yard and Switchyard.

Switchyard is located outside the Yard with a separate security access. CST, RWST, UAT, Main Transformer and SUT are located in the open in the Yard. All walls shown in Figures 1 through 8 (Drawings A-01 through A-08) should be assumed as

fire rated. All doors shown in Figures 1 through 8 (Drawings A-01 through A-08) should be assumed as

fire rated and normally closed. Battery rooms A and B are located inside the respective switchgear rooms with 1-hour rated

walls, ceilings and doors. All cable trays are open type. Vertical cable trays are designated as VCBT and horizontal

cable trays as HCBT. For horizontal cable trays, the number following the letters indicate the elevation of the cable tray. For example, HCBT+35A denotes a horizontal cable tray at elevation +35 ft.

The stairwell in the Aux. Building provides access to all the floors of the building. The doors and walls are fire rated and doors are normally closed.

2.4 SNPP Drawings

The following 12 pages (pages 2-7 through 2-18) provide schematic drawings of the SNPP. Drawings A-01 through A-09 are general physical layout drawings providing plan and elevation views of the plant. These drawings also identify the location of important plant equipment. Drawing A-10 provides a piping and instrumentation diagram (P&ID) for the primary coolant system, and drawing A-11 provides a P&ID for the secondary systems. Drawing A-12 is a simplified one-line diagram of the plant power distribution system.

SECURITYBLDG.

AUX BLDG.

DGBLDG.SWITCH

YARD

INTAKESTRUCTURE

TURBINE BLDG.

CONTAINMENT

RWST

AA

AA

CST

DG-B

DG-A

MAINTRANSFORMER

UAT

SUT

YARD

14

7

12

8B

8A

15

13

BATTERYROOM 1

1

6/22/09

A-01

Revision No.:

Date:

Drawing No.:

PLANT LAYOUT

GENERAL

SNPP

MAIN CONTROL ROOM

CABLE SPREADING ROOM

SWITCHGEAR ROOMS

RHR PUMP ROOM

CHARGING PUMP ROOM

+70 FT

+55 FT

+40 FT

+20 FT

+0 FT GRADE

-20 FT

MOV-7RCP-1

SOV-1

PRESSURIZER

AOV-1MOV-13

PLANT AIR

N2

5 6 9 10 11

2

4A 4B

7

1

3

2

7/21/11

A-02

Revision No.:

Date:

Drawing No.:

PLANT LAYOUT

SECTION AA

SNPP

NOTES:

1. VERTICAL PIPE PENETRATIONTO UPPER ELEVATION.

2. PENETRATION TO UPPERFLOOR IS SEALED.

UP

MOV-3MOV-4

RHRPUMP

RH

RH

X

VCBT 10(NOTE 2)

HCBT 0(NOTE 2)

HCBT: HORIZONTAL CABLE TRAYVCBT: VERTICAL CABLE TRAY

MOV-8

NOTE 1

MOV-20

4B

4A

MO

V-1

0

MO

V-1

1

MO

V-1

5MO

V-1

4

AFW

-A

AFW

-B

2

7/21/11

A-03

Revision No.:

Date:

Drawing No.:

AUX BLDG

EL. - 20FT

SNPP

NOTE:

1. VERTICAL PIPE PENETRATINGTHE FLOOR.

UP

VCBT 0AVCBT: VERTICAL CABLE TRAY

VCBT 0B

MOV-1MOV-9 AOV-3

HPI-AHPI-B MOV-6

MOV-5

NOTE 1 MOV-2VCT

AOV-2

RWSTLE

TDO

WN

HX

CV-3

CV-4 CV-2

1

6/22/09

A-04

Revision No.:

Date:

Drawing No.:

AUX BLDG

EL. 0FT

SNPP

UP

VCBT +20AVCBT +20B

HCBT +35A


HCBT +37B

SWITCHGEARROOM A

SWITCHGEARROOM B

BATTERYROOM A

BATTERYROOM B

120VAC-A

MCC-A

120V

AC

-B

MCC-B

SST-A

4KV

BU

S-A

4KV

BU

S-B

120V

DC

-A

120V

DC

-B

SST-B

LC-A

LC-B

BC

-AIN

V-A

PN

L-A

BC

-BIN

V-B

PN

L-B

SWG ACCESSROOM

65

7

10 11

9

HCBT +37B

1

6/22/09

A-05

Revision No.:

Date:

Drawing No.:

AUX BLDG

EL. +20FT

SNPP

UP


HCBT +50B

VCBT +40B

VCBT +40A

HCBT +50A

3

1

6/22/09

A-06

Revision No.:

Date:

Drawing No.:

AUX BLDG

EL. +40FTCABLE SPREADING ROOM

SNPP

UP

MAIN CONTROL ROOM

KITCHEN

MCB

SHIFTSUPERVISOR

OFFICE

CONTROLROOM

ACCESS

7

1

1

6/22/09

A-07

Revision No.:

Date:

Drawing No.:

AUX BLDG

MAIN CONTROL ROOM

SNPP

AOV-4

MOV-16

MOV-19

MOV-17

AFW-C

V-12

CST

HCBT +10A

HCBT +10B

4KV BUS 1

4KV BUS 2

SST-1 LC-1

SST-2 LC-2

250V/DC BUS-1

BC-1

MC

C-1

MC

C-2

AIR COMPRESSOR 1

MOV-18

ATS

BAT-1

BATTERY ROOM

15

12

1

(Diesel Driven)

1

6/22/09

A-08

Revision No.:

Date:

Drawing No.:

TURBINE BLDG

EL. 0FT

SNPP

ELECTRICAL TURBINE ANDMAIN FEED RX CONTROL

PR

ESS

UR

ELZE

RC

ON

TRO

L

AU

X.FE

ED

WA

TER

HPI + RHR SW +CCW

DG-A DG-B

AOV-1

PT-1

MOV-13

MOV-14, MOV-15MOV-17, MOV-16, MOV-19

AUX. FEEDWATER PUMP A, PUMP CMOV-11, MOV-10, MOV-18

SEE DETAILBELOW

LI-1LI-2

TI-1

MOV-6MOV-5

RHR PUMP

LI-3LI-4

MOV-4MOV-3

MOV-9MOV-1

AOV-3

PUMP HPI-BPUMP HPI-A

MOV-2AOV-2

MOV-8MOV-7

DETAIL FROMABOVE

TURBINE

MAIN FEED

3' 2' 2' 4'CB-6 CB-4 CB-2

CB-1

CB-3 CB-5CB-7

AOV-4

MOV-20

3'

