NUREG/CP-0194

Volume 1 of 3

EPRI 1020621

Final Report

Methods for Applying Risk

Analysis to Fire Scenarios

(MARIA FIRES) -2008

NRC-RES/EPRI Fire PRA Workshop

Volume 1

Overall Course

and

Module 1: PRA/HRA

Based on the Joint

NRC-RES/EPRI Training Workshops

Conducted in 2008

September 28 - October 2, 2008, and

November 17-20, 2008, Bethesda, MD

David Stroup

Felix Gonzalez

Hugh Woods

U.S. Nuclear Regulatory Commission

Office of Nuclear Regulatory Research

Washington, DC 20555-0001

Electric Power Research Institute

3420 Hillview Avenue

Palo Alto, CA 94303
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Methods for Applying Risk Analysis to

Fire Scenarios (MARIAFIRES)-2008

NRC-RES/EPRI Fire PRA Workshop

Volume 1: Overall Course and Module 1: PRA/HRA

NUREG/CP-0194 EPRI 1020621

Final Report

July 2010

U.S. Nuclear Regulatory Commission

Office of Nuclear Regulatory Research

Washington, DC 20555-0001

U.S. NRC-RES Project Manager

M. H. Salley

Electric Power Research Institute

3420 Hillview Avenue

Palo Alto, CA 94303

EPRI Project Manager

R. P. Kassawar
COURSE TRAINING INSTRUCTORS AND PROJECT MANAGERS

Sandia National Laboratories (SNL)

1515 Eubank SE

Albuquerque, NM 87185

Steven Nowlen

Mardy Kazarians (Kazarians & Associates)

Frank Wyant

Jeff LaChance

Science Applications International Corp. (SAIC)

1671 Dell Ave, Suite 100

Campbell, CA 95008

Bijan Najafi

Francisco Joglar-Biloch

Dan Funk (EDAN Engineering)

Richard Anoba (Anoba Consulting Services)

J.S. Hyslop/ Kendra Hill

U.S. NRC-RES Project Managers

CITATIONS

This report was prepared by:

R.P. Kassawara

EPRI Project Manager

U.S. Nuclear Regulatory Commission,

Office of Nuclear Regulatory Research (RES)

Washington, DC 20555-0001

Electric Power Research Institute (EPRI)

3420 Hillview Avenue

Palo Alto, CA 94303

Principal Investigators:

D. Stroup

F. Gonzales

Principal Investigator:

R. Kassawara

This report describes research sponsored jointly by the U.S. Nuclear Regulatory Commission,

RES, and EPRI.

The report is a corporate document that should be cited in the literature in the following manner:

Methods for Applying Risk Analysis to Fire Scenarios (MARIAFIRES), NRC-RES/EPRI Fire

PRA Workshop, Volume 1: Overall Course and Module 1: PRA/HRA, U.S. Nuclear Regulatory

Commission, Office of Nuclear Regulatory Research (RES), Washington, DC 20555-0001, and

Electric Power Research Institute (EPRI), Palo Alto, CA, NUREG/CP-0194 and EPRI 1020621.

NUREG/CP-0194 and EPRI 1020621, Volume 1,

has been reproduced from the best available copy

11
DISCLAIMER OF WARRANTIES AND LIMITATION OF LIABILITIES

THIS DOCUMENT WAS PREPARED BY THE ORGANIZATION(S) NAMED BELOW AS AN

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INSTITUTE, INC, (EPRI). NEITHER EPRI NOR ANY MEMBER OF EPRI, ANY COSPONSOR, THE

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U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research

NOTE

For further information about EPRI, call the EPRI Customer Assistance Center at

800.313.3774 or

e-mail askeDriaepri.com.

Electric Power Research Institute, EPRI, and TOGETHER ...SHAPING THE FUTURE OF ELECTRICITY

are registered service marks of the Electric Power Research Institute, Inc.

ABSTRACT

The U.S. Nuclear Regulatory Commission (NRC) approved the risk-informed and

performance-based alternative regulation 10 CFR 50.48(c) in July 2004, which allows licensees

the option of using fire protection requirements contained in the National Fire Protection

Association (NFPA) Standard 805, "Performance Based Standard for Fire Protection for LightWater Reactor Electric Generating Plants, 2001 Edition," with certain exceptions. To support

licensees's use of that option, NRC and the Electric Power Research Institute (EPRI) jointly

issued NUREG/CR-6850 (EPRI 1011989) "Fire PRA Methodology for Nuclear Power

Facilities," in September 2005. That report documents the state-of-the art methods, tools, and

data for conducting a fire probabilistic risk assessment (PRA) in a commercial nuclear power

plant (NPP) application. The report is intended to serve the needs of a fire risk analysis team by

providing a general framework for conduct of the overall analysis as well as specific

recommended practices to address each key aspect of the analysis. Participants from the U.S.

nuclear power industry supported demonstration analyses and provided peer review of the

program. Methodological issues raised in past fire risk analyses, including the Individual Plant

Examination of External Events fire analyses, are addressed to the extent allowed by the current

state-of-the-art and the overall project scope. Although the primary objective of the report is to

consolidate existing state-of-the-art methods, in many areas, the newly documented methods

represent a significant advance over previous methods.

NUREG/CR-6850 does not constitute regulatory requirements, and NRC participation in

this study neither constitutes nor implies regulatory approval of applications based on the

analysis contained in this document. The analyses/methods documented in this report represent

the combined efforts of individuals from RES and EPRI. Both organizations provided specialists

in the use of fire PRA to support this work. The results from this combined effort do not

constitute either a regulatory position or regulatory guidance.

In addition, NUREG/CR-6850 can be used for risk-informed, performance-based approaches and

insights to support fire protection regulatory decision-making in general.

On 14-16 June 2005, NRC's Office of Nuclear Regulatory Research (RES) and EPRI

conducted a joint public workshop for about 80 attendees at the EPRI NDE Center in Charlotte,

NC. A second workshop was held the following year, on 24-26 May 2006, in NRC's Two White

Flint North Auditorium in Rockville, MD. About 130 people attended the second workshop.

Based on the positive public response to these two workshops, a more detailed training class was

developed by the authors of NUREG/CR-6850. Two detailed training workshops were

conducted in 2007: on 23-27 July and again on 27-30 August, both at EPRI in Palo Alto, CA.

About 100 people attended each of these workshops. In 2008, two more workshops were held

from 29 September through 2 October, and again from 17-20 November in Bethesda, MD, near

NRC Headquarters. The two workshops attracted about 170 participants including domestic

representatives from NRC Headquarters and all four regional offices, U.S. Department of

Energy, National Aeronautics and Space Administration, EPRI, NPP licensees/utilities, Nuclear

Steam Supply System vendors, consulting engineering firms, and universities. Also in

iii
attendance were international representatives from Belgium, Canada, France, Japan, South

Korea, Spain, and Sweden.

The material in this NUREG/CP was recorded at the workshops in 2008 and adapted by

RES Fire Research Branch members for use as an alternative training method for those who were

unable to physically attend the training sessions. This report can also serve as a refresher for

those who attended one or more training sessions and would be useful preparatory material for

those planning to attend a session.

NRC Disclaimer: This document's text and video content are intended solely for use as training

tools. No portions of their content are intended to represent NRC conclusions or Regulatory

Positions, and they should not be interpreted as such.

iv
CONTENTS OF VOLUME 1

A B S T R A C T ........................................................................................................ iii

ACKNOWLEDGEMENTS ...................................................................................... vii

LIST OF ACRONYMS ......................................................................................... ix

OVERALL COURSE ................................................ .. ............................ 1-1

1 COURSE OVERVIEW ......................................................................................... 1-1

1.1 IN T R O D U C T IO N .......................................................................................................... 1-1

1 .1 .1 B a c kg ro u n d .......................................................................................................... 1-1

1.1.2 How to Use These Volumes ................................................................................. 1-4

1.1.3 E P R I P e rspective .................................................................................................. 1-4

1 .1 .4 R e fe re n ce s ........................................................................................................... 1-4

1.2 GENERAL PLANT INFORMATION .............................................................................. 1-7

1.2 .1 O verall P lant D escription ...................................................................................... 1-7

1.2 .2 S ystem s D escription ............................................................................................. 1-7

1.2.2.1 Primary Coolant System ................................................................................ 1-7

1.2.2.2 Chemical Volume Control and High Pressure Injection Systems .................... 1-7

1.2.2.3 Residual Heat Removal System ..................................................................... 1-9

1.2.2.4 Auxiliary Feedwater System .......................................................................... 1-9

1.2 .2 .5 E lectrica l S yste m ......................................................................................... 1-10

1.2 .2 .6 O ther S yste m s ............................................................................................. 1-11

1.3 M O D U LE 1: P R A /H R A ................................................................................................ 1-24

1.4 MODULE 2: Electrical Analysis ................................................................................... 1-25

1.5 M O D U LE 3: Fire A nalysis ............................................................................................ 1-25

1.6 FORMAT OVERVIEW ............................................. 1-26

1 .6 .1 V id e o S e rie s ....................................................................................................... 1-2 6

1 .6 .2 E xe rc ise s ............................................................................................................ 1-2 6

1.6.3 General Comments Regarding This Course ....................................................... 1-29

2 INTRODUCTION SLIDES ....................................................................................... 2-1

SESSION 1: NRC Welcome .................................................................................. 2-1

SESSION 2: Fire Risk Requantification Project ......................................................... 2-5

SESSiON 3a: Fire PRA Methodology and Course Structure ........................................ 2-11

V
Contents OF VOLUME 1 (CONTINUED)

SESSION 3b: PRA/HRA Module Overview ............................................................. 2-20

SESSION 3c: Circuits/Electrical Module Overview ................................................... .2-30

SESSION 3d: Fire Analysis Module Overview .......................................................... 2-36

SESSION 3e: Sample Exercises/Sample Plant ........................................................ 2-44

SESSION 4: Fire HRA Guidelines Overview ...................................................... . .2-61

MODULE 1: PRAJHRA (SYSTEMS ANALYSIS)

PRA/HRA .............................................................................................................. 3-1

INTRODUCTION..................................................................................................... 3-1

SESSION la: Fire PRA Component Selection ........................................................... 3-3

SESSION 1b: Example Exercises .............................................................................. 3-20

SESSION 2: Fire Induced Risk Model Development .................................................. 3-52

SESSION 3: Qualitative/Quantitative Screening ...................................................... 3-63

SESSION 4: Screening Post-Fire HRA ................................................................... 3-75

SESSION 5: Post-Fire HRA Detailed Analysis ......................................................... 3-81

SESSION 6: Fire Risk Quantification .................................................................... 3-89

SESSION 7: Uncertainty and Sensitivity Analysis ...................................................... 3-94

APPENDICES ........................................................................................................................ A-1

APPENDIX A: Questions Asked in Module 1 Sessions ........................................... A-1

APPENDIX B: Sample Exercises Solutions (Module 1, Task 2) .................................... B-1

vi
Acknowledgements

The authors of this report greatly appreciate the efforts and contributions made by Carli Follett, David

Yates, and Thomas DeBold, summer interns from the University of Maryland's Fire Protection

Engineering Department, during development of this NUREG/Conference Proceedings report. As part of

their summer internship, they were tasked to analyze, evaluate, organize, and edit information in the

videos of this report and to develop its first draft under direct supervision of the authors. They compiled

all the video files and presentation materials used during the training sessions and used their knowledge to

create, organize, edit, and deliver a user-friendly report of the highest quality possible.

We also acknowledge the other Fire Research Branch members Kendra Hill, Aixa Belen-Ojeda, Jason

Dreisbach, Jessica Kratchman, Gabriel Taylor, J.S. Hyslop, and Mark Henry Salley for their support and

contributions during its development. Their input and encouragement facilitated its timely completion

and greatly benefited the final product.

In addition, we want to extend our gratitude to Nathan White and Thomas Kardaras of the Information

Technology and Infrastructure Branch for their quick responses addressing all our computer needs and

issues during development of the report. We greatly appreciate the support of Linda Stevenson (the

agency publication specialist) and Guy Beltz (the agency printing specialist) whose invaluable support

and expertise were critical to ensuring the published report's quality.

Also, several other 2009 Office of Nuclear Regulatory Research summer interns, specifically Minh Dang,

Daniel Breedlove, Ashley Carnes, and Anders Gilbertson, made valuable contributions to this project,

both technical and editorial, that contributed to its overall quality.

vii

LIST OF ACRONYMS

ACB

ACRS

AEP

AFW

AGS

AOP

AOV

ATHEANA

ATS

ATWS

BAT

BNL

BWR

CBDT

CCDP

CF

CCPS

CCW

CDF

CFD

CFR

CLERP

CM

CR

CRS

CST

CVCS

CWP

DC

EDG

EDS

EF

El

EOP

EPR

EPRI

Air-cooled Circuit Breaker

Advisory Committee on Reactor Safeguards

Abnormal Event Procedure

Auxiliary Feedwater

Assistance General Supervisor

Abnormal Operating Procedure

Air Operated Valve

A Technique for Human Event Analysis

Automatic Transfer Switch

Anticipated Transient Without Scram

Boric Acid Tank

Brookhaven National Laboratory

Boiling-Water Reactor

Causal Based Decision Tree

Conditional Core Damage Probability

Cable (Configuration) Factors

Center for Chemical Process Safety

Component Cooling Water

Core Damage Frequency

Computational Fluid Dynamics

Code of Federal Regulations

Conditional Large Early Release Probability

Corrective Maintenance

Control Room

Cable and Raceway (Database) System

Condensate Storage Tank

Chemical and Volume Control System

Circulating Water Pump

Direct Current

Emergency Diesel Generator

Electrical Distribution System

Error Factor

Erroneous Status Indicator

Emergency Operating Procedure

Ethylene-Propylene Rubber

Electric Power Research Institute

ix
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

HTGR High Temperature Gas-cooled Reactor

HVAC Heating, Ventilation, and Air Conditioning

ICDP Incremental Core Damage Probability

ILERP Incremental Large Early Release Probability

1NPO Institute for Nuclear Power Operations

IPE Individual Plant Examination

IPEEE Individual Plant Examination of External Events

IS Ignition Source

ISLOCA Interfacing Systems Loss of Coolant Accident

KS Key Switch

LCO Limiting Condition of Operation

LERF Large Early'Release Frequency

LFL Lower Flammability Limit

LOC Loss of Control

LOCA Loss-of-Coolant Accident

LPG Liquefied Petroleum Gas

LWGR Light-Water-cooled Graphite Reactors (Russian design)

MCB Main Control Board

MCC Motor Control Center

MCR Main Control Room

MG Motor-Generator

MFW Main Feedwater

MOV Motor-Operated Valve

MQH McCaffrey, Quintiere and Harkleroad's Method

MS Main Steam

MSIV Main Steam Isolation Valve

NC No Consequence

NEI Nuclear Energy Institute

NEIL Nuclear Electric Insurance Limited

NFPA National Fire Protection Association

NPP Nuclear Power Plant

x
NPSH Net Positive Suction Head

NQ cable Non-Qualified (IEEE-383) cable

NRC U.S. 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

PTS Pressurized Thermal Shock

PVC Polyvinyl Chloride

PWR Pressurized-Water Reactor

Q cable Qualified (IEEE-383) cable

RBMK Reactor Bolshoy Moshchnosty Kanalny (high-power channel reactor)

RCIC Reactor Core Isolation Cooling

RCP Reactor Coolant Pump

RCS Reactor Coolant System

RDAT Computer program for Bayesian analysis

RES Office of Nuclear Regulatory Research (at NRC)

RHR Residual Heat Removal

RI/PB Risk-Informed / Performance-Based

RPS Reactor Protection System

RWST Refueling Water Storage Tank

SCBA Self-Contained Breathing Apparatus

SDP Significance Determination Process

SGTR Steam Generator Tube Rupture

SI Safety Injection

SMA Seismic Margin Assessment

SNPP Simplified Nuclear Power Plant

SO Spurious Operation

SOV Solenoid Operated Valve

SRV Safety Relief Valve

SSD Safe Shutdown

SSEL Safe Shutdown Equipment List

SST Station Service Transformer

SUT Start-up Transformer

SW Service Water

SWGR Switchgear

T/G Turbine/Generator

THERP Technique for Human Error Rate Prediction

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)

xi
VVER The Soviet (and now, Russian Federation) designation for light-water

pressurized reactor

XLPE Cross-Linked Polyethylene

ZOI Zone of Influence

xii
OVERALL COURSE

1 COURSE OVERVIEW

1.1 INTRODUCTION

1.1.1 Background

The U.S. Nuclear Regulatory Commission (NRC) and Electric Power Research Institute (EPRI),

under a Memorandum of Understanding (MOU) on Cooperative Nuclear Safety Research, have

been developing state-of-the-art methods for the conduct of fire probabilistic risk assessments

(PRAs). In September 2005, this work produced the joint EPRI and NRC report, "EPRI/NRCRES Fire PRA Methodology for Nuclear Power Facilities," EPRI 1011989 and NUREG/CR

6850.