1

6/22/09

A-09

Revision No.:

Date:

Drawing No.:

MAIN CONTROL ROOM

MAIN CONTROL BOARD

SNPP

HPI-A

HPI-B

MOV-1

MOV-9

RHR HEATEXCHANGER

CV-3

CV

-4

LETDOWN HEATEXCHANGER

FO

MO

V-5

MO

V-6

CV

-2M

OV

-2

AOV-2OPERATEDVIA SOV-2

(AIR)

TI-1FC

CCW

CCW

LI-2LI-1

SKID MOUNTEDAIR COMPRESSOR

AIRSUPPLY

VCTPZR

MOV-13FC

OPERATED VIASOV-1 WITH N2BACKED BY AIR

AOV-1(PORV)

PT-1SG

RVMOV-7 RCP-1

MOV-3

MOV-4

MOV-20

LI-3 LI-4

AOV-3OPERATED

VIA SOV(AIR)

RHR PUMPMOV-8

RWST

2

7/21/11

A-10

Revision No.:

Date:

Drawing No.:

PRIMARY SYSTEMP & I D

SNPP

SG

MOV-14 MOV-15

MOV-11

MOV-10

MOV-18

MOV-19

MOV-16

MOV-17

MSIV

CST

AFW-B(STEAM)

AFW-A

AFW-C(DIESEL DRIVEN)

V-11L.O.

GOVERNOR

FROMFEEDWATER

TO TURBINECONDENSER

ATOMOSPHERERELIEF VALVE

AOV-4

A-1

STEAMTURBINE

1

6/22/09

A-11

Revision No.:

Date:

Drawing No.:

SECONDARY SYSTEMP & I D

SNPP

SWGR-1

RCP-1

UAT-1

SWGR-2

Spare

COMP-1

SST-1

LC-1

MCC-1

SST-2

LC-2

MCC-2

BC-1

SWGR-A

AFW-AHPI-A

SUT-1

SST-A

LC-A

SWGR-B

RHR-B HPI-B

SST-B

LC-B

ATS-1

EDG-A EDG-BG

MCC-A1

BC-A MO

V-1

MO

V-3

MO

V-5

MO

V-7

MO

V-1

0

MO

V-1

3

MO

V-1

8

(Rac

ked

Out

)

BC-B

MCC-B1

MO

V-2

MO

V-4

MO

V-6

MO

V-8

MO

V-9

MO

V-1

7

MO

V-2

0

MO

V-1

9

MO

V-1

8

DC BUS-A

VITAL-A

TI-1SOV-1LI-3LI-1 ANN-1

INV-A

BAT-A

DC BUS-B

VITAL-B

LI-2 LI-4PT-1

INV-B

BAT-B

SO

V-3

SO

V-2

SW

GR

-B

SW

GR

-A

ED

G-A

PNL-A PNL-B

MO

V-1

4

MO

V-1

1

MOV-15

VolReg

DC BUS-1

BAT-1

AOV-4

NON-VITALDC LOADS

SWYD OFF-SITE POWER

2

7/21/11

A-12

Revision No.:

Date:

Drawing No.:

ELECTRICALONE-LINE DIAGRAM

SNPP

3-1

3 MODULE 1: PRA/SYSTEMS

The following is a short description of the Fire PRA technical tasks covered in Module 1. For further details, refer to the individual task descriptions in Volume 2 of EPRI 1011989, NUREG/CR-6850.

Fire PRA Component Selection (Task 2). The selection of components that are to be credited for plant shutdown following a fire is a critical step in any Fire PRA. Components selected would generally include many components credited in the 10 CFR 50 Appendix R post-fire SSD analysis. Additional components will likely be selected, potentially including any and all components credited in the plant’s internal events PRA. Also, the proposed methodology would likely introduce components beyond either the 10 CFR 50 Appendix R list or the internal events PRA model. Such components are often of interest due to considerations of multiple spurious actuations that may threaten the credited functions and components.




Post-Fire Human Reliability Analysis (Task 12). This task considers operator actions for manipulation of plant components. Task 12 is covered in limited detail in the PRA/Systems module. In particular, those aspects of Task 12 that deal with identifying and incorporating human failure events (HFEs) into the plant response model are discussed. Methods for quantifying human error probabilities (HEPs) are deferred to Module 4.


Uncertainty and Sensitivity Analyses (Task 15). This task describes the approach to follow for identifying and treating uncertainties throughout the Fire PRA process. The treatment

3-2

may vary from quantitative estimation and propagation of uncertainties where possible (e.g., in fire frequency and non-suppression probability) to identification of sources without quantitative estimation. The treatment may also include one-at-a-time variation of individual parameter values or modeling approaches to determine the effect on the overall fire risk (sensitivity analysis).

4-1

4 MODULE 2: ELECTRICAL ANALYSIS


Fire PRA Cable Selection (Task 3). This task provides instructions and technical considerations associated with identifying cables supporting those components selected in Task 2. In previous Fire PRA methods (such as EPRI FIVE and Fire PRA Implementation Guide) this task was relegated to the SSD analysis and its associated databases. This document offers a more structured set of rules for selection of cables.



5-1

5 MODULE 3: FIRE ANALYSIS



Fire Ignition Frequency (Task 6). This task describes the approach to develop frequency estimates for fire compartments and scenarios. Ignition frequencies are provided for 37 item types that are categorized by ignition source type and location within the plant. For example, ignition frequencies are provided for transient fires in the Turbine Buildings and in the Auxiliary Buildings. A method is provided on how to specialize these frequencies to the specific cases and conditions.

Scoping fire Modeling (Task 8). Scoping fire modeling is the first task in the Fire PRA framework where fire modeling tools are used to identify ignition sources that may impact the fire risk of the plant. Screening some of the ignition sources, along with the applications of severity factors to the unscreened ones, may reduce the compartment fire frequency previously calculated in Task 6.

Detailed Fire Modeling (Task 11). This task describes the method to examine the consequences of a fire. This includes consideration of scenarios involving single compartments, multiple fire compartments, and the main control room. Factors considered include initial fire characteristics, fire growth in a fire compartment or across fire compartments, detection and suppression, electrical raceway fire barrier systems), and damage from heat and smoke. Special consideration is given to turbine generator (T/G) fires, hydrogen fires, high-energy arcing faults, cable fires, and main control board (MCB) fires.

Seismic Fire Interactions (Task 13). This task is a qualitative approach for identifying potential interactions between an earthquake and fire.

6-1

6 MODULE 4: FIRE PRA HUMAN RELIABILITY ANALYSIS

The following is a short description of the Fire PRA technical tasks covered in Module 4. For further details relative to this technical task, refer to the individual task descriptions in Volume 2 of EPRI 1011989, NUREG/CR-6850.

Post-Fire Human Reliability Analysis (Task 12). This task considers operator actions for manipulation of plant components. The analysis task procedure provides structured instructions for identification and inclusion of these actions in the Fire PRA. The procedure also provides instructions for incorporating human error probabilities (HEPs) into the fire PRA analysis.

Note that NUREG/CR-6850, EPRI 1011989 did not develop a detailed fire HRA methodology. Training module 4 is instead based on a joint EPRI/RES project as documented in NUREG-1921, EPRI 1019196, EPRI/NRC-RES Fire Human Reliability Analysis Guidelines – Draft Report for Comment. Publication of the final report remains pending. The training materials presented here are based on the draft guidance including consideration of public review comments received and the team’s response to those comments.

7-2

7

MODULE 5: ADVANCED FIRE MODELING




Note that NUREG/CR-6850, EPRI 1011989 did not provide detailed guidance on the application of fire modeling tools. Rather, the base methodology document assumes that the analyst will apply a range of computation fire modeling tools to support the analysis, provides recommended practice relative to the general development/definition of fire scenarios and provides recommendations for characterizing of various fire sources (e.g., heat release rate transient profiles and peak heat release rate distribution curves). The question of selecting and applying appropriate fire modeling tools was left to the analyst’s discretion.

Training module 5 is instead based on a joint EPRI/RES project as documented in NUREG-1924, EPRI 1019195, Nuclear Power Plant Fire Modeling Application Guide – Draft Report for Comment. Publication of the final report remains pending. The training materials presented here are based on the draft guidance including consideration of public review comments received and the team’s response to those comments.

EPRI/NRC-RES Fire PRA Course

Revision Date: June 21, 2011

Electric Power Research Institute (EPRI) Division of Risk Analysis 3412 Hillview Avenue Office of Nuclear Regulatory Research (RES) Palo Alto, CA 94303 U.S. Nuclear Regulatory Commission Washington, DC 20555

2

iii

PREPARERS

TECHNICAL TEAM LEADS:

Bijan Najafi

Science Applications International Corp. 1671 Dell Ave, Suite 100 Campbell, CA 95008

Steven P. Nowlen

Sandia National Laboratories (SNL) PO Box 5800 Albuquerque, NM 87185-0748

PROJECT MANAGERS:

Richard Wachowiak

Electric Power Research Institute

J. S. Hyslop

U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research Fire Research Branch

v

CONTENTS

1 INTRODUCTION .................................................................................................................... 1-1

1.1 Background .................................................................................................................. 1-1

1.2 How to Use this Package ............................................................................................. 1-4

1.3 References .................................................................................................................. 1-4

2 EXAMPLE CASE PLANT - GENERAL INFORMATION ......................................................... 2-1

2.1 Overall Plant Description ............................................................................................. 2-1

2.2 Systems Description .................................................................................................... 2-1

2.2.1 Primary Coolant System .......................................................................................... 2-1

2.2.2 Chemical Volume Control and High Pressure Injection Systems ............................ 2-2

2.2.4 Residual Heat Removal System .............................................................................. 2-3

2.2.5 Auxiliary Feedwater System .................................................................................... 2-3

2.2.6 Electrical System ..................................................................................................... 2-5

2.2.7 Other Systems ......................................................................................................... 2-5

2.3 Plant Layout ................................................................................................................. 2-6

2.4 SNPP Drawings.............................................................................................................. 2-6

3 MODULE 1: PRA/SYSTEMS .................................................................................................. 3-1

4 MODULE 2: ELECTRICAL ANALYSIS ................................................................................... 4-1

5 MODULE 3: FIRE ANALYSIS ................................................................................................. 5-1

6 MODULE 4: FIRE PRA HUMAN RELIABILITY ANALYSIS .................................................... 6-1

7 MODULE 5: ADVANCED FIRE MODELING .......................................................................... 7-2

vii

LIST OF ACRONYMS

AFW Auxiliary Feedwater ATWS Anticipated Transient Without Scram BWR Boiling Water Reactor CCDP Conditional Core Damage Probability CF Cable (Configuration) Factors CCW Component Cooling Water CDF Core Damage Frequency CFD Computational Fluid Dynamics CFR Code of Federal Regulations CLERP Conditional Large Early Release Probability CM Corrective Maintenance CRS Cable and Raceway (Database) System CVCS Chemical and Volume Control System EDG Emergency Diesel Generator EF Error Factor EOP Emergency Operating Procedure EPR Ethylene-Propylene Rubber EPRI Electric Power Research Institute FEDB Fire Events Database FEP Fire Emergency Procedure FHA Fire Hazards Analysis FIVE Fire-Induced Vulnerability Evaluation (EPRI TR 100370) FMRC Factory Mutual Research Corporation FPRAIG Fire PRA Implementation Guide (EPRI TR 105928) FRSS Fire Risk Scoping Study (NUREG/CR-5088) FSAR Final Safety Analysis Report HEAF High Energy Arcing Fault HEP Human Error Probability HFE Human Failure Event HPI High Pressure Injection HPCI High Pressure Coolant Injection HRA Human Reliability Analysis HRR Heat Release Rate HVAC Heating, Ventilation, and Air Conditioning ICDP Incremental Core Damage Probability ILERP Incremental Large Early Release Probability