The course/seminar described in the three volumes of this NUREG/CP-0194 has been created to

train interested parties in the application of that fire PRA methodology. The course/seminar is

presented in the following three parallel parts:

" Module 1: PRA/Human Reliability Assessment (HRA) - This module covers the technical

tasks for development of the system and operator response to a fire. Specifically, this

module covers EPRI 1011989, NUREG/CR-6850, Volume 2, Sections (Technical Tasks) 2,

4, 5, 7, 12, 14, 15, and 16.

" 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 EPRI 1011989,

NUREG/CR-6850, Volume 2, Sections (Technical Tasks) 3, 9, and 10.

* 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

EPRI 1011989, NUREG/CR-6850, Volume 2, Sections (Technical Tasks) 1, 6, 8, 11, and 13.

Integral to the course is a set of hands-on exercises based on a fictitious Simplified Nuclear

Power Plant (SNPP). The same SNPP is used in all three modules. Clearly, the power plant

defined in this package is an extremely simplified one that in many cases does not meet

regulatory requirements or good engineering practices. Design features are focused on bringing

forward various aspects of the fire PRA methodology.

This Volume 1 includes a general description of the SNPP and the internal events (i.e., non-fire)

PRA needed as input for application of the fire PRA methodology. The instructions for specific

technical tasks are organized by modules (see above). A short description of all of the technical

tasks is provided below. For further details, refer to the individual fire PRA technical task

descriptions in EPRI 1011989, NUREG/CR-6850, Volume 2.

0 Plant Boundary Definition and Partitioning (Task 1). The first step in applying the fire

PRA methodology is to define the physical boundary of the analysis and to divide the area within

that boundary into analysis compartments.

1-1
0 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 safe shutdown (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.

e Fire PRA Cable Selection (Task 3). This task provides instructions and technical

considerations associated with identifying cables supporting those components selected in Task 2

above. 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. EPRI 1011989,

NUREG/CR-6850 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, without quantitative analysis, to have little or no risk significance. 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 because of either plant procedures, an automatic trip signal, or

technical specification requirements.

0 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., selfinduced station blackout).

0 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.

0 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.

0 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.

0 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

1-2
option for those fire scenarios that are significant contributors to the risk. The methodology

provided in EPRI 1011989, NUREG/CR-6850 benefits from the knowledge gained from the tests

performed in response to the circuit failure issue.

e 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. Considerable improvements can be found in

the method for this task over the EPRI FIVE and Fire PRA Implementation Guide in nearly all

technical areas.

* 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 estimating screening human error probabilities (HEPs) before detailed

fire modeling results (e.g., fire growth and damage behaviors) have necessarily been developed

or detailed circuit analyses (e.g., can the circuit spuriously actuate as opposed to simply

assuming it can actuate) have been completed. In a fire PRA, the estimation of HEP values with

high confidence is critical to the effectiveness of screening. This report does not develop a

detailed fire HRA methodology. A number of HRA methods can be adopted for fire with

appropriate additional instructions that superimpose fire effects on any of the existing HRA

methods such as the Technique for Human Error Rate Prediction (THERP), Causal Based

Decision Tree (CBDT), A Technique for Human Event Analysis (ATHEANA), etc. This would

improve consistency across analyses (i.e., fire and internal events PRA).

* Seismic Fire Interactions (Task 13). This task is a qualitative approach to help identify

the risk from any potential interactions between an earthquake and a fire.

0 Fire Risk Quantification (Task 14). The task summarizes what is to be done

for quantification of the fire risk results.

0 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 nonsuppression 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 (i.e.,

sensitivity analysis).

* Fire PRA Documentation (Task 16). This task describes the approach to follow for

documenting the Fire PRA process and its results.

Figure 1 shows the relationship between the above 16 technical tasks from EPRI 1011989,

NUREG/CR-6850, Volume 2.

1-3
1.1.2 How to Use These Volumes

These three volumes are intended to provide the background information necessary to perform

the exercise sets of the course/seminar and, ultimately, to participate in performing fire PRAs for

NPPs. Please note:

1. All course/seminar attendees were expected to review Section 1.2 of this Volume 1 and

become familiar with the power plant (i.e., the SNPP) defined in that section.

2. The instructors of each module provided questions or case study exercise sets and guided .the

attendees to material relevant to each specific exercise set. Attendees were expected to

review that relevant material and use the information or examples provided to complete the

assigned exercise sets.

3. Attendees were directed not to make any additional assumptions in terms of equipment,

systems, or plant layout other than those presented in the exercise without consulting the

instructor.

1.1.3 EPRI Perspective

"Methods for Applying Risk Analysis to Fire Scenarios (MARIAFIRES)" is a collection of the

materials that are presented at the fire PRA course provided by EPRI and RES. The training and

resulting presentation material is detailed and represents in excess of 60 hours of classroom

instruction. The training focuses on the fire PRA methods documented in the joint EPRIINRC

RES publication 1011989 and NUREG/CR-6850 along with clarifications, enhancements, and

additions provided via the frequently asked question (FAQ) process for NFPA 805.

The intent of the publication is to provide to the public the training material used at the fire PRA

training. This material is not intended to be a substitute for direct interaction that is provided in

the periodically offered fire PRA courses; rather, it is meant to augment that training and to serve

as a reference. Enthusiastic future students can use the material to become familiar with the

general principles of fire PRA prior to arrival at the course. Students who have already taken the

course can use the material for reference. The material consists of a series of reports that

document the presentations including some speakers' notes and text. In addition, an edited

version of a recorded training session is also available via a separate product number. This video

version can.be used in a similar manner to the documentation (e.g., for reference or in

preparation for the course) and includes the actual recorded and edited course.

In providing this material, it is hoped that those who plan to attend the course can arrived more

informed, those who have already attended can have a reference, and those who have been

unable to attend have a resource to gain a more complete understanding of the intent and goals of

EPRI 1011989 and NUREG/CR-6850.

1.1.4 Reference

EPRI 1011989, NUREG/CR-6850, "EPRI/NRC-RES Fire PRA Methodology for Nuclear Power

Facilities," September 2005.

1-4
Figure 1 - Relationship of Technical Tasks in NUREG/CR-6850, Volume 2

SUPPORT TASKA:

Plant Walk Downs

Hr

I

I

I

Mhduodue1PRA/HRAf

Module 2: Electrical Analysis

Module 3: Fire Analysis

SUPPORT TASK B: Fire PRA Database

ITASK Scoping 8Fire Modeling

I

EZ

1-5
Figure 1 (continued)

Note: "B" is from Task 7B, prev. pg.

Detailed Fire Scenario Analysis

" •TASK I 9:I I

*Detailed Circuit Failure An*lysis • I Detailed Fire Modeling

=_• ~A. Single CompartmentT B. Multi-Compartment •TASK 10: IC. Main Control Room "

Circuit Failure Mode & Likelihood C

Analysis

S l

TASK 13:

Seismic-Fire Interactions

Module 1: PRA/HRA

Module 2: Electrical Analysis

D Module 3: Fire Analysis

1-6
1.2 GENERAL PLANT INFORMATION

1.2.1 Overall Plant Description

The following notes generally describe the SNPP, including its layout:

1. The plant is a pressurized-water reactor (PWR) consisting of one primary coolant loop that

consists of one steam generator, one reactor coolant pump, and the pressurizer. A chemical

volume control system and multiple train 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 of feed and bleed, etc.

4. It is assumed that the reactor is initially at 100-percent power.

5. The plant is laid out in accordance with Figures 2 through 10. 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.

1.2.2 Systems Description

This section provides a more detailed description of the various systems within the plant

addressed in the case studies. Each system is described separately.

1.2.2.1 Primary Coolant System

The following notes and Figure 11 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. 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 the chemical and volume control

system (CVCS) to the high-pressure injection (HPI) configuration.

4. A nitrogen bottle provides the necessary pressurized gas to operate the PORV in case of loss

of plant air.

1.2.2.2 Chemical Volume Control and High-Pressure Injection Systems

The following notes and Figure 11 define the shared CVCS and HPI system:

1. The CVCS normally operates during power generation.

1-7
2. Valve type and position information include:

Valve Type Status on Loss of Power Position During

(or Air as applicable) Normal Operation

AOV-2 Air-Operated Valve Fail Closed Open

AOV-3 Air-Operated Valve Fail Closed Open

MOV-1 Motor-Operated Fail As Is Closed

Valve

MOV-2 Motor-Operated Fail As Is Open

Valve

MOV-3 Motor-Operated Fail As Is Closed

Valve

MOV-4 Motor-Operated Fail As Is Closed

Valve

MOV-9 Motor-Operated Fail As Is Closed

Valve

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.

6. On a need for safety injection, the following lineup takes place automatically:

" AOV-2 and AOV-3 close.

" 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. The HPI is used for recirculating sump water after a loss-of-coolant accident (LOCA) using

the pathway involving MOV-3 and MOV-4. For recirculation, the operator manually opens

MOV-3 and MOV-4 and closes MOV-5 and MOV-6 upon proper indication of low refueling

water storage tank (RWST) level and sufficient sump level.

8. RWST provides the necessary cooling water for the HPI pumps. It has enough capacity to

depressurize the primary loop for residual heat removal (RHR) cooling.

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.

1-8
10. The air compressor provides the motive power for operating the AOVs, but the detailed

connections to the various valves are not shown (nor needed for exercises).

1.2.2.3 Residual Heat Removal System

The following notes and Figure 11 define the 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 Position During

Power Normal Operation

MOV-7 Motor-Operated Fail As Is Closed

Valve

MOV-8 Motor-Operated Fail As Is Closed

Valve

3. Operators have to align the system for low-pressure cooling after reactor vessel

depressurization from the control room by opening MOV-7 and MOV-8, turning the RHR

pump on, and establishing cooling in the RHR heat exchanger.

1.2.2.4 Auxiliary Feedwater System

The following notes and Figure 12 define the auxiliary feedwater (AFW) system:

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-B is steam turbine driven.

3. Valve type and position information include:

Status on Loss Position During

Valve Type of Power Normal

Operation

MOV-10 Motor-Operated Fail As Is Closed

Valve

MOV-1 1 Motor-Operated Fail As Is Closed

Valve

MOV-14 Motor-Operated Fail As Is Closed

Valve

MOV-15 Motor-Operated Fail As Is Closed

Valve

MOV-16 Motor- Operated Fail As Is Closed

Valve

1-9
MOV-1 7 Motor-Operated Fail As Is Closed

Valve

MOV-18 Motor-Operated Fail As Is Closed

Valve

MOV-19 Motor-Operated Fail As Is Closed

Valve

4. Upon a plant trip, main feedwater isolates and AFW automatically initiates by starting AFWA and AFW-C pumps, opening the steam valves MOV- 14 and MOV- 15 to operate the AFWB steam-driven pump, and opening valves MOV- 10, MOV- 11, and MOV- 18.

5. The condensate storage tank (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.

1.2.2.5 Electrical System

Figure 13 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-B 1, etc.) while the nonsafety

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 that are normally powered from the startup transformer SUT-1. In the event that

offsite power is lost, these switchgears 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-A 1 and MCC-B 1) 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 PNLB. The 120 VAC vital loads are powered from buses VITAL-A and VITAL-B that in turn

receive their power from the DC buses through inverters INV-A and INV-B.

The nonsafety portions of the EDS reflect a similar hierarchy of power flow; however, important

differences exist. For example, 4160-volt SWGR-1 and SWGR-2 are normally energized from

1-10
the unit auxiliary transformer (UAT-1) with backup power available from SUT-1. A cross-tie

breaker allows one nonsafety switchgear bus to provide power to the other. Nonsafety 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 nonsafety MCCs. The nonvital DC bus

(DC BUS-i) can be powered from either MCC via an automatic transfer switch (ATS-1) and

battery charger BC-I or directly from the 250-volt DC battery BAT-1.

1.2.2.6 Other Systems

The following systems and equipment are mentioned in the plant description but not explicitly

included in the fire PRA:

" Service Water (SW). Provides cooling to recirculating and residual heat removal heat

exchangers - assumed to be available at all times.

" Component Cooling Water (CCW). Provides cooling to letdown 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.)

1-11
Figure 2 - SNPP General Layout

S- - -- - - - --- -- - -- - - --- -- - -- - - - -- -- - -- --

I T* II

L '$AL I -- ] I

7RAHSFCRM4ER

LFT-j

I ~BATTERY

8 IA

---------------------- ---------- I

SNPP

A-q1

PLANT LAYOUT 7o0

GENERAL

1-12
Figure 3 - SNPP Plant Layout Section AA

WVv-1 1

PRESSURIZER

008

SNPP

PLANT LAYOUT

SEC1ON AA

4-t'~ I.

402

-U~.

7g7~!Q7

C

I. I

1-13
Figure 4 - SNPP Aux Building Layout at Elevation Minus 20 ft

I

HCOBT 0

(NOTE 2)

NOTTE

I- VRTICAL PIPE PFRATION

TO UPPER ELEVATION2. PENETRATION TO UPPER

FLOOR IS SEALED.

HC1T: Hof-IZONTAL CAB - TRAY

VCBT: VERTICAL CABLE TRAY

VCST 10

(NOTE 2)

SNPP

AUX BLDG

EL - 2OFT

A-03

7M7107

0

1-14
Figure 5 - SNPP Aux Building Layout 0 ft

4

NOTE:

I VERTICAL PIPE PENETRATING

THE FLOOSL

VCBT: VERTICAL. CABLE TPRAY

\_VCBT DB

VCB1 flA SNPP

AUX BLDG

EL. OFT

1-15
Figure 6 - SNPP Aux Building Layout +20 ft

SWITCHGEAR

ROOM 8

4

D BATTERY 4 ROOM B

HCST: HORIZONTAL CABLE TRAY

VCST: VERTICAL CABLE TRAY

SNPP

AUX BLDG

EL +20FT

A435

N17107

0

1-16
Figure 7- SNPP Aux Building Layout +40 ft

VOST +406 4

VCBT -40A HCOST +506

HCOST: HORIZONTAL CABLE TRAY

VCB!: VERTICAL CABLE TRAY

SNPP

AUX BLDG

EL +4OFT

CABLE SPREADING ROOM

A-06

rq7707

1-17
Figure 8 - SNPP Aux Building Main Control Room

MCB

4

MAIN CONTROL ROOM

A.,

SNPP

AUX BLDG

MAIN CONTROL ROOM

A-07

7/7/07

oQI

1-18
Figure 9 - SNPP Turbine building

4

SNPP

TURBINE BLDG

EL- OFT

ArOS

71710l7

I I

1-19
Figure 10 - SNPP Main Control Room

MOV- 14. 11V-1 5

'Ov-. MOV-16. MoW19'

AUX- FE-.EIOVOITE V f A. PUMP C

.OV- 1., MOYV- MOV-18

MOV-13 -'

AOV-11 --

TURBINE 

I

Eu n111

PT-1 - PT-I 0 El []V El SEE DETAII" L

El MAIN FEED El El El

I0 I- 0G-A -/ -

I II rI I

IELEr-CTRICAl. TURBINE AND 0

IMAIN FEED T

CB-G C3-2

RX CONTROL

CB-1

I I HPI * RR I I"

C"- -- 4- 4'C1 cs,-5

sw+ I ccwI

Ce-7I

LI-3

FUIV~P HPI-A

PUMP HPI-S

AOV-2

MDV-2

MOV-~3

MOV-I

AOV-3

SNPP

MAIN CONTROL ROOM

MAIN CONTROL BOARD

DETAIL FROM

AEOV•

A-09

7717.07

a &

1-20
Figure 11 - Primary Coolant System Including RHR, HPI, and CVC Systems

-----------

Skid Mounted

Air Compressor

L1-21
Figure 12 - Aux. Feedwater System

Aunnapherlc

Relie

MOV-1 7

1-22
Figure 13 - Electrical Distribution System

SWYD OFF-SITEPOWER

LIAT-1 SUT-1

l1

0 >> 0 0 0-~0 00 >0>> >0 >0 00 0 0

Non-Vital DC

Loads

00 0a0 0 ý: w ~ ~ t tin e

(n U)

LI-1 LI-3 SOV-1 TI-i PT-A LI-2 LI-4

1-23
1.3 MODULE 1: PRA/HRA

The following is a short description of the fire PRA technical tasks covered in this module. For

details regarding these tasks, 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, 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 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 as well as concerns about fire effects on instrumentation used by the plant crew to

respond to the event.

* Qualitative Screening (Task 4). This task identifies fire analysis compartments that can be

shown without quantitative analysis to have little or no risk significance. 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.

0 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., selfinduced station blackout).

* 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.

0 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 estimating screening human error probabilities (HEPs) before detailed

fire modeling results have been developed (e.g., fire growth and damage behaviors or detailed

circuit analyses). In a fire PRA, estimation of HEP values with high confidence is critical to the

effectiveness of screening. This report does not develop a detailed fire HRA methodology. A

number of HRA methods can be adopted for fire with appropriate additional instructions that

superimpose fire effects on any of the existing HRA methods such as SHARP, ATHEANA, etc.

This would improve consistency across analyses (i.e., fire and internal events PRA). This task is

subdivided into Task 12a, "Screening Post-Fire HRA," and Task 12b, "Post-Fire HRA Detailed

Analysis."

1-24
* 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 nonsuppression 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 (i.e., sensitivity

analysis).

e Fire PRA Documentation (Task 16). This task describes the approach to follow for

documenting the fire PRA process and its results.

1.4 MODULE 2: Electrical Analysis

The following is a short description of the fire PRA technical tasks covered in this module. For

details regarding these tasks, refer to the individual task descriptions in Volume 2 of EPRI

1011989, NUREG/CR-6850.

9 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.

* 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 EPRI 1011989, NUREG/CR-6850 benefits from the knowledge gained from the tests

performed in response to the circuit failure issue.

1.5 MODULE 3: Fire Analysis

The following is a short description of the fire PRA technical tasks covered in this module. For

details regarding these tasks, refer to the individual task descriptions in Volume 2 of EPRI

1011989, NUREG/CR-6850.

* 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 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

1-25
ignition frequency model (as opposed to the location/component-based model used, for example,

in FIVE).

* Scoping Fire Modeling (Task 8). Scoping fire modeling is the first task in the fire PRA

framework where fire modeling tolls 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. There are considerable improvements in the

method for this task over the EPRI FIVE and Fire PRA Implementation Guide in nearly all

technical areas.

* Seismic Fire Interactions (Task 13). This task is a qualitative approach to help identify the

risk from any potential interactions between an earthquake and a fire.

1.6 FORMAT OVERVIEW

1,.6.1 Video Series

This NUREG-CP centers on video recordings of the NRC/EPRI course/seminars. Thus, the

course material is presented primarily in video format on the enclosed DVDs. A noteworthy

ramification is the absence of written text independently relating the course material. It was

decided that the words and gestures of the six lecturers were a necessary component of the

course that were preserved on the video recordings as authentically as possible.

Each module is broken into a number of sessions, corresponding to the associated tasks. Some

tasks correspond to multiple sessions, as indicated in each of the video menus. In a conventional

DVD player, these sessions appear as scenes within the DVD menu, which also indicates the

associated content. Important to note is that these DVDs comprise videos of both

course/seminars (i.e., each session was presented on two different dates). Careful selection was

made to provide the best presentation of each session's material.

Onscreen text indicates which lecture slide is currently being discussed. In addition, a fullscreen snapshot of each slide will appear to indicate what information is currently being

discussed, and a smaller-scale version of the slide will continuously be displayed onscreen while

the instructors are lecturing. These are the same slides that are present in this NUREG/CP as

well as within the associated session folders on the DVDs, accessible only by computer.

1.6.2 Exercises

The three modules of this course contain exercises that were distributed by the instructors for

completion as reinforcement of certain concepts introduced within the sessions. However, the

1-26
videos included with these three NUREG/CP volumes do not adequately describe the locations

and instructors' intentions regarding the exercises. Because of that inadequacy, this section

provides further available information regarding the exercises.

The exercises are not included in the three "Module Overview" slide presentations in the "2

INTRODUCTION SLIDES" sectionof this "OVERALL COURSE" part of Volume 1 (i.e., they

are not provided in SESSION 3b: PRA/HRA Module Overview, SESSION 3c:

Circuits/Electrical Module Overview, and SESSION 3d: Fire Analysis Module Overview).

Instead, the exercises are presented in the detailed, separate, parallel sessions for each module.

Thus, the PRA/HRA exercises are presented in the "Module 1: PRA/HRA (Systems Analysis)"

part of Volume 1; the "Module 2: Electrical Analysis" exercises are presented in Volume 2; and

the "Module 3: Fire Analysis" exercises are presented in Volume 3. The exercises are organized

differently within each of the three volumes, and it should be noted that they do not cover all 16

of the EPRI 1011989, NUREG/CR-6850 tasks.

MODULE 1: PRA/HRA

Exercises for this module begin with Task 2, "Component Selection," as presented in the

"MODULE 1: SESSION lb: Example Exercises" section of this Volume 1. These exercises

have seven steps. Step 1 asks whether a fire PRA model should incorporate a series of initiating

events. Students are asked to fill in the last two columns, explaining whether or not (and why)

each initiating event should be included in a fire PRA. For example, students are asked whether

initiator %Tl, a reactor trip, is a reasonable consequence of a fire in most NPPs. Step 2 asks the

same question, but this time about whether certain systems rather than events should reasonably

be included. Step 3, as introduced in the lecture, combines the first two steps and examines

further which fire initiators are relevant to the particular equipment selected in Step 2. Step 4

identifies possible failure paths for each of the components that remain after Step 3 and also

identifies the need for modeling each of the failure paths depending upon its relative significance

within the PRA context. Step 5 incorporates the human element of the PRA. At this point,

instrumentation issues as a consequence of fire are discussed. Students are asked to identify

specific failure mechanisms that could occur as a result of the faulty operation of each of the

pieces of instrumentation. Step 6 examines multiple failure modes that are possible due to

interactions between pieces of equipment. Only one interaction is identified in this simplified

plant. Finally, in Step 7, the fire PRA equipment list is assembled. Students are given all of the

components as well as the event identifiers and descriptions and are asked to identify the

operating and failed positions of each component, according to different modes. Appendix B

provides the solution to these Task 2 exercises.

Exercises for Task 5 (mentioned in Slide 15 of the "MODULE 1: SESSION 2: Fire Induced Risk

Model Development" section of this Volume 1 but not included in the video for that Session)

focus on development of the fire-induced risk model for those components deemed necessary

within the screening process of Task 2. There are two steps in the task. First, students are asked

to use enclosed event trees and fault trees to develop a cumulative distribution function

(CDF)/conditional core damage probability (CCDP) model that identifies where changes are

required from the existing model to facilitate event mapping. Second, students develop a large

early release frequency (LERF) or conditional large early release probability (CLERP) model,

1-27
asking essentially the same questions except about fire-initiating events. The same event and

fault trees are employed. These "Task 5" exercises and their supporting information occupy 53

pages and are not included in this NUREG/CP report. However, they are available on NRC's

public Web site (www.nrc.gov): using the Web-based ADAMS search, enter ML083330431 in

the "Search" box, right click on the "Image File" icon, select "Open Link," and see pages 68 thru

120. However, the solutions are not currently available in this NUREG/CP report nor on the

Web.

MODULE 2: ELECTRICAL ANALYSIS

The first five sessions of the electrical task develop a structured approach for identifying cables

important to the fire PRA, performing circuit analysis to refine failure modes associated with

fire-induced equipment, and assigning likelihoods of spurious operations. The approach then

determines the importance of cable failures to fire events by using electrical diagrams and logic

trees to show which outcomes are possible given unsuccessful operations of cables. An

overview of the exercises for this module is presented in the sixth session (Volume 2 of this

NUREG/CP, Session 6, "Electrical Exercises Overview"). The exercises draw from all of the

modules and employ drawings that the developers provided for completion of the analyses. One

solution is included as a video, and the rest have instructor solutions with the completed charts as

described below.

An exercise sheet for each of the 16 exercises is provided in Volume 2 in the "3 EXAMPLE

EXERCISES" section. Another copy of each exercise sheet, along with the drawings needed to

perform it, is provided in Volume 2, Appendix B, "Exercise Problems and Solutions," in its

"Exercises with Reference Figures" subsection. Solutions are provided in Volume 2, Appendix

B, "Exercise Problems and Solutions," in its "Instructors Exercise Solutions" subsection. The

exercises include analyses of various failure modes of AOVs and MOVs, failure of a reactor

coolant system pressure indicator, a spurious annunciator alarm, an HPI pump failure to start and

failure to run, an instrument air compressor failure to run, and a 480V load center fault.

It should be noted that incomplete and sometimes conflicting documentation of Exercise 14 was

available for inclusion in this NUREG/CP report that does not acceptably describe details of the

exercise nor provide its solution.

MODULE 3: FIRE ANALYSIS

Module 3 includes exercises that correspond to the associated sessions, providing reinforcement

of the material presented in the session. Instructions for each of the fire analysis exercises are

specific to the exercise and are included at the beginning of each exercise on the videos. Most

involve completing a chart with missing information, and the solutions are included within each

module and labeled as the instructor set.

In this NUREG/CP, the exercises are presented in Volume 3 in five "Example Problems" subsessions, numbered Ib, 2b, 6b, lOb, and 1 lb. The solutions for all exercises are presented in

Volume 3, Appendix C, "Exercise Set Solutions." The exercises include, respectively, selection

of the plant analysis boundary and its partitioning into fire compartments, mapping ignition

1-28
sources, preparation for walkdown, characterization of compartments with respect to features

important to fires such as fire detection and suppression features and ignition sources, and fire

modeling in the main control room.

1.6.3 General Comments Regarding This Course

The material within the DVD includes solutions to many of the example sets as well as full-size

versions of the lecture slides that are printable in assorted formats. Specifically, the solutions to

the electrical exercise sets include instructor interpretations that are challenging to read and,

therefore, were edited for use in the videos.

Questions raised by participants at the conference were removed from the videos. Generally, the

available video footage had questions that were largely inaudible. For the convenience of the

user, selected questions and answers have been transcribed and included in the associated session

folder on the DVDs. In this NUREG/CP, each volume includes selected questions and responses

in its Appendix A, "Questions Asked in Module { 1, 2, or 3 } Sessions."

To obtain the most complete set of materials for this course, instructors provided portions

(particularly of the exercises and their solutions) that they had prepared for use in these courses

over a period of time. Because this material is being continuously improved, minor

discrepancies resulted between sections (e.g., between the statement of certain exercises and

their solutions). However, these differences are not sufficiently numerous nor significant to

affect accomplishment of the overall objectives of the course for its users.

1-29

2 INTRODUCTION SLIDES

SESSION 1: NRC Welcome

2-1
Slide 1 Notes:

*-U.S.NRC

SM&

EIelfl Rt CH 11SIT1

EPRI/NRC-RES FIRE PRA

METHODOLOGY

Welcome

Mark Henry Salley P.E.

Chief, Fire Research Branch

U.S. Nuclear Regulatory Commission

Office of Nuclear Regulatory Research

Joint RES/EPRI Fire PRA Workshop

September 29 - October 2 and November 17-20, 2008

Bethesda, MD.

A Collabortion of U.S. NRC Offle of NucMe-r Reguatory ResR.roh (RES) & Electric Pofer Resear.h Institute (EPRO

Slide 2 Notes:

! Welcome

* Unique Training Opportunity

- State-of-the-Art Fire PRA for NPP application

- Presented by Leading Experts

* Joint Research Project

- Memorandum of Understanding (MOU) between NRC Office of

Nuclear Regulatory Research (RES) and Electric Power Research

Institute (EPRI)

- Performing the Training since 2005

FiR PRA W000080, 2008, B.-- Da 0S82, I0 US 00 &tUNd..,S,00815o

Module Alf: F"0.Risk 1 qonOfioat08 Proq,, R.0n 0RE0)8&.0 h 0lecto rR .. 00 0,08Lge (0PR00

2-2
Slide 3 Notes:

! Introduction

O Domestic Attendees:

- NRC Headquarters and all 4 Regional Offices

- U.S. Department of Energy (DOE)

- National Aeronautics and Space Administration (NASA)

- EPRI

- NPP Licensees/Utilities

- NSSS Vendors

- Consulting Engineering Firms

- Universities

Fee PRA WnShoP. 2000, Bftaesa. MD Modul t.1: R. Risk Rs•qufnn fchh/o Project aeae h REnS) I entrace, CeSeA MS/ud. (EPtIý

Slide 4 Notes:

! Introduction

International Attendees

- Regulators and NPP Operators

Belgium

Canada

France

Japan

South Korea

Spain

Sweden

Fire pA Wo*kMmo, 20M8, Selthsda, M0

M.Wul 1f: Rr. Rift- uatf Pmjec FPaee.u ne4IA CoIlaoro US. OffiedNc/Rgato.ary -Re Coda. 51. FS /ataeqa,,ucataa eeRod rojct Rens, & fleetrl • eaae•re tn=es (retR

2-3
Slide 5 Notes:

! Introduction

* Enjoy the Training Session

- Interact with your Peers

* Participate in the Sessions

- Ask Questions

* If you have any needs, please contact one of the members of

the NRC

Fif PRA WorksAop 2008, B00328. 0 Sd 5 C0,ofllabo iU.us, NRC Offl d N.fM RfeMO

ModuM 81: Rr Risk Req-anficatn PnJpt R5*00 (RES) a 8SPownRn• ýERO 205840 

2-4
SESSION 2: Fire Risk Requantification Project

2-5
Slide 1 Notes:

TIC I~R CRtC POWtR U.S.NRC ,-Erm EACHINST

EPRI/NRC-RES FIRE PRA

METHODOLOGY

Module I-1: Fire Risk Requantification

Project

J.S. Hyslop - NRC/RES

K. Canavan - EPRI

Joint RES/EPRI Fire PRA Workshop

September 29 - October 2 and November 17-20, 2008

Bethesda, MD

A Collaboretlon of US. NRC Office of Noclear Reglatory Research (RES) & Elecitric Poe, Research lnstitute (EPRO

Slide 2 Notes:

! 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

* Similar training conducted last year in Palo Alto, CA

* Course does not provide official NRC positions, but does

represent the expertise of authors of NUREG/CR-6850

(EPRI 1011989)

FPe PR4 WOrShop, 2008, Betheeda, MOD __ _ a Cd2aboW ondUS. NRC Office of WWP 9 toay

Module I-: Flm ReSk R flr ýdw Project Resnf flES) 8ed• eud ..,ffiem- -,80e (EPRo

2-6
Slide 3 Notes:

! Quality Product

* NRC and EPRI brought diverse group of specialists together

* Volunteer pilot plants support testing

* Peer review by other participating licensees

* Public review and comment

* Formal process to resolve technical disputes

FA PRA RNolýh. 2008, B MThesd& MD s A Colabo.di, IS. S, NRC e o3]flANuUSkN R t-y

Uoduft PI: RH, Rik equw Uflon Project Re0h IRESI S P.0500050,tf O (E-RR)

Slide 4 Notes:

! Advancement To State Of Art

* Improvements made in areas important to fire risk (resource

constraints considered)

* Means to advance

- Consolidate existing research

- Analyze more extensive data

- Modify existing methods

- Develop new approaches

Fir PRA Worshop, 2008M , [hedD] A CdlbtiOO 02 US NRC O o MD0 N laRuV

Modue I-M: Rý Risk R .quanR9cafloa PmJect Re0t0h (RES) 00020 PowE Res00 5.tu0 (EPRF

2-7
Slide 5 Notes:

! Uses Of Methodology

* Support for new rule 1 OCFR50.48c implementation - NFPA pilot plants using technology for fire PRA development/upgrade

* ANS fire risk standard development

- Typically defines state-of-art, although supports lesser capability

categories as well

* Reactor Oversight Process analyses

- Refined phase 3 analyses

- Development of phase 2 Fire Protection SDP (IMC 0609, Appendix F)

* Basis for review guidance that RES developed for NFPA 805

related changes

* Other expected uses

- Analyses under the current fire protection regulations (i.e.