viii

IPE Individual Plant Examination IPEEE Individual Plant Examination of External Events IS Ignition Source ISLOCA Interfacing Systems Loss of Coolant Accident KS Key Switch LERF Large Early Release Frequency LFL Lower Flammability Limit LOC Loss of Control LOCA Loss of Coolant Accident MCC Motor Control Center MCR Main Control Room MG Motor-Generator MOV Motor Operated Valve MQH McCaffrey, Quintiere and Harkleroad’s Method MS Main Steam NC No Consequence NEI Nuclear Energy Institute NEIL Nuclear Electric Insurance Limited NFPA National Fire Protection Association NPP Nuclear Power Plant NPSH Net Positive Suction Head NQ cable Non-Qualified (IEEE-383) cable NRC Nuclear Regulatory Commission P&ID Piping and Instrumentation Diagram PE Polyethylene PM Preventive Maintenance PMMA Polymethyl Methacrylate PORV Power Operated Relief Valve PRA Probabilistic Risk Assessment PSF Performance Shaping Factor PVC Polyvinyl Chloride PWR Pressurized Water Reactor Q cable Qualified (IEEE-383) cable RCP Reactor Coolant Pump RCS Reactor Coolant System RDAT Computer program for Bayesian analysis RES The Office of Nuclear Regulatory Research (at NRC) RHR Residual Heat Removal RPS Reactor Protection System RWST Refueling Water Storage Tank SDP Significance Determination Process SGTR Steam Generator Tube Rupture SI Safety Injection SO Spurious Operation SOV Solenoid Operated Valve SRV Safety Relief Valve

ix

SSD Safe Shutdown SSEL Safe Shutdown Equipment List SUT Start-up Transformer T/G Turbine/Generator TGB Turbine-Generator Building TSP Transfer Switch Panel UAT Unit Auxiliary Transformer VCT Volume Control Tank VTT Valtion Teknillinen Tutkimuskeskus (Technical Research Centre of Finland) XLPE Cross-Linked Polyethylene ZOI Zone of Influence

1-1

1 INTRODUCTION

1.1 Background

The U.S. Nuclear Regulatory Commission and Electric Power Research Institute under a Memorandum of Understanding (MOU) on Cooperative Nuclear Safety Research have been developing state of the art methods for conduct of fire PRA. In September 2005, this work produced the “EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities,” EPRI 1011989, and NUREG/CR-6850 [1].

A Fire PRA Course has been put together to train interested parties in the application of this methodology. The Course/Seminar is provided in five parallel modules. The first three modules are based directly on Reference [1]. However, that document did not cover fire human reliability analysis (HRA) methods in detail. For 2010, the training materials were enhanced to include a fourth module based on a more recent EPRI/RES collaboration and a draft guidance document, EPRI 1019196, NUREG-1921 [2] published in late 2009. The training materials are based on this draft document including the consideration of public comments received on the draft report and the team’s responses to those comments. For 2011 a fifth training module on Advanced Fire Modeling techniques and concepts has been added to the course. This module is based on the another joint RES/EPRI collaboration and a draft guidance published in January 2010, EPRI 1019195, NUREG-1934 [3].

The four training modules are:

Module 1: PRA/Systems Analysis - This module covers the technical tasks for development of the system response to a fire including human failure events. Specifically, this module covers Tasks/Sections 2, 4, 5, 7, 14, and 15 of Reference [1].

Module 2: Electrical Analysis – This module covers the technical tasks for analysis of electrical failures as the result of a fire. Specifically, this module covers Tasks/Sections 3, 9, and 10 of Reference [1].

Module 3: Fire Analysis – This module covers technical tasks involved in development of fire scenarios from initiation to target (e.g., cable) impact. Specifically, this module covers Tasks/Sections 1, 6, 8, 11, and 13 of Reference [1].

Module 4: Fire Human Reliability Analysis: This module covers the technical tasks associated with identifying and analyzing operator actions and performance during a postulated fire scenario. Specifically, this module covers Task 12 as outlined in Reference [1] based on the application of the approaches documented in Reference [2].

1-2

Module 5: Advanced Fire Modeling: This module is new for the 2011 training course and covers the fundamentals of fire science and provided practical implementation guidance for the application of fire modeling in support of a fire PRA. Module 5 covers fire modeling applications for Tasks 8 and 11 as outlined in Reference [1] based on the material presented in Reference [3].

Integral to Modules 1, 2 and 3 is a set of hands-on problems based on a fictitious, simplified nuclear power plant. The same power plant is used in all three modules. This document provides the background information for the problem sets of each module. Clearly, the power plant defined in this package is an extremely simplified one that in many cases does not meet any regulatory requirements or good engineering practices. Design features presented are focused on bringing forward the various aspects of the Fire PRA methodology. This package includes a general description of the power plant and the internal events PRA needed as input to the Fire PRA.

For Module 4 and 5, independent sets of examples are used to illustrate key points of the analysis procedures. The examples for these two modules are not tied to the simplified plant. Module 4 uses examples that were derived based largely on pilot applications of the proposed fire HRA methods and on independent work of the EPRI and RES HRA teams. The examples for Module 5 were taken directly from Reference [3] and represent a range of typical NPP fire scenarios across a range of complexity and that highlight some of the computation challenges associated with the NPP fire PRA fire modeling applications.

The instruction package for specific technical tasks is provided in Sections 3, 4, 5 and 6 which are organized by Modules (see above). A short description of the Fire PRA technical tasks is provided below. For further details, refer to the individual task descriptions in EPRI 1011989, NUREG/CR-6850, Volume 2. The figure presented at the end of this chapter provides a simplified flow chart for the analysis process and indicates which training module covers each of the analysis tasks.


Fire PRA Component Selection (Task 2). The selection of components that are to be credited for plant shutdown following a fire is a critical step in any Fire PRA. Components selected would generally include many, but not necessarily all components credited in the 10 CFR 50 Appendix R post-fire SSD analysis. Additional components will likely be selected, potentially including most but not all components credited in the plant’s internal events PRA. Also, the proposed methodology would likely introduce components beyond either the 10 CFR 50 Appendix R list or the internal events PRA model. Such components are often of interest due to considerations of multiple spurious actuations that may threaten the credited functions and components; as well as due to concerns about fire effects on instrumentation used by the plant crew to respond to the event.

Fire PRA Cable Selection (Task 3). This task provides instructions and technical considerations associated with identifying cables supporting those components selected in Task 2. In previous Fire PRA methods (such as EPRI FIVE and Fire PRA Implementation

1-3

Guide) this task was relegated to the SSD analysis and its associated databases. This document offers a more structured set of rules for selection of cables.



Fire Ignition Frequency (Task 6). This task describes the approach to develop frequency estimates for fire compartments and scenarios. Significant changes from the EPRI FIVE method have been made in this task. The changes generally relate to use of challenging events, considerations associated with data quality, and increased use of a fully component-based ignition frequency model (as opposed to the location/component-based model used, for example, in FIVE).


Scoping Fire Modeling (Task 8). This step provides simple rules to define and screen fire ignition sources (and therefore fire scenarios) in an unscreened fire compartment.