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

specifications)

F~ePRA Wb 0. 2M08, Sthnda. MD 5 A CObl oo aO, NR0 0 01c l aNaV

?*dUM M: Rn 08 Rýqu-dflctfo PftJ~t F;;;08 7 OS) &8 t0.~n20s0 (M008

Slide 6 Notes:

g Implementation Impacts

* Fire PRA Methodology Standard clarifies requirements

- Particular impact on spurious actuations

* FAQ process for NFPA pilots requires implementation

guidance

- Several FAQs related to fire ignition frequency and counting of

sources solved

* Newer FAQs have been identified that address many areas

of fire PRA

- Identified in next presentation

- At various stages of completion

F0, P W ,. 200W8, bSed, M A cooton oU, NRC 08fice ofN 000008by

ModuM (-:I R R i R"qaan al, on Poje-t Re.h (fES) & E-m -nP rR-008 - (EPR00

2-8
Slide 7 Notes:

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 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 101999/NUREG-1824

F0 t PRA Wo r*08oR. 2M08. B.th.0 da, MD S A C-O -o nU.S. NRC O t 0r• N reg-00

ModuR ý1.: Rn Ris R"qu•anhcan Pr,*ct R_,rt {RSS & aeMDC fofw _ 14O ( ,

Slide 8 Notes:

! Related Activities

- EPRI 1011989/NUREG/CR-6850

- Publication Sept 2005

- First Workshop Jun 2005

- Second Workshop May 2006

- First detailed course Jul, Aug 2007

* EPRI 1011999/NUREG-1 824 May 2007

* Current Fire PRA Course Sept, Nov 08

* Fire Modeling User's Guide mid 2009

* Fire HRA Methodology Development Fall 2009

* FAQ Support Ongoing

* Fire Modeling Training Ongoing

* Low Power/Shutdown Fire PRA Methods NRC

FR. PRA WorksoWp. 2008, B0hesda, MD4 A Cd~bo0do SU SNC 0ff0 oi u 05Reuto

Modue M:12 RIk Rik cf0 . mqun project Re~toi (REs) &8 o R 5500 (EPRI)

2-9
Slide 9 Notes:

! Project Team

Covers all technical disciplines critical to Fire PRA

4k Technical Lead: B. Najafi, S. Nowlen

& General PRA & plant systems analysis: A. Kolaczkowski, R. Anoba

4k Circuit Analysis and Appendix R: D. Funk, F. Wyant

4L Human Reliability Analysis: J. Forrester, W. Hannaman, A.

Kolaczkowski

4a Fire analysis: F. Joglar, M. Kazarians

& Consultants: A. Mosleh, D. Bley

* Collectively, over 250 years of relevant experience

* Principal authors of documented Fire PRA methods in the US

for the past 2 decades

* Experience with use of previous methods; their strengths and

weaknesses

* The Methodology reflects the consensus of this team, EPRI and

RES

FRI PRA WM**. 2008, B00.•W0, MD 5 A Clbttn0°,ROS. NRC 0fiicedNCM&R0W48tGY

Uoduft 1.1: RM Risk RMquOnRtcston Project Rer, (RPS) & 8.ec Ro 0 nn3 t uOA (FFR,)

2-10
SESSION 3a: Fire PRA Methodology and Course Structure

2-11
Slide 1 Notes:

~U.S.N RC EIPf2l1RSAC INSITC4UTE

EPRI/NRC-RES FIRE PRA

METHODOLOGY

Introduction and Overview: the Fire PRA

Methodology and Course Structure

Bijan Najafi - Science Applications International Corp.

Steve Nowlen - Sandia National Laboratories

Joint RES/EPRI Fire PRA Training Workshop September and November, 2008

Bethesda, MD

I A &oIbrtoo U.S. NRC -fl.o N I.1.11Reguloty)o.. Rsn~rd (RES) &Eleotdc Poo,w Rese.,oh Ionstitte (EPRI)

Slide 2 Notes:

! Overview

1. An Overview of the EPRI/NRC-RES Fire PRA

Methodology

- Fire PRA Course:

Structure/Content

Objectives

Logistics

II. Ongoing Activities Related to Fire PRA

- ANS/ASME Fire PRA Standard

- Fire HRA, joint EPRI & RES project

PEA EE,. Eoo~S.,d F 2008R- pA__ N* ., ~Reubln , E C o (RElS) U• , Poem; Reooch trireme (EPRO)

2-12
Slide 3 Notes:

PART I

An Overview of the

EPRI/NRC-RES Fire PRA Methodology

&

The Fire PRA Course

ZP pRA r-t S.# , 2A-8 d 'S NR . 8W -Y d, Sm.), (RESJ • ) PcP., . Rý1 •t• - • XER0)

Slide 4 Notes:

! BACKGROUND

* Prior to IPEEE; Mostly simple approximate method for orderof-magnitude assessment of fire risk, e.g. NUREG/CR-2258,

Fire Risk Analysis for Nuclear Power Plants.

* EPRI FIVE (1992)

- A "vulnerability evaluation" methodology developed in response to

IPEEE program

* EPRI Fire PRA Implementation Guide (1995)

- Developed as a complement to FIVE for detailed evaluation of

unscreened fire areas/compartments

- More robust methods (compared to FIVE) for:

* Development and evaluation of fire risk model, including human actions

* Assessment of fire growth and damage, detection and suppression

* Control room and multi-compartment fire risk

F_____p oM.,&R2ASV8 AAdTm,., d U. NRCS a RS Reach (RES) A 808,,) Pow, Rea8 a R;A; (EARW2-13
Slide 5 Notes:

! EPRIINRC-RES FIRE PRA METHODOLOGY

-The methodology is presented in the form technical

task procedures within an overall process

-The process is intended as a guide and should fit

most cases

- User may adjust process based on plant-specific

information, efficiency, economy and desired

applications

A____ C ýoaRC iUSNRC , Ofice 80wRely d Pd, ~ 200 ~ ~ ~ ~ ~ ri-,~etw ~ ~ (S ,R _ _RRtf~*R~.~,t (SE-0

Slide 6 Notes:

! EPRI/NRC-RES FIRE PRA METHODOLOGY

Procedures cover the following technical areas

- Plant analysis boundary and partitioning

- Fire PRA component selection and risk model

- Circuit/cable selection, routing and failure modes analysis

- Screening, qualitative and quantitative

- Fire ignition frequency

- Fire modeling; fire growth, damage and detection/suppression

- Post-fire human reliability analysis (HRA)

- Seismic-fire interactions, and

- Fire risk quantification, including uncertainties, and documentation

i PRO Tduton SAW, WOv . M00000 8 A C - ES) l df ectrC P" N o-RsVhn Sýtt P

_ _ Wco, _~~cee [:Wid.RL R-,s~h (RES) &8000 P-.,eRes-,h XPRO0(00

2-14
Slide 7 Notes:

! PROCEDURE CONTENT

1. Purpose

2. Scope

3. Background information: General approach and

assumptions

4. Interfaces: Input/output to other tasks, plant and other

information needed, walk-downs

5. Procedure: Step-by-step instructions for conduct of the

technical task

6. References

Appendices: Technical bases, data, examples, special models

or instructions, tools or databases

F•PAT S•.e~ o, 2008 ~AC~o,0W&NC0k ~..~,a0 Rp.0.00 iOES) S BOES CtC• 58, Rn..r 5,0. t# f-PRE)

Slide 8 Notes:

Course Structure

- Three parallel modules:

- Module 1: Systems/PRA

- Module 2: Electrical Analysis

- Module 3: Fire Analysis.

* General structure for each module:

- PowerPoint presentations designed to convey key concepts and

the general "how to" of each task

- Example problems designed.to illustrate key elements of the

procedures (more on this shortly)

FR. ERA Ts SES. 2 S B5 R 5 . 5 2 R . .(sh RES) E B- .5,0 PI Resnh CE- fe BEPR)

2-15
Slide 9

Notes:

2-16
Slide 10

OI MODULE VERVIEW STRUCTURE OF FIRE PRA (2)PROCESS AND

ITDetailed Fire Scenario Analysis

Analyis TASK 11: Detailed Fire Modeling

SA. Single Compnprnent

TASK 10: CkcultFawre Mode & a Mufti-Compartment

Likelihood Ana ysis C. Main Control Room

I i I TASK 13: Seisnmic-Fire TASK 14; FRe Risk QaVcta

es• y AniysaEL Fire Module

ISystems Module

T ASK 15: F" MA

D n Electrical Module

Fire PRA TrainingS Sept. and Nov., 2008 S ie A Collaboration of U.S. NRC Office of Nuclear Regulatory

Introduction and Over.iew 2006 . Research (RES) & Electric Power Research Institute (EPRI)

Notes:

2-17
Slide 11 Notes:

! Training Objectives (1 of 3)

* Target audience:

- FPRA practitioners

- FPRA reviewers

* The "doer" versus the "reviewer"

- We are targeting both types of users, but the needs are really quite

similar

- The key elements for implementation are the same as the key

elements for review

- Understanding how and why the "doer" does what they do is one key

to understanding the analysis itself

Slide 12 Notes:

Training Objectives (2 of 3)

* Our intent:

- To deliver practical implementation training at a higher level of detail

than provided via the previous "roll-out" workshops

- To illustrate and demonstrate key aspects of the procedures

* We expect and want significant participant interaction

- Class size was limited specifically to allow for questions and

discussion

- We will take questions about the methodology

- We cannot answer questions about a specific application

- We will moderate discussions, and we will judge when the course

must move on

Fft PPA rT, nS!=ýNo, •

FENC•f •rCS•NC SRS 12 Ecottonf d u.s NRC Cffl of Nc1aRRC'yr

R~h(RCS) S BEnc, pN'e.,R.~RC irslCt* (EPRRS

2-18
! Training Objective

This training is about

and the EPRI/RES tea

method

- It is not about regulator

- It is not about alternati

standard and its "Capa

- We cannot cover areas

(e.g., detailed HRA)

Your next step:

- You should not expect

- We do expect that afte

knowledge needed to'

method or to review an

- Ultimately, you learn b

FIZ pRT1 ý , S"•L atiN•,

Slide 13 Notes:

(3 of 3)

NUREG/CR-6850, EPRI TR-1O011989

ams' intent for implementation of that

ry compliance (talk to NRR about that)

ve methods, NFPA-805, or the ANS FPRA

bility Categories"

s that are outside the scope of the document

to come out of this training an "instant expert"

r the training you will possess the fundamental

'hit the ground running" and to implement the

application involving the method

y doing, so get out there and do it!

Slide 4O(N** Cabornn US N1O RNoates

Slide 14 Notes:

! What's next:

* The rest of these introductory slides will provide a short

overview of each of the three modules

* Intent is to ensure that those of you. in one module are aware

of what is being covered in the other two modules

* FPRA requires a team effort and integration is critical to

success

* That will close out the introductions, and then you will be

breaking up and going to your individual modules

Fft TniIQ Sp.d # [:S 4 AC 5050 d2GhS lgd-y oe.arch (RS) & E-ectr P O R .s h 5*. (5500

2-19
SESSION 3b: PRA/HRA Module Overview

2-20
Module 1: PRAJHR

* This module will cover

accident response mo

quantification tasks

* Specific tasks covered

- Task 2: Equipment Sel

- Task 4: Qualitative Scr

- Task 5: Fire-Induced R

- Task 7: Quantitative S

- Task 12: Human Relial

- Task 15: Risk Quantific

- Task 16: Uncertainty A

F•ire HRA is the subject of an ongoi

guidance in current methodology

FnPeeceheeSW,~ 0WO

Slide 15 Notes:

A

r all aspects of the plant systems

deling, human reliability analysis, and

d are:

ection

eening

.isk Model

creening

bility Analysis (HRA)*

cation

nalysis

ng RES-EPRI collaboration. Results are expected to supersede

document. More on this later in this presentation...

Sd1 coube• wo s. secon NRCO a Rulaeae

Slide 16 Notes:

Task 2: Equipment Selection (1 of 3)

* Objective: To decide what subset of the plant equipment will

be modeled in the FPRA

* FPRA equipment will draw from:

- Equipment from the internal events PRA

- We do assume that an internal events PRA is available!

- Equipment from the Post-Fire Safe Shutdown analysis

e.g., the Appendix R analysis or the Nuclear Safety Analysis under

NFPA-805

- Other "new" equipment not in either of these analyses

Pu. PeA rTeuno. S•cc aco N 00'* 2•08 ___ _ A c~doabt.- (R & UES) c Rec uceegue -N 

2-21
Task 2: Equipment

Many choices to be m

influence these decisii

- Fire-induced failures th

- Mitigating equipment a

- Fire-induced failures th

- Fire-induced failures th

operator actions

:h. PRTmAC kk 20

Slide 17 Notes:

t Selection (2 of 3)

ade in this task, many factors will

ons

at might cause and initiating event

nd operator actions

at adversely impact credited equipment

at could lead to inappropriate or unsafe

[lide A COletaoRtS, NRC OfNoet oNtt:RegaMo,

Slide 18 Notes:

! Task 2: Equipment Selection (3 of 3)

Choices are important in part because "selecting" equipment

implies a burden to Identify and Trace cables

- Cable selection is covered in Module 2 (Electrical) but can represent

a significant commitment of time and effort to the FPRA

R=,ch ORES) SEIPCEO PrRR OnsiRAe OEPRIO

2-22
Slide 19 Notes:

Task 4: Qualitative Screening (1 of 2)

* Objective: To identify fire compartments that can be

screened out as insignificant risk contributors without

quantitative analysis

* This is an Optional task

- You may choose to bypass this task which means that all fire

compartments will be treated quantitatively to some level of analysis

(level may vary)

5i 9 Rema (RES) &Cetr = • 0 n 0 (EPRI)

Slide 20 Notes:

l Task 4: Qualitative Screening (2 of 2)

* Criteria are established that consider:

- Trip initiators

- Presence of selected equipment

- Presence of selected cables

* Note that any compartment that is "screened out" in this step

is reconsidered in the multi-compartment fire analysis as a

potential source of multi-compartment fires

- See Module 3, Task 11c

SRs.2 - =Erf h (OES) eoR.,Rds0, IsO,,t (EO)

2-23
! Task 5: Fire-Induc

- Objective: To construc

- Functional relationship

- Equipment failure mod

- Human Failure Events

F-k PRA Trah00,Sl., .CNdov, 200

Slide 21 Notes:

ed Risk Model (1 of 2)

ct the FPRA plant response model

s among selected equipment

es (including spurious actuation)

(HFEs)

[]jt. =21 A C ý O U N 0Sfli d 0 Rlaty

P..- (RES) & EI.•t PfO Rejt- R .4. (EMI)

Slide 22 Notes:

Task 5: Fire-Induced Risk Model (2 of 2)

- Covers both CDF and LERF

- Builds on/from the internal events model but more than just a "tweak"

- Adds fire unique equipment

- Adds fire-specific equipment failure modes (e.g., spurious actuations)

- Adds fire-specific operator actions

Ff pRAATrenoLaN.. 20ý 2 d S S, RC N .

(RES) , 20.5r0 20,.3 R. . ,,st0*A (EPo

2-24
Slide 23 Notes:

Task 7: Quantitative Screening (1 of 2)

* Objective: To identify compartments that can be shown to be

insignificant contributors to fire risk based on limited

quantitative considerations

* This task is Optional

- Analyst may choose to retain all compartments for more detailed

analysis

=P PRO T=S.0 m. 20 e23-] A Cu VS W•'- P RES) & SPP 0 R.Oeeh -~tOA (CERD)

Slide 24 Notes:

Task 7: Quantitative Screening (2 of 2)

" Screening may be performed in stages of increasing

complexity

" Consideration is given to:

- Fire ignition frequency

- Screening of specific fire sources as non-threatening (no spread, no

damage)

- Impact of fire-induced equipment and cable failures

- conditional core damage probability (CCDP)

R00 h tOES) 0000000 Powe Ro rch Iooitt (6PRm)

2-25
Slide 25 Notes:

! Post-fire Human Reliability Analysis (1 of 5)

* Objective: To assess Human Error Probabilities (HEPs) for

the Human Failure Events (HFEs) included in the FPRA plant

response model

- For those already in the level 1 PRA, and

- Those to be added specific to post-fire condition considerations

* Fire HRA is being addressed under a separate RES-EPRI

collaboration

- Dealing with both screening and quantification

- Intent is that new guidance will supersede that currently in

NUREG/CR-6850 EPRI TR-1 011989

- More on this later

* In the mean time, back to the current guidance ...

A= p.,~.~~ (EPRQ

Slide 26 Notes:

! Post-fire Human Reliability Analysis (2 of 5)

Focus on the Screening HRA - Rule-based (in the absence of detailed fire scenario information)

quantitative screening approach

* Fire conditions; environment and accessibility

* Fire-induced component/system faults, such as spurious actuations

* Location and timing to diagnose and perform the required actions

* Availability of the crew to perform the actions, not impacted by dual

responsibilities

- Instructions provided to develop screening HEPs based on these

conditions

Fire pRA TradCLb. $epU Nc. 2 A ldS FRPATm Dp 2 (RES) & RC.. OW R h ANtA, (EPRI)

2-26
Slide 27 Notes:

! Post-fire Human Reliability Analysis (3 of 5)

Use of existing HRA methods in fire conditions

- Fire performance shaping factors (PSF) defined and described

1. Available staffing resources

2. Applicability and suitability of training/experience

3. Suitability of relevant procedures and administrative controls

4. Availability and clarity of instrumentation

5. Time available and time needed to complete the action, including impact of

concurrent and competing activities

6. Environment

7. Accessibility and operability of equipment

8. Special tools

9. Communication

10. Crew dynamics and characteristics

11. Special~fitness needs

- Operator Manual Action (OMA) feasibility criteria and Post-Fire HRA

PSFs

PR rv'1 5 Sw~ :t. -0 sld j;727 A boýad. WS.NRClifl.NCrRegýdan R.st lidli & Eledgapýwýn.&h I#.A. (E-1~)

Slide 28 Notes:

Post-fire Human Reliability Analysis (4 of 5)

- Quantitative link between these PSFs & best-estimate HEPs

specifically for fire not developed but much can be handled

with existing methods and guidance.

F,. epRA T,.flng' S~"edE N~v S2ide 2 A C~tliORlit°e RfUNS NRC Olffc. ONu•tw Regflt~y

linii.tlodc~o aid liekim Re.a~h (dEli) & Uledird PowaiR.&.8R l~ifdA (CURI)

2-27
Slide 29 Notes:

! Post-Fire Human Reliability Analysis (5 of 5)

Special Case: Main Control Room (MCR) abandonment

- Additional considerations as part of PSF evaluation, e.g.,

* The decision and timing of control room evacuation

* The number, complexity and multiple-location local manual actions

* Adequacy of human-machine interface at the remote shutdown and/or

local panels

Guidance is method-neutral

- Specific HRA method is analyst choice

- General considerations are outlined and applicable regardless of

method chosen

- Specifics of treatment are method-dependent

- e.g., how you handle/quantify a performance shaping factor (PSF)

hR~dRIO~a N R.R~asR(RES) A8.00 PR. R-se MONA (EPRI

Slide 30 Notes:

! Task 14: Fire .Risk Quantification

* Objective: To quantify fire-induced CDF and LERF

* Covered in limited detail

* Relatively straight-forward roll-up for fire scenarios

considering

- Ignition frequency

- Scenario-specific equipment and cable damage

- Equipment failure modes and likelihoods

- Credit for fire mitigation (detection and suppression)

- Fire-specific HEPs

- Quantification of the FPRA plant response model

P PRR r ,S. NR&. 20E (RR CO 0 R . R ROCSNh (RES) & EetrOMPowrRBRSW R h nRRCu (EPRO

2-28
Slide 31 Notes:

Task 15: Uncertainty and Sensitivity

* Objective: Provide a process for identifying and quantifying

uncertainties in the FPRA and for identifying sensitivity

analysis cases

* Covered in limited detail

* Guidance is based on potential strategies that might be

taken, but choices are largely left to the analyst

- e.g., what uncertainties will be characterized as distributions and

propagated through the model?

(waN (RES) & 0 k - -- .,N - M-0PRO

2-29
SESSION 3c: Circuits/Electrical Module Overview

2-30
Module 2: Circuits

* This module covers th

related to the identific

analysis of electrical c

* Tasks covered are:

- Task 3: Cable Selectio

- Task 9: Detailed Circu

- Task 10: Failure Mode

- Support Task B: FPRA

Slide 32 Notes:

/Electrical

hose parts of the method specifically

ation and tracing of cables, and the

:ircuit failure modes and likelihood

n (and Routing)

it Analysis

Likelihood Analysis

Database

S 3 •i, co/OaC /u s NRC OýWdkr•dRWR,9 y

Slide 33 Notes:

1Task 3: Cable Selection (1 of 2)

* Objectives:

- To identify/select cables whose fire-induced failure could adversely

effect the operation of selected equipment

- To locate selected cables

* Cables may include Power, Control/Indication, and

Instrumentation

Fk0A T.td -, RtktNo. Rý -30 0 Nn010 00

-~,R (-)S & 051 8055I ROO_ - (-o0

2-31
Task 3: Cable Sele

* Selected cables need

* Cable routing can be

resources

- Depends a lot on status

. Scope, quality, vintage

- Tracing cables is a tim

- Intent is to allow for "W

. Iteration to identify an

- Allowances are made

assumptions about a

F=P0T0.S.0=1 00o. 200

Slide 34 Notes:

ction (2 of 2)

to be routed/located

a major commitment of FPRA

s of existing plant cable information

e, method of documentation

e consuming activity

ork smart" approaches

d route more cables as needed to support FPRA

for making "conservative"

cable's routing if unknown

______ A Co000,,oW US. NRCW O MO k0Reg.Mty

RSlid (RES & 0t02d0 (EPRo Notr e 

Slide 35 Notes:

]Support Task B: FPRA Database (1 of 2)

* Objective: Develop a database to support query needs of the

FPRA

* FPRA will ask: "if a fire damages everything within some

spatial region, what equipment and cables are lost?"

- The regions is defined by the fire scenarios (covered in Module 3)

- The region may be as large as a combination of two or more fire areas

or as small as a single raceway

* Because cables tend to be the primary driver, the FPRA

Database is covered in the electrical module

Fire PRO T'r~rg' 00se0l e 2000 S/ 35• A CRJAO°,88 of Ua SNRC Offý ý t tN,*&ý RW-°0t bOSoduct 00on and Ren,0h CRES) & 000t00 PFo,•R.0arch* 0oute rEPRO

2-32
Slide 36 Notes:

! Support Task B: FPRA Database (2 of 2)

The electrical module will describe the needs served by the

database:

- Database functionality needs

- How to assess capability of exiting systems

- How to implement a structured process to obtain required capability

- Some discussion of the potential role of new software and data

management tools

Slide 37 Notes:

Task 9: Detailed Circuit Failure Analysis (1 of 2)

- Objectives:

- To identify circuit responses (failure modes) to fire-induced cable

failures

- To screen out cables that do not impact the ability of a component to

complete its credited function

* This is not about failure mode likelihoods (that is task 10)

* This is about defining the effects that cable failure can (or

cannot have) on selected equipment

- e.g., what cables can, or cannot, cause spurious actuations?

F, PRA irm SI.P a• MNI. s A -

R,._R (RES) & 2At PR, RSCR ,,SI&A, (EPRO

2-33
Task 9: Detailed C

* Fundamentally a detei

modes and effects

* Module will cover:

- Those failure modes th

various cable/circuit co

- Underlying assumption

- Role of existing analyse

- Steps of the analysis

Slide 38 Notes:

ircuit Failure Analysis (2 of 2)

rministic analysis of cable failure

at are, and are not, considered plausible for

nfigurations and applications

s of the analysis

es (e.g., Appendix R SSD analysis)

Slide 39 Notes:

Task 10: Circuit Failure Modes Likelihood (1 of 2)

* Objective: To establish first order estimates of the

conditional probability, given failure of a specific cable, that

the circuit will respond in. a specific way

* This one is about the likelihood that certain equipment failure

modes will be observed given fire-induced cable failure

- Will the equipment spuriously actuate, or

- Will it be a loss of function failure?

- What is the relative likelihood of each failure mode of interest?

F PR- n T ot 4*0o. 2--088 A C ý (PR

2-34
Slide 40 Notes:

Task 10: Circuit Failure Mode Likelihood (2 of 2)

* This is a probabilistic analysis

* Based largely on existing data including

- The EPRI/NEI cable tests including the NRC/RES collaboration

- The EPRI expert panel

* Module will include

- Existing knowledge base

- Underlying assumptions

- Key factors in the analysis

- Analysis approach and methods

FnPpRA uM S.I . mt OV.. 2000

2-35
SESSION 3d: Fire Analysis Module Overview

2-36
Slide 41 Notes:

! Module 3: Fire Analysis

* This module covers those parts of the method specifically

related to the identification and analysis of fires, fire damage,

and fire protection systems and features

* Tasks covered are:

- Task 1: Plant Partitioning

- Task 6: Fire Ignition Frequency

- Task 8: Scoping Fire Modeling

- Task 11: Detailed FireScenario Analysis

- Task 13: Seismic/Fire Interactions (briefly)

- Support Task A: Plant Walkdowns

'We 0.0.8,00 f-O(ES) &E.0 0.n 0.000 0s0& (SPRO

Slide 42 Notes:

Task 1: Plant Partitioning (1 of 3)

* Objectives:

- To define the global analysis boundary of the FPRA

- To divide the areas within the global analysis boundary into fire

compartments

* The fire compartments become the "basic unit" of analysis

- Generally we screen based on fire compartments

- Risk results are often rolled up to a fire compartment.level

F2e PISA Tr $00•2p an200wRC P 0 . -. E8 R,.rd LRES) 000.Ott oerR~aC ,O~s,0002C h (5500

2-37
Slide 43 Notes:

!Task 1: Plant Partitioning (2 of 3)

The global analysis boundary is intended to be a liberal

definition of the region potential interest

- It will likely encompass areas of essentially no risk, but that is OK,

screening steps will identify these

The fire compartments are a matter of analysis convenience

- Fire compartments may equal fire areas if you so choose

- You can also subdivide fire areas into multiple compartments

- The sum of the fire compartments must equal the global analysis

boundary

No omissions, no overlap between compartments

Slide 44 Notes:

!Task 1: Plant Partitioning (3 of 3)

*Ultimately, the FPRA is expected to provide some resolution

to each defined fire compartment and to all locations within

the global analysis boundary

*Module will cover:

- Guidance and criteria for defining the global analysis boundary\

- Guidance and criteria for defining fire compartments

*Ultimately, there is not a lot of new guidance in this task

-A lot like what was done in the IPEEE days

2-38
Task 6: Fire Ignitic

* Objective: To define fi

of fire scenarios at va

* Fire frequencies will b

resolution:

- An entire fire area,

- A fire compartment

- A group of fire ignition

- A single ignition source

Slide 45 Notes:

on Frequency (1 of 3)

ire frequencies suitable to the analysis

rious stages of the FPRA

e needed at various levels of

sources (e.g., a bank of electrical cabinets)

e (e.g., one electrical panel)

RSlid (RES) & ENotres PowRflnW SNtlu (EPR5

Slide 46 Notes:

Task 6: Fire Ignition Frequency (2 of 3)

- Task begins with generic industry-average statistics on fire

- EPRI fire event database

- Events filtered for applicability and sorted into ignition source bins

- Plant-wide fire frequency is provided for each bin

* The real "trick'" is to convert the generic values into values

specific to your plant and to a given fire scenario

- Approach is based on ignition source counting and apportionment of

the plant-wide frequency based on local population

tnRdCton and S. le 4 A CobtRRa Ct o US NRC Rf l' Rulator R-h (ýRES) ERRU C R.- •R R, •60

2-39
Task 6: Fire Ignitic

Quite a bit is new rela

- The fire event data hav

method

- That means older IPE

- There has been a swit

and away from generic

- Some areas have rece

* e.g., main control rooi

- P . Z 0 ., SA N0v., 2008

Slide 47 Notes:

)n Frequency (3 of 3)

tive to fire frequency:

ve been re-analyzed entirely to suit the new

EE-vintage frequencies are obsolete

6h towards component-based fire frequencies

room-based fire frequencies

ived special treatment

m

Slid4 * US. NRC a Note.Ns: 

Slide 48 Notes:

Task 8: Scoping Fire Modeling (1 of 2)

* Objective: To identify (and screen out) fire ignition sources

that are non-threatening and need not be considered in

detailed fire modeling

* Non-threatening means they cannot:

- Spread fire to other combustibles, or

- Damage any FPRA equipment item or cable

PFk70 Tmtf . 0tER0q ld 40' CUO7R Rd.NC 0dNMR08 ."a=k -l - R,..RO - ~ (-J i I -8 R.Rl.,i - - - r(70

2-40
Task 8: Scoping F

- Scoping fire modeling

associated with the tri

targets

- The Fire Severity Profi

- Damage criteria for cal

- Assumptions associate

Slide 49 Notes:

ire Modeling (2 of 2)

introduces a number of key concepts

eatment of fire sources and damage

le approach

bles and equipment

td with specific fire sources

Slide 50 (RES) lRS O - (E NtO

Slide 50 Notes:

Task 11: Detailed Fire Modeling (1 of 3)

* Objective: To identify and analyze specific fire scenarios

* Divided into three sub-tasks:

- 12a: General fire compartments (as individual risk contributors)

- 12b: Main Control Room analysis

- 12c: Multi-Compartment fire scenarios

F, ERA T0 , S iseNfov,. 2008 A Ou 0,8.oEh ICES) 00)080 Po•R,.80600,E 00002 (EEO

2-41
Task 11: Detailed

Task 11 involves man

- Selection of specific fir

. Combinations of fire

- Analysis of fire growth/

- Application of fire mo

- Analysis of fire damag

- Time to failure

- Analysis of fire detectie

F" rA , ld08o., 2000

Slide 51 Notes:

Fire Modeling (2 of 3)

y key elements

e scenarios

ources and damage targets

spread

dels

e

on and suppression

0020.01 oACAo8t• I AUS Ofi 0. WNu0000g00y

Slide 52 Notes:

Task 11: Detailed Fire Modeling (3 of 3)

* Task 11 comes with a wide range of supporting appendices

including:

- Specific fire sources such as high energy arc faults, turbine generator

fires, and hydrogen fires

- Treatment of fire severity and severity factors

- Treatment of manual fire suppression

- Treatment for main control board fires

* Module will cover key appendices

Jtdfk T-N n SI• f. 8 5 A Cd0hoaort U.S NRCS O f Nbes W r WR.5A320 R.-..b (RES) A ýeAtt - -ow ( #eR5

2-42
Slide 53 Notes:

! Task 13: Seismic/Fire Interactions

* Objective: A qualitative assessment of potential fire/seismic

interactions

* Module will cover this task briefly

- No significant changes from IPEEE guidance (e.g., the Fire PRA

Implementation Guide)

F. RCA r, ". * No%0. 2&00 (PR 4

2-43
SESSION 3e: Sample Exercises / Sample Plant

2-44
Slide 54 Notes:

! Sample Problems / Sample Plant

- All modules will involve hands-on exercises

- Intent: To illustrate key aspects of the methodology through a

cohesive set of sample problems

* All exercises are built around a common sample plant - the

SNPP

* The exercises are designed such that taking all modules

together presents a fairly complete picture of the FPRA

methodology

- Not every task is covered by the SNPP sample problems

- Not every aspect of covered tasks are illustrated

- R~~aNR ~(RES) Ba S ý*DnRsN hfstcA (EPRO

Slide 55 Notes:

The SNPP: Intent and Approach

* The SNPP is not intended to reflect either regulatory

compliance or good engineering practice

- It is purely an imaginary construct intended to highlight key aspects

of the methodology - nothing more!