Detailed Fire Modeling (Task 11). This task describes the method to examine the consequences of a fire. This includes consideration of scenarios involving single compartments, multiple fire compartments, and the main control room. Factors considered include initial fire characteristics, fire growth in a fire compartment or across fire compartments, detection and suppression, electrical raceway fire barrier systems, and damage from heat and smoke. Special consideration is given to turbine generator (T/G) fires, hydrogen fires, high-energy arcing faults, cable fires, and main control board (MCB) fires. There are considerable improvements in the method for this task over the EPRI FIVE and Fire PRA Implementation Guide in nearly all technical areas.

1-4

Post-Fire Human Reliability Analysis (Task 12). This task considers operator actions for manipulation of plant components. The analysis task procedure provides structured instructions for identification and inclusion of these actions in the Fire PRA. The procedure also provides instructions for incorporating human error probabilities (HEPs) into the fire PRA analysis. (Note that NUREG/CR-6850, EPRI 1011989 did not develop a detailed fire HRA methodology. Fire-specific HRA guidance can be found in NUREG-1921, EPRI 1019196, EPRI/NRC-RES Fire Human Reliability Analysis Guidelines – Draft Report for Comment, November 2009. Publication of the final Fire HRA report remains pending.)

Seismic Fire Interactions (Task 13). This task is a qualitative approach to help identify the risk from any potential interactions between an earthquake and fire.


Uncertainty and Sensitivity Analyses (Task 15). This task describes the approach to follow for identifying and treating uncertainties throughout the Fire PRA process. The treatment may vary from quantitative estimation and propagation of uncertainties where possible (e.g., in fire frequency and non-suppression probability) to identification of sources without quantitative estimation. The treatment may also include one-at-a-time variation of individual parameter values or modeling approaches to determine the effect on the overall fire risk (sensitivity analysis).

1.2 How to Use this Package

This package is intended to provide the background information necessary to perform some of the problem sets of the Course/Seminar. Please note:

1. All Course/Seminar attendees are expected to review Section 2 of this document and become familiar with the power plant defined in that section.

2. The instructors of each module will provide questions or case study problem sets and will guide the attendees to sections relevant to each specific problem set. Attendees will be expected to review those relevant sections and use the information or examples provided in those sections to complete the assigned problem set.

3. Do not make any additional assumptions in terms of equipment, systems, or plant layout other than those presented in the problem package without consulting the instructor.

1.3 References 1. EPRI 1011989, NUREG/CR-6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear

Power Facilities, September 2005. 2. EPRI 1019196, NUREG-1921, EPRI/NRC-RES Fire Human Reliability Analysis Guidelines

– Draft Report for Comment, Technical Update, November 2009.

3. EPRI 1019195, NUREG-1934, Nuclear Power Plant Fire Modeling Application Guide – Draft Report for Comment, January 2010.

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1-6

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2 EXAMPLE CASE PLANT - GENERAL INFORMATION

2.1 Overall Plant Description

This chapter provides background information about the fictitious plant used in the hands-on problem sets of Modules 1, 2 and 3. Note that the examples used in Module 4 (HRA) are not based on the example case plant.

The following notes generally describe the example case plant, including its layout:

1. The plant is a Pressurized Water Reactor (PWR) consisting of one Primary Coolant Loop, which consists of one Steam Generator, one Reactor Coolant Pump and the Pressurizer. A Chemical Volume Control System and multiple train High Pressure Injection system, as well as a single train Residual Heat Removal system interface with the primary system

2. The secondary side of the plant contains a Main Steam and Feedwater loop associated with the single Steam Generator, and a multiple train Auxiliary Feedwater System to provide decay heat removal.

3. The operating conditions and parameters of this plant are similar to that of a typical PWR. For example, the primary side runs at about 2,200 psi pressure. The steam generator can reject the decay heat after a reactor trip. There is a possibility for feed and bleed.

4. It is assumed that the reactor is initially at 100% power. 5. The plant is laid out in accordance with Figures 1 through 9. The plant consists of a

Containment Building, Auxiliary Building, Turbine Building, Diesel Generator Building and the Yard. All other buildings and plant areas are shown but no details are provided.

2.2 Systems Description

This section provides a more detailed description of the various systems within the plant and addressed in the case studies. Each system is described separately.

2.2.1 Primary Coolant System

The following notes and Figure 10 define the Primary Coolant System:

1. The Primary Coolant Loop consists of the Reactor Vessel, one Reactor Coolant Pump, and one Steam Generator and the Pressurizer, along with associated piping.

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2. The Pressurizer is equipped with a normally closed Power Operated Relief Valve (PORV), which is an air operated valve (AOV-1) with its pilot solenoid operated valve (SOV-1). There is also a normally open motor operated block valve (MOV-13) upstream of the PORV.

3. The Pressure Transmitter (PT-1) on the pressurizer provides the pressure reading for the Primary Coolant Loop and is used to signal a switch from Chemical Volume Control System (CVCS) to High Pressure Injection (HPI) configuration. That is, PT-1 provides the automatic signal for high pressure injection on low RCS pressure. It also provides the automatic signal to open the PORV on high RCS pressure.

4. A nitrogen bottle provides the necessary pressurized gas to operate the PORV in case of loss of plant air but does not have sufficient capacity to support long-term operation.

2.2.2 Chemical Volume Control and High Pressure Injection Systems

The following notes and Figure 10 define the shared CVCS and HPI System:

1. The CVCS normally operates during power generation. 2. Valve type and position information include:

Valve Type Status on Loss of Power (or Air as applicable)


Motor Power (hp)

AOV-2 Air Operated Valve Fail Closed Open N/A

AOV-3 Air Operated Valve Fail Open Open N/A




Fail As Is Open <5











3. One of the two HPI pumps runs when the CVCS is operating. 4. One of the two HPI pumps is sufficient to provide all injection needs after a reactor trip and

all postulated accident conditions. 5. HPI and CVCS use the same set of pumps.

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6. On a need for safety injection, the following lineup takes place automatically: AOV-3 closes MOV-5 and MOV-6 open MOV-2 closes. Both HPI pumps receive start signal, the stand-by pump starts and the operating pump

continues operating. MOV-1 and MOV-9 open.

7. HPI is used for re-circulating sump water after a Loss of Coolant Accident (LOCA). For recirculation, upon proper indication of low RWST level and sufficient sump level, the operator manually opens MOV-3 and MOV-4, closes MOV-5 and MOV-6, starts the RHR pump, and aligns CCW to the RHR heat exchanger.