* The SNPP has been kept as simple as possible while still

serving the needs of the training modules

* Aspects of the plant are assumed for purposes of the

training exercises, e.g.:

- BOP equipment not covered in detail

- Some systems are assumed to remain available

- Details will be provided in each module

F.NM PRA Tr(EP S ,RCAMNY 0.,*rh (00S)05 * ('ee • w 00r ,0 000h 0*0*00P0

2-45
Slide 56 Notes:

The SNPP: Plant Characteristics

- PWR with one primary coolant loop

- One steam generator, one RCP, one pressurizer

- Shared chemical volume control system

- Multiple train injection system

- Residual heat removal system

* Secondary side includes:

- Main steam and feedwater loop for the single steam generator

- Multiple train aux. feedwater to provide decay heat removal

* Additional specifications in Chapter 2 of handouts

=RA MrJ*dV~t~~ -SW ~ ~ AC~0,U NCOidN 0 IaO

2-46
Slide 57

I The SNPP: Primary Systems P&ID

Fire PRA Training, Sept. and Nov., 2008

Introduction and Overview A Collaboration of U.S. NRC Office of Nuclear Regulatory

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

Notes:

2-47
Slide 58

IThe SNPP: Secondary Systems P&ID

..... .. ....... . . ..... -

A Collaboration of U.S. NRC Office of Nuclear Regulatory .... ........................ Research (RES) & Electric Power Research Institute (EPRI) Fire PRA Training, Sept. and Nov., 2008 Introduction and Overview

Notes:

2-48
Slide 59

I The SNPP: Electrical One-Line Diagram

swG

OF-STPOWER

Fire PRA Training, Sept. and Nov., 2008

Introduction and Overview Slide 59 1 A Collaboration of U.S. NRC Office of Nuclear Regulator, L Research (RES) & Electric Power Research Institute (EPRI)

Notes:

2-49
Slide 60

The SNPP: General Plant Layout - Plan

S-------------------------- STRUCTURE I

"• ~ ~SECUR"T~~G

SWITCH O Lg

YARD

TURBINE BLDG.

.AA

Intodctonan OvrvewI .......... ........... ........... Research (RES) & Electric Power Research Institute (EPR/)

Notes:

2-50
Slide 61

The SNPP: Plant Layout - Elevation

Containment and Auxiliary Building

MAIN CONTROL ROOM

CABLE SPREADING ROOM

4 Ad

+2 SWITCHGEAR ROOM

4 ~2

OFT

5FT

OFT

0OFT

FT GRADE

FT

CHARGING PUMP ROOM

+'0

RHR PUMP ROOM

,-20

~-Th

Fire PRA Training, Sept. and Nov., 2008

Introduction and Overview Slide 61 A Collaboration of U.S. NRC Office of Nuclear Regulatory 1. Slide... Research (RES) & Electric Power Research Institute (EPRI)

Notes:

2-51
Slide 62

I The SNPP: Aux. Bid. - RHR Pump Room

R X -• NOTES:

1. VERTICAL PIPE PENETRATION

TO UPPER ELEVATION.

2. PENETRATION TO UPER FLOOR

IS SEALED.

N CBT 0

(NOTE 2) HCBT: HORIZONTAL CABLE TRAY

VCBT: VERTICAL CABLE TRAY

Fire PRA Training, Sept. and Nov., 2008 Introduction and Overview i.Slide 62 I A Collaboration of U.S. NRC Office of Nuclear Regulatory

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

Notes:

2-52
Slide 63

I The SNPP: Aux. Bid. - Charging Pump Rm.

NOTE:

1. VERTICAL PIPE PENETRATING

THE FLOOR.

VCBT: VERTICAL CABLE TRAY

Collaboation eA of U.S. NRC Office of Nuclear RIgulatoty

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

Fire PRA Training, Sept. and Nov., 2008

Introduction and Overview

Notes:

2-53
Slide 64

Notes:

2-54
Slide 65

L.

The SNPP: Aux. Bid. - Cable Spreading Rm.

VCBT +40A _

HCBT +40A-'

-1 FM

-- VCBT +408

'NHCBT +50B

HCBT: HORIZONTAL CABLE TRAY

VCBT: VERTICAL CABLE TRAY

Fire PRA Training, Sept. and Nov., 2008

Introduction and Overview PSlide 65 I A Collaboration of U.S. NRC Office of Nuclear Regulatory

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

Notes:

2-55
Slide 66

The SNPP: Aux. Bid. - Main

MAIN CONTROL ROOM

KICE SHIFT

KITCHENSUPERVISOR

OFFICE

Fire PRA Training, Sept. and Nov., 2008 i Slide 66 i

Introduction and Overview Sie6 n ot..........................e.s............

Notes:

2-56
Slide 67

I The SNPP: Main Control Board Layout

IELECTRICA TURBINE AND:I~

MAW FEEDI

A :'(

RX CONTROL. H IP W

III*I w.

TI-i 1.-- RHR PUMP

LI-2 - \ \ -MOV-5

L1-1',,- r / MOV-6

PUMP HPI-A -L8 I] u..- MOV-3

AOV-2 0 0 MOV-2 AOV-3

MOV-7 MOV-8

DETAIL FROM

ABOVE

Fire PRA Training, Sept. and Nov., 2008

/ntioduction and Overview Slid 6.'7 i A Collaboration of U.S. NRC Office of Nuclear Regulatory

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

Notes:

2-57
Slide 68

I The SNPP: Turbine Building

Fire PRA Training, Sept. and Nov., 2008

Introduction and Overview . i .. A Collaboration of U.S. NRC Office of Nuclear Regulatory i lde6 Research (RES) .& Electric Power Research Institute (EPRI)

Notes:

2-58
Slide 69 Notes:

PART II

EPRI and RES Fire Research Activities

Relevant to Fire PRA

Topics:

The ANSIASME FPRA Standard

The RES-EPRI Collaboration on Fire HRA

Rm rh(RES) A El'M Pow -et~ ýt l (EPRO

Slide 70 Notes:

The ASME/ANS Fire PRA Standard (briefly..)

SeR a 2w A CoSaboIA RO UDS NRC Sfl2. d - R.•Yat

R-Sou Ro R R .- R2-59
Slide 71 Notes:

! The Fire PRA Standard

* The ANS Fire PRA Standard has been published

* The technical content has been adopted in whole into Part III of the

ASME/ANS PRA standard

- Pending as a proposed revision to RA-S-2002 (RA-S-200X)

* NRC issued Draft for Comment R.G.1.200, Rev. 2, June 2008

* Be aware that some areas of the standard have evolved beyond the

guidance of NUREG/CR-6850 - EPRI TR 1011989

* Particular areas to note:

- Quantitative screening criteria

- Treatment of concurrent spurious actuations

- Instrument cable tracing (to verify diversity/availability)

P .•RA71 .SUS F A - rOvca RE RbXfrUeaUch (RES) &E. c p R.s..ch mao.- (F-RO

2-60
SESSION 4: Fire HRA Guidelines Overview

2-61
Slide 1 Notes:

UR" E I' I'2

Joint EPRI / NRC

Fire Human Reliability Analysis

Guidelines Overview

Susan Cooper, USNRC

EPRI-NRC Fire PRA Training,

September and November, 2008

A CohbO,.tion of U.S. NRC Office of Noudea Regalino"y ReA.MIh (RES) & EMOctric Power Re-s.th Insfftuft (EPRIO

1

Slide 2 Notes:

<*-RC Presentation Outline

• Project Overview

- Team

- Tasks

* Technical Overview

* Status

(R -) f-) 2

2-62
Slide 3 Notes:

<*UM- RC Problem Statement

* NUREG/CR-6850 addresses screening, with

conservative data, but not detailed fire HRA

* 6850 Fire HRA approach applied at Nine Mile

Point (but not released)

* Over 40% of the USA plants are transitioning

to NFPA-805, 3-year time window to complete

transition

-- * AC°b~.tkdU.&`NRCO e~u~Rgu°R N~(RESt&E J~cP s h•R t•hk~sfueEPO 3

2-63
Slide 4

q*U.S.NRC Fire HRA Team

EPRI NRC

SFrank Rahn (Bob Kassawara) 1 Susan Cooper (Kendra Hill)

EPRI Project Manager p NRC-RES Project Manager

EPRI RES

J. Julius & B. Najafi John Forester (Lead)

Jan Grobbelaar & K. Kohlhepp Susan Cooper

G. William Hannaman Stacey Hendrickson

Independent Review Team **

NRC Reps

Utility HRA Reps

S-• A Collaboration of U.S. NRC Office of Nuclear Regulatory Research (RES) & Electric Power Research Institute (EPRI) 4

Notes:

2-64
Slide 5 Notes:

Project Overview

1) Data collection

* Fire events (reviewed historical event data).

* Considered a range of plant responses (fire response strategies).

* Some plant interviews (with more during the testing phase).

2) Method Development

* Started with a review of the EPRI Fire HRA Guideline.

* Updated NUREG/CR-6850 screening for long time windows.

* Adding a Fire HRA Scoping (flowchart) method.

* Detailed Fire HRA based on EPRI approach or ATHEANA.

3)Testing

4) Documentation

-- Joint NUREG/CR & EPRI report, similar to NUREG/CR-6850

A Codtotk s U.& NRC O dNNýI, R. 9 oyWW ec (ýRES) &Ekcr ,ower NRRA(EPRO 5

Slide 6 Notes:

, U .NRC Fire HRA Guidelines:

Objectives and Scope

* Identify/analyze existing post-initiator HFEs

* Identify/analyze post-initiator fire response (App R) HFEs

- Includes Main Control Room abandonment.

* Identify/analyze post-initiator undesired spurious HFEs

* Implement post-initiator fire HEPs in fire PRA model(s)

- First quantification/screening and/or detailed fire PRA model.

- Including dependency analysis.

* Out of Scope

- Pre-initiators (per NUREG/CR-6850).

- Fire brigade response.

• -- . A ob U.S NRC •O *dN - Re, A -Rmtao r t• (S) &EMA(EPPoRnrRnmh 6

2-65
Slide 7 Notes:

<US.NRC Fire HRA Approach

* Examine HRA process, see how the process and tasks

would change in a post-fire environment or post-fire

accident response scenarios

* Done for each of the 3 categories of post-fire actions

- Existing post-initiator HFEs

- Post-initiator fire response (App R) HFEs

' Includes Main Control Room abandonment.

- Undesired response to spurious HFEs

* Progressive fire HRA tasks reflect fire PRA development

- Rough, quantitative screening per NUREG/CR-6850.

- Scoping fire HRA approach for scoping fire models.

- Detailed fire HRA using EPRI or NRC methods (joint project).

• -- , A Cotlabb,~n S. NRC Of Fit Nulear Rg~ Y (RES) S E fletric P~erma OIAA (EPRO 7

Slide 8 Notes:

*USjNRC Existing Post-lE HFEs in

Base PRA

* Identification

- Base PRA HFEs modeled in fault & event trees.

- Those portions use to develop the plant response to fire-induced

initiating events.

* Definition - HFE modeling modified based on fire impacts:

- Cues and indications may be inaccurate.

- Time available may be reduced.

- Response time may increase.

- Workload may be high.

- Crew credited for recovery may not be available anymore.

- Stress may be higher.

- Accessibility may be impeded due to smoke, heat.

- Communication may be impacted.

• -• A ColaboSto dEtlS. NRC 055. ofN~tk.,r IguAto•y R.,..h (RES( & EleofriC posr R ch 5550,ut (EPRI) 8

2-66
Slide 9 Notes:

CgU&NRC Fire Response

Operator Actions

Required in response to a fire, as directed by the fire

procedure/s e.g.

- Mitigate or prevent fire damage to equipment

- Recover existing internal events operator actions

- Mitigate undesired operator actions in response to spurious indications or actuations

- Abandon main control room and perform safe shutdown outside

the main control room

* Identification process can be

- Iterative as required in fire PRA

- Comprehensive based on fire procedure/s

* Similar definition considerations, but procedures, training,

and cues can be significantly different that for existing.

Example HFEs on next slide

S A h~A oýdUS. NRC OlýarR-oyReýft eRES) &EMPon,eý EPRo 9

Slide 10 Notes:

,USNRC Fire Response Action

Examples

* Identify protected instrumentation channels

(to mitigate spurious indications)

* Defeat solid state protection system (to prevent spurious

safety injection)

* Control auxiliary feedwater locally by throttling valves

manually and starting / stopping pumps

* Place back-up indication panels in service

* Obtain steam generator level locally

* De-energize all ADS valves

* Close HPCI steam supply valve locally

* Align 4 kV bus by locally operating breakers

A CO nCOA dUS.NRC Numc earcMRendu~ataryOSCeRRES) & EMCbP~rRem~fltReAO(E•RR 10

2-67
* A definition of an und

intentioned operator

the scenario

- Operators are trained

follow their procedure

* To be identified within

progression

- Emergency operating

- Annunciator response

A - U&. NCýr

Slide 11 Notes:

Undesired Response to

Spurious Operations

esired operator action is a well

action that unintentionally aggravates

to (1) believe their instrumentation and (2) to

s

n the context of the accident

procedures

procedures

SNlRdeay e (RES) &ENotes

Slide 12 Notes:

"*u'yRc" I Fire HRA Guideline

Summary

1. Standard HRA process used for Fire HRA modeling:

Based on other processes (e.g., NUREG-1792, EPRI Fire HRA

Guidelines, SHARP1, ATHA NA)

2. Fire HRA steps:

- Identification & definition of human failure events (HFEs):

Substantial guidance provided, including feasibility test

Feasibility Evaluation (Go / No-Go) example criteria - Sufficient time avai!able to complete action - Sufficient manpower - Procedures & cues exist

- NUREGICR-6850 screening:

Refinement/relaxation for areas identified in NUREG/CR6850 implementation

- Qualitative HRA analysis activities described:

* Certain activities required for all analyses; others only for

specific detailed HRA methods

A C -hdanU.S. NRC Off detRwa-r yRn h (RES) A S'AC P°• Reach inE- ý (ECR) 12

2-68
Slide 13 Notes:

< ]'_S•.RC Fire HRA Guideline

Summary (cont.)

2. Fire HRA steps: (continued)

- Scoping Fire HRA method added (new):

Developed to address the majority of HFEs, thereby

conserving HRA resources

Guidance being developed to aid reproducibility & ability to

review

Can be used for defined scenario contexts that are

generalized & constrained with respect to PSFs, etc.

Current format: decision trees

- Used existing methods for detailed Fire HRA, with

performance shaping factors modified for Fire PRA:

EPRI Cause-Based Decision Tree & HCR/ORE; plus THERP

ATHEANA

• -i A C °fU-t 00. 303 OttO dflcttR~g• Rn (REStE) tP0 Ot an e(E5R 13

Slide 14 Notes:

<*,USNRC, Qualitative Analysis

includes....

Performance shaping factors that may

impact specific fire scenarios, such as:

- Timing (primarily increased response times)

- Instrumentation (i.e., cues for action & potential

unavailability or spurious actuation)

- Stress

- Procedure quality

- Workload

- Complexity

A C°#t.'.ta°,Wx U S" A/RC 000 ON of Nuclear Ra~uflOby Rench (RESt A 300c.'00 0 (EPRIO 14

2-69
Slide 15 Notes:

<UMNRC Quantification Methods

Fire HRA quantification method can vary, commensurate

with the Fire PRA task where it is modeled, such as:

- NUREG/CR-6850 HRA screening for first quantification

- Revised NUREG/CR-6850 to reduce conservatism for long term

actions in sets 1 and 2 (for actions after the fire is out).

- Scoping Fire HRA method added as fire PRA iteratively quantified to

refine the quantification of areas

- Detailed Fire HRA for more detailed scenarios

* Using existing methods with performance shaping factors

modified for Fire PRA

- ATHEANA

- EPRI Cause-Based Decision Tree & Human Cognitive Reliability /

Operator Reliability Experiments

• -- , A C--tofUS N&nC •OMfNunRgýyRe.A hRES) AEedcp.,R.,e..hAM~m.e(EPR 15

2-70
Slide 16

Scoping HRI U.S.NRC HFEs in the New & Existing

HFES In MCR

Derronstrated

feasibility

All necessary &

sufficient cues & Y controls for rd Yes

actions are not

irripacted by fire.

No

Need to c onsider

spunousIn idicators

OO t. bot nh Unk is bColeO 

S purInd IdicatrrUnk own ch enrate Id cto Floy

Transfer to Spunous Transfer to Spur

F o wnia cart indicator Flowcl

frEC for bOO,

A Collaboration of U.S. NRC Office of Nuclear Regulatory Rese

Notes:

•,Analysis for New & Existing VlCR (part one)

In MCR Actions HFE

(transfer from MCRI

Abandonmnent

Ti- OfTransfer to

the fire effects New HIFE: HEP = I E-3

have ended Existing HFE: existing

>=100%

so- 99%

0- 49°9Eitn F :ices,

1.0o HEP by o

nous

earch (RES) & Electric Power Research Institute (EPRI) 16

2-71
Slide 17

• [,•.=,_•Scoping HRW

.NARC HFEs in the

D thef to , I , >=100 poe - Yes hazardus ýeffhect Yes $CBA Ye!