8. RWST provides the necessary cooling water for the HPI pumps during injection. During the recirculation mode, HPI pump cooling is provided by the recirculation water.

9. There are level indications of the RWST and containment sump levels that are used by the operator to know when to switch from high pressure injection to recirculation cooling mode.

10. The Air Compressor provides the motive power for operating the Air Operated Valves but the detailed connections to the various valves are not shown.

2.2.4 Residual Heat Removal System

The following notes and Figure 10 define the Residual Heat Removal (RHR) System:

1. The design pressure of the RHR system downstream of MOV-8 is low. 2. Valve type and position information include:



Motor Power (hp)


Fail As Is Closed (breaker racked out)

>5




Fails As Is Closed >5

3. Operators have to align the system for shutdown cooling, after reactor vessel de-pressurization from the control room by opening MOV-7 and MOV-8, turn the RHR pump on and establish cooling in the RHR Heat Exchanger.

2.2.5 Auxiliary Feedwater System

The following notes and Figure 11 define the Auxiliary Feedwater (AFW) System:

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1. One of three pumps of the AFW system can provide the necessary secondary side cooling for reactor heat removal after a reactor trip.

2. Pump AFW-A is motor-driven, AFW-B is steam turbine-driven, and AFW-C is diesel-driven.

3. Valve type and position information include:


Position During Normal

Operation Motor

Power (hp)

















4. Upon a plant trip, Main Feedwater isolates and AFW automatically initiates by starting AFW-A and AFW-C pumps, opening the steam valves MOV-14 and MOV-15 to operate the AFW-B steam-driven pump, and opening valves MOV-10, MOV-11, and MOV-18.

5. The CST has sufficient capacity to provide core cooling until cold shutdown is achieved. 6. The test return paths through MOVs-16, 17, and 19 are low flow lines and do not represent

significant diversions of AFW flow even if the valves are open 7. There is a high motor temperature alarm on AFW pump A. Upon indication in the control

room, the operator is to stop the pump immediately and have the condition subsequently checked by dispatching a local operator.

8. The atmospheric relief valve opens, as needed, automatically to remove decay heat if/should the main condenser path be unavailable.

9. The connections to the Main Turbine and Main Feedwater are shown in terms of one Main Steam Isolation Valve (MSIV) and a check valve. Portions of the plant beyond these interfacing components will not be addressed in the course.

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10. Atmospheric dump valve AOV-4 is used to depressurize the steam generator in case of a tube rupture.

2.2.6 Electrical System

Figure 12 is a one-line diagram of the Electrical Distribution System (EDS). Safety related buses are identified by the use of alphabetic letters (e.g., SWGR-A, MCC-B1, etc.) while the non-safety buses use numbers as part of their designations (e.g., SWGR-1 and MCC-2).

The safety-related portions of the EDS include 4160 volt switchgear buses SWGR-A and SWGR-B, which are normally powered from the startup transformer SUT-1. In the event that off-site power is lost, these switchgear receive power from emergency diesel generators EDG-A and EDG-B. The 480 volt safety-related load centers (LC-A and LC-B) receive power from the switchgear buses via station service transformers SST-A and SST-B. The motor control centers (MCC-A1 and MCC-B1) are powered directly from the load centers. The MCCs provide motive power to several safety-related motor operated valves (MOVs) and to DC buses DC BUS-A and DC BUS-B via Battery Chargers BC-A and BC-B. The two 125 VDC batteries, BAT-A and BAT-B, supply power to the DC buses in the event that all AC power is lost. DC control power for the 4160 safety-related switchgear is provided through distribution panels PNL-A and PNL-B. The 120 VAC vital loads are powered from buses VITAL-A and VITAL-B, which in turn receive their power from the DC buses through inverters INV-A and INV-B.

The non-safety portions of the EDS reflect a similar hierarchy of power flow. There are important differences however. For example, 4160 volt SWGR-1 and SWGR-2 are normally energized from the unit auxiliary transformer (UAT-1) with backup power available from SUT-1. A cross-tie breaker allows one non-safety switchgear bus to provide power to the other. Non-safety load centers LC-1 and LC-2 are powered at 480 volts from the 4160 volt switchgear via SST-1 and SST-2. These load centers provide power directly to the non-safety MCCs. The non-vital DC bus (DC BUS-1) can be powered from either MCC via an automatic transfer switch (ATS-1) and battery charger BC-1 or directly from the 125 volt DC battery, BAT-1.

2.2.7 Other Systems

The following systems and equipment are mentioned in the plant description but not explicitly included in the fire PRA:

Component Cooling Water (CCW) – provides cooling to Letdown Heat Exchanger and the RHR Heat Exchanger– assumed to be available at all times.

It is assumed that the control rods can successfully insert and shutdown the reactor under all conditions.

It is assumed that the ECCS and other AFW related instrumentation and control circuits (other than those specifically noted in the diagrams) exist and are perfect such that in all cases, they would sense the presence of a LOCA or otherwise a need to trip the plant and provide safety injection and auxiliary feedwater by sending the proper signals to the affected components (i.e., close valves and start pumps, insert control rods, etc.).

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Instrument air is required for operation of AOV-1, AOV-2, AOV-3, and AOV-4.

2.3 Plant Layout

The following notes augment the information provided in Figures 1 through 9 (Drawings A-01 through A09):

The main structures of the plant are as follows: - Containment - Auxiliary Building - Turbine Building - Diesel Generator Building - Intake Structure - Security Building

In Figure 1 (Drawing A-01), the dashed lines represent the fence that separates two major parts: the Yard and Switchyard.

Switchyard is located outside the Yard with a separate security access. CST, RWST, UAT, Main Transformer and SUT are located in the open in the Yard. All walls shown in Figures 1 through 8 (Drawings A-01 through A-08) should be assumed as

fire rated. All doors shown in Figures 1 through 8 (Drawings A-01 through A-08) should be assumed as

fire rated and normally closed. Battery rooms A and B are located inside the respective switchgear rooms with 1-hour rated

walls, ceilings and doors. All cable trays are open type. Vertical cable trays are designated as VCBT and horizontal

cable trays as HCBT. For horizontal cable trays, the number following the letters indicate the elevation of the cable tray. For example, HCBT+35A denotes a horizontal cable tray at elevation +35 ft.