F--a_---(-p,1 NoNewN HFE: HEP - 0.0-3

E.ting HFE : existing HEP

W atis D theSmoke cr other

50-99 pr rs Yes Yes SB the tinnehazardous effect SI

M rgi? h (e.g., toxic gas) neequired?

s a sin MGR?

NO , I

HE--=1.0 (L e H........... - xtin, HIE: Inc"•

E~tlngHFE: i~creas HEP by facto HRb of atro, 

25-49% prcdue hazrdus effect E 1.

snratnch the(g,n 9 toxc gas}

NO No

Nt. HFE: HP= 0.0

•_.Bexst ing HFE: ncrae-

-- HEP by factor df SO

• :A ----- Collaboration of U.S. NRC Office of Nuclear Regulatory Rese

Notes:

Analysis for New & Existing

OCR (part two)

earch (RES) & Electric Power Research Institute (EPRI) 17

2-72
Slide 18 Notes:

<*UaNRC Joint NRC / EPRI

Fire HRA Schedule

-Started March 5, 2007

- Meetings held June, July, Sept & Nov'07

- First integrated draft May 2008

- 1st peer review in June 2008

-Testing at 2 plants (in progress)

- 2nd peer review planned

- Public comment period planned for 2008/2009

- Publication scheduled for Summer 2009

A ý Cb dU.S NRC O .WNu de Rguaý y RN.h ý S, AE~a• Pa R(RES) 18tEPRO 18

Slide 19 Notes:

-LU&NRC Fire HRA Peer Review

& Testing

Peer Review Phase-I: NRC and industry team with 8 reviewers

* In general, was the right approach taken and implemented? - Is the technical approach sound and reasonable?

- Are the selected HRA models appropriate for the application?

- Are the assumptions presented in this methodology reasonable?

- Does the guidance meet its stated objectives?

Testing

" Plant 1: Conducted in August, tested the flowcharts

* Plant 2: Scheduled for September

Peer Review Phase-2:

* Phase-2 to be conducted after testing - Were the issues identified during the phase one review and testing addressed? - After addressing any issues does the method still meet the Phase-1

criteria?

- a ~ta.,aJS RC RC. U, R., IC ICSR,.CC R~s a RCNC RC,,C CC~flCCQ 19

2-73
Slide 20 Notes:

_S_.N.RC Fire HRA Guideline

Summary

1. Standard HRA process used for Fire HRA modeling

2. Developing Fire HRA methodology & guidelines:

- Relaxation added to NUREG/CR-6850 screening

- Scoping Fire HRA method added (new)

- Used existing methods for detailed HRA, with

performance shaping factors modified for Fire PRA

* ATHEANA

' EPRI Cause-Based Decision Tree & HCR/ORE

3. Test methods & guidelines:

- Screening HEPs as well as HEPs supporting detailed

fire scenarios

4. Coordinated between the NRC and Industry via the

EPRI HRA Users Group A CO#.hOtUCn Of US. OU NRC II of fudar feguatdy R - (RESC & EedcbtPo Rosm INtw, (Em) 20

2-74
MODULE 1:

PRA/HRA (SYSTEMS ANALYSIS)

INTRODUCTION

With the adoption of NFPA 805 as an acceptable alternative to existing standards for fire protection in

NPPs, a systematic approach to performance-based analysis was developed for the quantification of risks

associated with fire. An integrated approach to conducting a fire PRA was necessary to combine accident

response modeling in response to a fire (including manual actions) and the frequency and progression of

fire with fire protection features and systems. PRA is also highly iterative to enable successive refining of

results as needed. The seven sessions related to the PRA/HRA (Systems Analysis) module in this part of

the NUREG/CP report provide an introduction to many important elements in this risk quantification.

As discussed in the preceding "OVERALL COURSE" part of this report, the PRA/HRA Module includes

Technical Tasks 2, 4, 5, 7, 12, 14, 15, and 16 from NUREG/CR-6850. All of these tasks (except Task 16,

"Fire PRA Documentation") are discussed in the MODULE 1 SESSIONS summarized below and

presented in detail in the remainder of this part of the NUREG/CP report.

SESSION 1 a discusses Task- 2, "Fire PRA Component Selection," in which the scope of the particular

PRA is defined. A procedure is developed for choosing particular plant components to be modeled,

generating a list that forms the basis of subsequent systems, electrical, and fire tasks. This task defines

the scope of the selected components including equipment that could mitigate core damage, equipment

that could cause initiating events, and equipment that could cause spurious actuations.

SESSION lb presents the exercises for all seven of the Task 2 steps as discussed in Section 1.6.2,

"Example Exercises," in the first part of this Volume 1, "OVERALL COURSE," in its Section 1,

"COURSE OVERVIEW." Appendix B of this Volume 1 presents the instructor's solution for these

exercises.

SESSION 2 discusses Task 5, "Fire Induced Risk Model Development." This task discusses

development of a logic model that reflects plant response following a fire. Specific instructions are

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).

SESSION 3 discusses Task 4, "Qualitative Screening," and Task 7, "Quantitative Screening." Task 4

identifies fire analysis compartments that can be shown to have little or no risk significance without

relying on a quantitative risk analysis of those compartments. Fire compartments may be screened out if

they contain no components or cables identified in Task 2 and Task 3', and if they cannot lead to a plant

trip because of plant procedures, an automatic trip signal, or technical specification requirements.

Qualitative screening decreases the work required in later parts of the PRA. Task 7, "Quantitative

Screening," discusses an approximate evaluation of quantitative risk associated with each compartment to

determine which compartments screen out (i.e., for which detailed analyses will not be performed).

1 Task 3 from NUREG/CR-6850 is "Fire PRA Cable Selection," in which electric cables are identified that support

components selected in Task 2. Task 3 is discussed in detail in Volume 2 ("Module 2: Electrical Analysis") of this

NUREG/CP report.

3-1
However, approximate (conservatively bounding) residual risks for "screened out" compartments are

retained in the final detailed quantification (i.e., Task 14) to ensure that fire risk is not underestimated.

SESSION 4 discusses Task 12a, "Screening Post-Fire HRA," in which human interactions with the plant

components vulnerable to fire risk are considered. The task provides structured instructions for

identification and inclusion of these actions in the fire PRA. It also provides instructions for estimating

screening human error probabilities (HEPs) before detailed fire modeling results (e.g., fire growth and

damage behaviors) or detailed circuit analyses (e.g., can the circuit spuriously actuate as opposed to

simply assuming it can) have been developed. It does not develop a detailed fire HRA methodology. A

number of HRA methods can be adopted for fire with appropriate additional instructions that superimpose

fire effects on any of the existing HRA methods such as THERP, CBDT,ATHEANA, etc. This would

improve consistency across analyses (i.e., fire and internal events PRA).

SESSION 5 discusses Task 12b, "Post-Fire HRA Detailed Analysis," in which nonscreened HRA-related

contributions to the fire risk from plant ignition sources are quantified. For example, it discusses

performance shaping factors that should be considered such as staff resources available at the time of the

fire, applicability and suitability of their training and experience, suitability of relevant procedural and

administrative controls, availability and clarity of instrumentation, time available vs. time needed to

complete actions, the environment in which the actions are to be performed, and the accessibility and

operability of needed equipment.

SESSION 6 discusses Task 14, "Fire Risk Quantification," in which risks are determined for non

"screened out" compartments, and inputs from all other Tasks are compiled (e.g., the residual risk from

screened-out compartments from Task 7, etc.).

SESSION 7 discusses Task 15, "Uncertainty and Sensitivity Analysis," where treatment may vary from

identification of sources of uncertainty to quantitative estimation and propagation of uncertainties. The

treatment may also include one-at-a-time variation of individual parameter values to determine the effect

on the overall fire risk (sensitivity analysis).

3-2
Session 1 a: Fire PRA CompiOnent Selection

3-3
Notes:

Slide 1

A,=~

mri f I :FRSEARCH INSTITUTE

EPRIINRC-RES FIRE PRA

METHODOLOGY

Task 2 - Fire PRA Component Selection

Joint RES/EPRI Fire PRA Workshop

September and November 2008

Bethesda, MD

I A Coflaboration of U.S. NRC Offic of Nuclear Regulatory Resea..h (RES) & Elctrc Power Research Institute (EPRI)

Slide 2 Notes:

I Component Selection

Purpose

Purpose:, describe the procedure for selecting plant

components to be modeled in a Fire PRA

Fire PRA Component List

- Key source of information for developing Fire PRA

Model (Task 5)

- Used to identify cables that must be located (Task 3)

Process is iterative to ensure appropriate agreement

among fire PRA Component List, Fire PRA Model, and

cable identification

FbV PRA WSh•p. Behesda MO. 2008 F- 2 A C!On US NRC Ohh a oute ReI•-ofy

Task 2: Co= rn( SaeN Oh Reashf (REhS & Petrc RhSOah his e (C? Rd

3-4
Slide 3 Notes:

! Task 2: Fire PRA Component Selection Scope

Fire PRA Component List should include following major categories of equipment:

* Equipment whose fire-induced failure (including spurious actuation) causes an

initiating event

- Model initiating events, not the specific equipment

- Need to identify worst-case initiator for each compartment

* Equipment needed to perform mitigating safety functions and to support

operator actions

* Equipment whose fire-induced failure or spurious actuation may adversely

impact credited mitigating safety functions

* Equipment whose fire-induced failure or spurious actuation may cause

inappropriate or unsafe operator actions

* Recommended to include passive systems in list even though they are not

susceptible to fire

Fre PEA WOMa•hon. 5.e,uda MD, 2008 A--J A US)NRC ONW-R"

Ta-k Z Gnp n t St nesc et R (RES) & Ec Pe Reh (EM)

Slide 4 Notes:

! Component Selection

Approach

* Step 1: Identify Internal Events PRA sequences to include in fire PRA Model

(necessary for identifying important equipment)

* Step 2: Review Internal Events PRA model against the Fire Safe Shutdown

(SSD) Analysis and reconcile differences in the two analyses

* Step 3: Identify fire-induced initiating events based on equipment affected

* Step 4: Identify equipment subject to fire-induced spurious operation that

may challenge the safe shutdown capability

* Step 5: Identify additional mitigating, instrumentation, and diagnostic

equipment important to human response

* Step 6: Include "potentially high consequence" related equipment

* Step 7: Assemble the Fire PRA Component List

Firm PRA Workshop, Bethesda MD, 2508 $ . , A C -bo.o I d U.S. Nnh Etio oh* udNýW Rý4-y

Trsk 2: Comwoenh neoetn Re..h (RhO) &El- P.cp - Res~h nst-tt (EPRIO

3-5
Slide 5 Notes:

Component Selection

General Observations

* Two major sources of existing information are used to generate the Fire PRA

Component List: * Internal Events PRA model

* Fire Safe Shutdown Analysis (Appendix R assessment)

* Just "tweaking" your Internal Events PRA is probably NOT sufficient - requires

additional effort - Consideration offim-induced spurious operation of equipment - Potential for undesirable operator actions due to spurious alarmshndications - Additional operator actions for responding to fire (e.g., opening breakers to prevent spurious

operation)

* Just crediting Appendix R components may NOT be conservative

- True that all other components in Internal Events PRA will be assumed to fail, but:

* May be missingcomponents with adverse risk implications (e.g., event initiators or complicate SSD response)

SMa p miss effects of non-modeled components on credited (modeled) systemstcomponents and on operator performance

* Sill need to consider non-credited components as sources of fires

Fue PRA Ww*or , Be thva MD, 2008 g A Cu Ia - ;°nefU-S Ns c ae N r ,•r Rruo a tar Y

Toak 2: Cý_ - sRovro (RE= r ) 0- (EPRI)

3-6
Slide 6

Notes:

3-7
Slide 7 Notes:

! Task Key Assumptions 2: Fire PRA Component Selection

The following key assumptions underlie this procedure:

* A good quality Internal Events PRA and App. R Safe Shutdown Analysis

are available

* Analysts have considerable collective knowledge and understanding of

plant systems, operator performance, the Internal Events PRA, and App.

R SSD analysis

* Steps 4 thru 6 are applied to determine an appropriate number of

spurious actuations to consider

- Configurations, timing, length of sustained spurious actuation, cable

material, etc. among reasons to limit what will be modeled

- Note that HS duration is a current FAQ topic...

R e PRA W tks Bethe ,p ,da S MD. 20 08 h A Cd boO LB o o U& N RC OfficefN u- R A -ut oy

rTs, 2: Cý Men: Senosn -R-h (RES) I Eett P _ _.mh -rfa (EPR-0

Slide 8 Notes:

F In-process rom: Lessons FAQs Learned ... and Insights

FAQ 08-0051

- Issue:

* The guidance does not provide a method for estimating the

duration of a hot short once formed

* This could be a significant factor for certain types of plant

equipment that will return to a "fail safe" position if the hot short is

removed

- General approach to resolution:

* Analyze the cable fire test data to determine if an adequate basis

exists to establish hot short duration distributions

- Status:

* Initial data analysis has been completed and results are under

team review

* Staff and industry review pending

* Final revisions, as needed, pending - remove

Fe PRA Worshop, Bethesda SD, 2008 L A CdDOon odV.S NRC Ofl N&PROARy Task 2; compo~nen Selection RSee~ch (RED) Eecri F Pm. R~e.sdm I05tuhe (EPRRS

3-8
Slide 9 Notes:

Task 2: Fire PRA Component Selection

Inputs/Outputs

Task inputs and outputs:

* Inputs from other tasks: equipment considerations for

operator actions from Task 12 (Post-Fire HRA)

* Could use inputs from other tasks to show equipment does

not have to be modeled (e.g., Task 9 - Detailed Circuit

Analysis or Task 11 - Fire Modeling to show an equipment

item cannot spuriously fail or be affected by possible fires)

* Outputs to Task 3 (Cable Selection) and Task 5 (Risk

Model)

* Choices made in this task set the overall analysis scope

IM PRa CQ Wo, -Uf B.n M, 2008 * Cont6p NRC OT oNudW RgASEN Tea? : O~pnn .Seco Rea? IRESI & B. c PwR0Ite• • rau (OCR?

Slide 10 Notes:

! Task 2: Fire PRA Component Selection Steps In Procedure/Details

Step 1: Identify sequences to include and exclude from Fire

PRA

- Some sequences can generally be excluded - Sequences requiring passive/mechanical failures that can not be initiated by

fires (e.g., pipe-break LOCAs, SGTR, vessel rupture)

- Sequences that can be caused by a fire but are low frequency (e.g., ATWS)

- It may be decided to not model certain systems (i.e., assume failed for Fire

PRA) thereby excluding some sequences (e.g., main feedwater as a mitigating

system not important)

Possible additional sequences (recommend use of expert panel to

address plant specific considerations) - Sequences associated with spurious operation (e.g., vessel/SG overfills,

PORV opening, letdown or other pressure/level control anomalies)

- MCR abandonment scenarios and other sequences arising from Fire

Emergency Procedures (FEPs) and/or use of local manual actions

FPe PRA W ., SePh00a MD, 2008 A C 1 d US. NRC ORCedNu..,Reuefoy

Tea 2: C,-,,St , SlcinRelt (Su Dec0n 0 PZ 0 e-tAft R.SA- (EPRO

3-9
Task 2: Fire PRA C

Steps In Procedure/D

Step 2: Compare Internal Eve

Identify and reconcile:

- differences in functions,

no feed/bleed; PRA-feed

- front-line and support sy

PRA-do not need HVAC

- system and equipment d

considerations (e.g., App

- other miscellaneous equ

Include review of manual ac

in conjunction with Task 12

FR PRA Workshop, B.the.aA MD, 2008

Taýk 2. Componnt Selection

Slide 11 Notes:

3omponent Selection

'etails

*nts PRA model to App. R SSD list

success criteria, and sequences (e.g., App. RI/bleed)

stem differences (e.g., App. R-need HVAC;

ifferences due to end state and mission

p. R-cold shutdown; PRA-hot shutdown)

ipment differences.

tions (e.g., actions needed for safe shutdown)

A Cflfl~bfti flWU S. NRC OCAce of Nucannr R Afo

Raaaah (RES) A 80,15 Pswen = = a nt=oa (=ePR)

Slide 12 Notes:

! Task 2: Fire PRA Component Selection Steps In Procedure/Details

Step 3: Identify fire-induced initiating events. Consider:

* Equipment whose failure (including spurious actuation) will cause

automatic plant trip

* Equipment whose failure (including spurious actuation) will likely result in

manual plant trip, per procedures

* Equipment whose failure (including spurious actuation) will invoke Tech.