The stairwell in the Aux. Building provides access to all the floors of the building. The doors and walls are fire rated and doors are normally closed.

2.4 SNPP Drawings

The following 12 pages (pages 2-7 through 2-18) provide schematic drawings of the SNPP. Drawings A-01 through A-09 are general physical layout drawings providing plan and elevation views of the plant. These drawings also identify the location of important plant equipment. Drawing A-10 provides a piping and instrumentation diagram (P&ID) for the primary coolant system, and drawing A-11 provides a P&ID for the secondary systems. Drawing A-12 is a simplified one-line diagram of the plant power distribution system.

SECURITYBLDG.

AUX BLDG.

DGBLDG.SWITCH

YARD

INTAKESTRUCTURE

TURBINE BLDG.

CONTAINMENT

RWST

AA

AA

CST

DG-B

DG-A

MAINTRANSFORMER

UAT

SUT

YARD

14

7

12

8B

8A

15

13

BATTERYROOM 1

1

6/22/09

A-01

Revision No.:

Date:

Drawing No.:

PLANT LAYOUT

GENERAL

SNPP

MAIN CONTROL ROOM

CABLE SPREADING ROOM

SWITCHGEAR ROOMS

RHR PUMP ROOM

CHARGING PUMP ROOM

+70 FT

+55 FT

+40 FT

+20 FT

+0 FT GRADE

-20 FT

MOV-7RCP-1

SOV-1

PRESSURIZER

AOV-1MOV-13

PLANT AIR

N2

5 6 9 10 11

2

4A 4B

7

1

3

2

7/21/11

A-02

Revision No.:

Date:

Drawing No.:

PLANT LAYOUT

SECTION AA

SNPP

NOTES:

1. VERTICAL PIPE PENETRATIONTO UPPER ELEVATION.

2. PENETRATION TO UPPERFLOOR IS SEALED.

UP

MOV-3MOV-4

RHRPUMP

RH

RH

X

VCBT 10(NOTE 2)

HCBT 0(NOTE 2)


MOV-8

NOTE 1

MOV-20

4B

4A

MO

V-1

0

MO

V-1

1

MO

V-1

5MO

V-1

4

AFW

-A

AFW

-B

2

7/21/11

A-03

Revision No.:

Date:

Drawing No.:

AUX BLDG

EL. - 20FT

SNPP

NOTE:

1. VERTICAL PIPE PENETRATINGTHE FLOOR.

UP

VCBT 0AVCBT: VERTICAL CABLE TRAY

VCBT 0B

MOV-1MOV-9 AOV-3

HPI-AHPI-B MOV-6

MOV-5

NOTE 1 MOV-2VCT

AOV-2

RWSTLE

TDO

WN

HX

CV-3

CV-4 CV-2

1

6/22/09

A-04

Revision No.:

Date:

Drawing No.:

AUX BLDG

EL. 0FT

SNPP

UP

VCBT +20AVCBT +20B

HCBT +35A


HCBT +37B

SWITCHGEARROOM A

SWITCHGEARROOM B

BATTERYROOM A

BATTERYROOM B

120VAC-A

MCC-A

120V

AC

-B

MCC-B

SST-A

4KV

BU

S-A

4KV

BU

S-B

120V

DC

-A

120V

DC

-B

SST-B

LC-A

LC-B

BC

-AIN

V-A

PN

L-A

BC

-BIN

V-B

PN

L-B

SWG ACCESSROOM

65

7

10 11

9

HCBT +37B

1

6/22/09

A-05

Revision No.:

Date:

Drawing No.:

AUX BLDG

EL. +20FT

SNPP

UP


HCBT +50B

VCBT +40B

VCBT +40A

HCBT +50A

3

1

6/22/09

A-06

Revision No.:

Date:

Drawing No.:

AUX BLDG

EL. +40FTCABLE SPREADING ROOM

SNPP

UP

MAIN CONTROL ROOM

KITCHEN

MCB

SHIFTSUPERVISOR

OFFICE

CONTROLROOM

ACCESS

7

1

1

6/22/09

A-07

Revision No.:

Date:

Drawing No.:

AUX BLDG

MAIN CONTROL ROOM

SNPP

AOV-4

MOV-16

MOV-19

MOV-17

AFW-C

V-12

CST

HCBT +10A

HCBT +10B

4KV BUS 1

4KV BUS 2

SST-1 LC-1

SST-2 LC-2

250V/DC BUS-1

BC-1

MC

C-1

MC

C-2

AIR COMPRESSOR 1

MOV-18

ATS

BAT-1

BATTERY ROOM

15

12

1

(Diesel Driven)

1

6/22/09

A-08

Revision No.:

Date:

Drawing No.:

TURBINE BLDG

EL. 0FT

SNPP

ELECTRICAL TURBINE ANDMAIN FEED RX CONTROL

PR

ESS

UR

ELZE

RC

ON

TRO

L

AU

X.FE

ED

WA

TER

HPI + RHR SW +CCW

DG-A DG-B

AOV-1

PT-1

MOV-13

MOV-14, MOV-15MOV-17, MOV-16, MOV-19

AUX. FEEDWATER PUMP A, PUMP CMOV-11, MOV-10, MOV-18

SEE DETAILBELOW

LI-1LI-2

TI-1

MOV-6MOV-5

RHR PUMP

LI-3LI-4

MOV-4MOV-3

MOV-9MOV-1

AOV-3

PUMP HPI-BPUMP HPI-A

MOV-2AOV-2

MOV-8MOV-7

DETAIL FROMABOVE

TURBINE

MAIN FEED

3' 2' 2' 4'CB-6 CB-4 CB-2

CB-1

CB-3 CB-5CB-7

AOV-4

MOV-20

3'

1

6/22/09

A-09

Revision No.:

Date:

Drawing No.:

MAIN CONTROL ROOM

MAIN CONTROL BOARD

SNPP

HPI-A

HPI-B

MOV-1

MOV-9

RHR HEATEXCHANGER

CV-3

CV

-4

LETDOWN HEATEXCHANGER

FO

MO

V-5

MO

V-6

CV

-2M

OV

-2

AOV-2OPERATEDVIA SOV-2

(AIR)

TI-1FC

CCW

CCW

LI-2LI-1

SKID MOUNTEDAIR COMPRESSOR

AIRSUPPLY

VCTPZR

MOV-13FC

OPERATED VIASOV-1 WITH N2BACKED BY AIR

AOV-1(PORV)

PT-1SG

RVMOV-7 RCP-1

MOV-3

MOV-4

MOV-20

LI-3 LI-4

AOV-3OPERATED

VIA SOV(AIR)

RHR PUMPMOV-8

RWST

2

7/21/11

A-10

Revision No.:

Date:

Drawing No.:

PRIMARY SYSTEMP & I D

SNPP

SG

MOV-14 MOV-15

MOV-11

MOV-10

MOV-18

MOV-19

MOV-16

MOV-17

MSIV

CST

AFW-B(STEAM)

AFW-A

AFW-C(DIESEL DRIVEN)

V-11L.O.

GOVERNOR

FROMFEEDWATER

TO TURBINECONDENSER

ATOMOSPHERERELIEF VALVE

AOV-4

A-1

STEAMTURBINE

1

6/22/09

A-11

Revision No.:

Date:

Drawing No.:

SECONDARY SYSTEMP & I D

SNPP

SWGR-1

RCP-1

UAT-1

SWGR-2

Spare

COMP-1

SST-1

LC-1

MCC-1

SST-2

LC-2

MCC-2

BC-1

SWGR-A

AFW-AHPI-A

SUT-1

SST-A

LC-A

SWGR-B

RHR-B HPI-B

SST-B

LC-B

ATS-1

EDG-A EDG-BG

MCC-A1

BC-A MO

V-1

MO

V-3

MO

V-5

MO

V-7

MO

V-1

0

MO

V-1

3

MO

V-1

8

(Rac

ked

Out

)

BC-B

MCC-B1

MO

V-2

MO

V-4

MO

V-6

MO

V-8

MO

V-9

MO

V-1

7

MO

V-2

0

MO

V-1

9

MO

V-1

8

DC BUS-A

VITAL-A

TI-1SOV-1LI-3LI-1 ANN-1

INV-A

BAT-A

DC BUS-B

VITAL-B

LI-2 LI-4PT-1

INV-B

BAT-B

SO

V-3

SO

V-2

SW

GR

-B

SW

GR

-A

ED

G-A

PNL-A PNL-B

MO

V-1

4

MO

V-1

1

MOV-15

VolReg

DC BUS-1

BAT-1

RCP-1

NON-VITALDC LOADS

SWYD OFF-SITE POWER

3

10/19/11

A-12

Revision No.:

Date:

Drawing No.:

ELECTRICALONE-LINE DIAGRAM

SNPP

3-1

3 MODULE 1: PRA/SYSTEMS


Fire PRA Component Selection (Task 2). The selection of components that are to be credited for plant shutdown following a fire is a critical step in any Fire PRA. Components selected would generally include many components credited in the 10 CFR 50 Appendix R post-fire SSD analysis. Additional components will likely be selected, potentially including any and all components credited in the plant’s internal events PRA. Also, the proposed methodology would likely introduce components beyond either the 10 CFR 50 Appendix R list or the internal events PRA model. Such components are often of interest due to considerations of multiple spurious actuations that may threaten the credited functions and components.




Post-Fire Human Reliability Analysis (Task 12). This task considers operator actions for manipulation of plant components. Task 12 is covered in limited detail in the PRA/Systems module. In particular, those aspects of Task 12 that deal with identifying and incorporating human failure events (HFEs) into the plant response model are discussed. Methods for quantifying human error probabilities (HEPs) are deferred to Module 4.


Uncertainty and Sensitivity Analyses (Task 15). This task describes the approach to follow for identifying and treating uncertainties throughout the Fire PRA process. The treatment

3-2

may vary from quantitative estimation and propagation of uncertainties where possible (e.g., in fire frequency and non-suppression probability) to identification of sources without quantitative estimation. The treatment may also include one-at-a-time variation of individual parameter values or modeling approaches to determine the effect on the overall fire risk (sensitivity analysis).

4-1

4 MODULE 2: ELECTRICAL ANALYSIS


Fire PRA Cable Selection (Task 3). This task provides instructions and technical considerations associated with identifying cables supporting those components selected in Task 2. In previous Fire PRA methods (such as EPRI FIVE and Fire PRA Implementation Guide) this task was relegated to the SSD analysis and its associated databases. This document offers a more structured set of rules for selection of cables.



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5 MODULE 3: FIRE ANALYSIS



Fire Ignition Frequency (Task 6). This task describes the approach to develop frequency estimates for fire compartments and scenarios. Ignition frequencies are provided for 37 item types that are categorized by ignition source type and location within the plant. For example, ignition frequencies are provided for transient fires in the Turbine Buildings and in the Auxiliary Buildings. A method is provided on how to specialize these frequencies to the specific cases and conditions.



Seismic Fire Interactions (Task 13). This task is a qualitative approach for identifying potential interactions between an earthquake and fire.

6-1

6 MODULE 4: FIRE PRA HUMAN RELIABILITY ANALYSIS


Post-Fire Human Reliability Analysis (Task 12). This task considers operator actions for manipulation of plant components. The analysis task procedure provides structured instructions for identification and inclusion of these actions in the Fire PRA. The procedure also provides instructions for incorporating human error probabilities (HEPs) into the fire PRA analysis.

Note that NUREG/CR-6850, EPRI 1011989 did not develop a detailed fire HRA methodology. Training module 4 is instead based on a joint EPRI/RES project as documented in NUREG-1921, EPRI 1019196, EPRI/NRC-RES Fire Human Reliability Analysis Guidelines – Draft Report for Comment. Publication of the final report remains pending. The training materials presented here are based on the draft guidance including consideration of public review comments received and the team’s response to those comments.

7-2

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MODULE 5: ADVANCED FIRE MODELING




Note that NUREG/CR-6850, EPRI 1011989 did not provide detailed guidance on the application of fire modeling tools. Rather, the base methodology document assumes that the analyst will apply a range of computation fire modeling tools to support the analysis, provides recommended practice relative to the general development/definition of fire scenarios and provides recommendations for characterizing of various fire sources (e.g., heat release rate transient profiles and peak heat release rate distribution curves). The question of selecting and applying appropriate fire modeling tools was left to the analyst’s discretion.

Training module 5 is instead based on a joint EPRI/RES project as documented in NUREG-1924, EPRI 1019195, Nuclear Power Plant Fire Modeling Application Guide – Draft Report for Comment. Publication of the final report remains pending. The training materials presented here are based on the draft guidance including consideration of public review comments received and the team’s response to those comments.

Documents

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