Spec. LCO necessitating a forced shutdown while fire may still be present

(prior EPRI guidance recommended consideration of <8-hr LCO)

* Compartments with none of the above need not have initiator though can

conservatively assume simple plant trip

FIf PRA Workshop. BethesdaMD, 2008 S 12 dUS Nn M . ,., PtWto

Task 2: componnt PR~ath -e-ec-on (PES) . pP R-e- - a ,St- - (Ep-I)

3-10
Slide 13 Notes:

! Task 2: Fire PRA Component Selection Steps In Procedure/Details

* Since not all equipment/cable locations in the plant (e.g., all BOP) may

be identified, judgment involved in 'likely' cable, paths

- Need a basis for any case where routing is not verified

- Routing by exclusion (e.g., from a fire area, compartment,

raceway...) is a common and acceptable approach

* Identify worst-case initiator based on possible initiators and other

mitigating equipment likely to be affected

* Should consider spurious event(s) contributing to initiators

Fke PRA Wo,&WW, Beesa MD. 2008 1o U. I R;-;;

Task 2: CGwnPo S00RCtn (RWS) IeP a- 0S -2s 1 (EPlI)

3-11
Slide 14

Task 2: Fire PRA Component Selection

Steps In Procedure/Details

Instrument

Air

Compressor

Cables judged

to be here

Fires cause loss of

instrument air Fires assumed to cause loss of

instrument air

Fires assumed to cause loss

of MCC(s) & subsequent

effects (including loss of

instrument air)

Fire PRA Workshop, Bethesda MD, 2008

Task 2: Component Selection

14 .Slide A Collaboration of U.S. NRC Office of Nuclear Regulatory

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

Notes:

3-12
Slide 15 Notes:

! Task 2: Fire PRA Component Selection Steps In Procedure/Details

Step 4: Identify equipment whose spurious actuation may

challenge ability to avoid core damage/large early release

Consider multiple spurious events within each system

considering success criteria

* Involves review of system P&IDs and other drawings

• Focus on equipment or failure modes not already on the

component list (e.g., flow diversion paths)

- Any new equipment/failure modes should be added to

component list for subsequent cable-tracing and circuit

analysis

FR PRA WoWWW. Be AI MD. 2008 SI- 5 A db U.S. NWU N RgA y

T.k 2. Cr-r bs . (RES) S BflýcpbaRlbsn mt&t (EPRO

3-13
Slide 16

! Task 2: Fire PRA Component Selection .Flow Diversion Path Examples

from main hý A hý A

flowpath P, "N V 'N

Div A MOV Div B MOV

to diversion

path

takes 2 spurious

hot shorts to

open diversion

path

Included in model

from main

flowpath

CheckValve

to diversion

path

takes I spurious

hot short &

failure of check

valve to open

diversion path

Screened from model

if not potential high

consequence event

Fire PRA Workshop, Bethesda MD, 2008

Task 2: Component Selection

Slide 16 A Collaboration of U.S. NRC Office of Nuclear Regulatory -Research (RES) & Electric Power Research Institute (EPRI)

Notes:

3-14
Slide 17

Task 2: Fire PRA Component Selection

Example of a New Failure Mode of a Component

App. R ensures MSIVs

will close / remain closed

Containment so as to isolate vessel1

Reo Main Steam Line

Vessel Inboard MSIý Outboard MSIV

SFe PRA concerned with

MSIVs closing / remaining

closed AND will not

spuriously close when want

valves to remain open so as to

use condenser as heat sink'

different cables and corresponding

circuits and analyses may need to

be accounted for

Fire PRA Workshop, Bethesda MD, 2008 Slide 17 A Collaboration of U.S. NRC Office of Nuclear Regulatory

Task 2: Component Selection Research (RES) & Electric Power Research Institute (EPRI)

Notes:

3-15
Slide 18 Notes:

! Task Steps 2: In Procedure/Details Fire PRA Component Selection

Step 5: Identify additional instrumentation/diagnostic equipment important

to operator response (level of redundancy matters!)

* Identify human actions of interest in conjunction with Task 12

* Identify instrumentation and diagnostic equipment associated with

credited and potentially harmful human actions considering spurious

indications related to each action

- Is there insufficient redundancy to credit desired actions in

EOPs/FEPs/ARPs in spite of failed/spurious indications?

- Can a spurious indication(s) cause an undesired action because action is

dependent on an indication that could be 'false'?

- If yes - put indication on component list for cable/circuit review

* Watch for new/expanded guidance to be developed by the RES/EPRI

fire HRA collaboration...

Fk. PRA W.S, Bheoda 2O. 2008 s 1 A CRdRo U.& NRC OtM•N'e-RegatoY TI~s 2: Compoj'ent s.MCCo,, • (RES & E•k Po~ ,RUNR h Ntsk EPRR

Slide 19 Notes:

! Task 2: Fire PRA Component Selection Steps In Procedure/Details

Guidance on identification of harmful spurious operating

instrumentation and diagnostic equipment:

* Assume instrumentation is its normal configuration

* Focus on instrumentation with little redundancy

- Note that fire PRA standard has language on this subject (i.e.,

verification of instrument redundancy in fire context)

* When verification of a spurious indication is required (and reliably

performed), it may be eliminated from consideration

* When multiple and diverse indications must spuriously occur, those

failures can be eliminated if the HRA shows that such failures would

not likely cause a harmful operator action

* Include spurious operation of electrical equipment that would cause a

faulty indication and harmful action

* Include inter-system effects

Fim PRA WNýshop, .tesda MD, 2W)8 L A (RESo & U.RC OR o. N O .Rut

Task 2. CRomPRn SToR E&RE

3-16
Slide 20 Notes:

Task Steps 2: In Procedure/Details Fire PRA Component Selection

Step 6: Include "potentially high consequence" related equipment

* High consequence events are one or more related failures at least

partially caused by fire that:

- by themselves Cause core damage and large early release, or

- single component failures that cause loss of entire safety function and lead

directly to core damage

* Example of first case: spurious opening of two valves in highpressure/low pressure RCS interface, leading to ISLOCA

* Example of second case: spurious opening of single valve that drains

safety injection water source

FRe PRA WoS*tmp. eSdhka MO. 2008 5 A CO ýSI U - 5 500r °

Tak 2, ConyzeM S.Mettio. R fth (RES) & Ra* ltiý. fEPRO

Slide 21 Notes:

Task 2: Fire PRA Component Selection

Steps In Procedure/Details

Step 7: Assemble Fire PRA component list. Should include

following information:

* Equipment ID and description (may be indicator or alarm)

* System designation

* Equipment type and location (at least compartment ID)

* PRA event ID and description

* Normal and desired position/status

* Failed electrical/air position

* References, comments, and notes

Fire PP. Wortshop, Bethesda MD, 2008 Sd 2 ctaboat o, Uis. NRC 0IW ( dtEU-k)Ratoy

Task 2; Componnt Steecton MEPt Roe.ert tPESJ SEtettc FoaR th r#tef 8 

3-17
! Sample Problem E

* Distribute blank han

* Distribute completed

* Question and Answe

Fir PRA WoýrShop, Beftida MD, 2008

TaSk 2: GomRoRM SNCfl.

Slide 22 Notes:

:xercise for Task 2, Step I

lout for Task 2, Step 1

handout for Task 2, Step 1

r Session

fl A CallbSotjnolf, d N• NRC~ oRC dkuda Re~dfoV

Re$oach (RES) & 0Ad Row Beaach Istu W (EPRO)

Slide 23 Notes:

! Sample Problem Exercise for Task 2, Steps 2 and 3

* Distribute blank handout for Task 2, Step 2

* Distribute completed handout for Task 2, Step 2 Question

and Answer Session

* Discuss Step 3

* Question and Answer Session

FiB PRA WotBSho, Rethesda MD, 20C8 R WS 23 A C W 0 US Nn O WN a R -a

Task 2: Compoýnt Selecton (RES) &EICUC Pofl W RnEI h )Sh S*5 (EPRI)

3-18
! Sample through Problem 6

* Distribute blank hand(

* Distribute completed

* Question and Answer

Fir PRA Workshop. Bethesda MD, 2008

Tas 2 CoXs pSnetSeectisn

Slide 24 Notes:

Exercise for Task 2, Steps 4

ut for Task 2, Steps 4 through 6

handout for Task 2, Steps 4 through 6

Session

Slide A Coeborateo8ou. NRC Notes :ONr.regt~h

Slide 25 Notes:

! Sample Problem Exercise for Task 2, Step 7

* Distribute blank handout for Task 2, Step 7

* Distribute completed handout for Task 2, Step 7

* Question and Answer Session

Fire PRA Wo•rOhop, Bethesda MD, 2008 o 2 CotIaoreto ofS NR00fic othNuceer eootetsy

Task 2. Component Setsction -eMhrhl (RES) & 0-Octr, Poeerheseh ttkttoo (EPRI)

3-19
Session lb: Example Exercises

3-20
Step 1: Identify Internal Events PRA Sequences to be Included (and

those to be excluded) in the Fire PRA Model

INITIATING EVENTS IN THE PRA MODEL

Initiator Average Description Accident Address Comments

Frequency Sequence in Fire

(per yr) or Event PRA

Tree Model?

Model (Y or N)

%T1 7.23E-01 Reactor Trip Transient

Event

Tree

%T2 9.33E-02 Loss of Transient

Condenser Event

Vacuum Tree

%T3 4.13E-01 Turbine trip Transient

Event

Tree

%T4 3.73E-02 Loss of Main Transient

Feedwater Event

Tree

%T5P 4.25E-02 Loss of Offsite Transient

Power (Plant- Event

Centered) Tree

%T5C 1.02E-02 Loss of Off-Site Transient

Power (Grid- Event

Related) Tree

%T5D 6.26E-03 Loss of Off-Site Transient

Power (Weather- Event

Induced) Tree

%T6 7.35E-03 Steamline/Feed Main

line Break Steamline

Upstream of Break

Main Steam Event

Isolation valves Tree

or Downstream

of Feedwater

Isolation Valves

(Includes StuckOpen Secondary

relief valves)

%T7 5.44E-03 Steamline Break Main

Downstream of Steamline

3-21
INITIATING EVENTS IN THE PRA MODEL

Initiator Average Description Accident Address Comments

Frequency Sequence in Fire

(per yr) or Event PRA

Tree Model?

Model (Y or N)

Main Steam Break

isolation valves Event

(Includes StuckOpen Secondary

relief valves)

%T8 2.94E-04 Loss of 4160 V Transient.

Bus 1 Event

Tree

%T9 2.94E-04 Loss of 4160 V Transient

Bus A Event

Tree

%T1 0 2.94E-04 Loss of 4160 V Transient

Bus B Event

Tree

%T11 2.94E-04 Loss of 4160 V Transient

Bus 2 Event

Tree

%T12 3.00E-03 Loss of 125 VDC Transient

Bus A Event.

Tree

%T13 3.OOE-03 Loss of 125 VDC Transient

Bus B Event

Tree

%T1 5 Fault Tree Loss of CCW Transient

Model System Event

%T15-1NIT Tree

%T16 Fault Tree Loss of Service Transient

Model Water System Event

%T16-1NIT Tree

%T1 7 Fault Tree Loss of Transient

Model Instrument Air Event

%T17- Tree

INIT

%T21 3.41 E-02 Closure of MSIV Transient

(1. SG Loop) Event

Tree

%T22 1.24E-02 Closure of both Transient

MSIVs Event

3-22
INITIATING EVENTS IN THE PRA MODEL

Initiator Average Description Accident Address

Frequency Sequence in Fire C

(per yr) or Event PRA

Tree Model?

Model (Y or N)

Tree

%T23 1.78E-01 Partial Load Transient

Rejection Event

Tree

%T24 5.79E-02 Spurious Steam Transient

Gen. Isolation Event

Signal Tree

%T25 7.23E-02 Reactor Trip With Transient

PORV Event

Opening/Demand Tree

%T26 Fault Tree Loss of Power Transient

Model from 120 VAC Event

%T26- Buses A & B Tree

INIT

%S 6.8E-03 Small LOCA Small

(pipe breaks and LOCA

RCP seal LOCA) Event

Tree

%M 9.60E-06 Medium LOCA Medium

(pipe breaks) LOCA

Event

Tree

%A 7.77E-05, Large LOCA Large

(pipe breaks) LOCA

Event

Tree

%R 7.93E-03 Steam Generator SGTR

Tube Rupture Event

Tree

%12 2.OOOE-07 Interfacing ISLRHR

Systems LOCA Sequence

at RCS/RHR (single

Interface (2 event

MOVs in series) model)

%13 Fault Tree Interfacing ISLCCW

Model Systems LOCA Sequence

13QINIT at RCS/CCW

interface

(Reactor Coolant

3-23
INITIATING EVENTS IN THE PRA MODEL

Initiator Average Description Accident Address Comments

Frequency Sequence in Fire

(per yr) or Event PRA

Tree Model?

Model (Y or N)

Pump Cooler

rupture)

%VR 2.70E-07 Reactor Vessel Single

Rupture Event in

Master

Fault Tree

3-24
ACCIDENT SEQUENCE OR EVENT TREE MODELS IN THE PRA

Accident Description Additional Details Address in Comments

Sequence or Fire PRA

Event Tree Model?

Model (Y or N)

TRA Transient Includes transientinduced LOCAs such

as stuck-open PORV

and RCP seal LOCA

SLOCA Small LOCA Pipe breaks & RCP

seal LOCA

MLOCA Medium LOCA Pipe breaks

LLOCA Large LOCA Pipe breaks

ATWS Anticipated Reactor Protection

Transients System fails safe on

Without Scram loss of power. Trip

circuits are highly

redundant and

confirmed to be

physically separated.

SGTR Steam Generator

Tube Rupture

MSLB Main Steamline Includes spurious

Break opening of secondary

relief valves.

ISLCCW Interfacing Rupture of Reactor

Systems LOCA at Coolant Pump Cooler

RCS/CCW

interface

ISLRHR Interfacing Fire-induced opening

Systems LOCA at of RHR suction valves

RCS/RHR

Interface

New?

New?

3-25
Step 2: Review of the Internal Events PRA Against the Fire Safe

Shutdown Analysis

TABLE 1: SYSTEMS IN PRA MODEL

System Description Additional Details Address Comments

in Fire

PRA

Model?

(Y or N)

RCS Reactor PORV for pressure relief and feed

Coolant & bleed. Stuck-open PORV

System causes small LOCA.

CVCS Chemical and Normal charging and letdown

Volume Control functions are not modeled.

System However, components required to

isolate charging and letdown are

modeled for HPI mode.

HPI High Pressure The charging pumps in the CVCS

Injection also function as safety injection

System pumps.

RHR Residual Heat Shutdown cooling is not modeled

Removal

System

AFW Auxiliary Only Trains A and B are modeled

Feedwater

System

MFW Main Feed Would take considerable effort to

Water get cables involved and their

locations

MS Main Steam Stuck-open secondary relief

System Valves could cause equivalent of

mainsteam line break.

3-26
TABLE 1: SYSTEMS IN PRA MODEL

System Description Additional Details Address Comments

in Fire

PRA

Model?

(Y or N)

CS Containment Required for recirculation during

Spray LOCA

CF Containment Required for recirculation during

Fan Coolers LOCA

Cl Containment Modeled in LERF

Isolation

ESFAS Emergency

Safeguards

Actuation

System

CCW Component

Cooling Water

System

SW Service Water

System

AC AC Power (all To extent power is needed to

voltage levels) support equipment in the PRA.

DG Emergency

Diesel

Generators

DC DC Power To extent power is needed to

support equipment in the PRA.

IA Instrument Air Required for PORV and other

valves. Backup nitrogen is

provided for PORV and is what is

credited.

3-27
TABLE 1: SYSTEMS IN PRA MODEL

System Description Additional Details Address Comments

in Fire

PRA

Model?

(Yor N)

HVAC- HVAC in HPI HVAC is required during 24-hr

HPI Pump Room PRA mission

3-28
TABLE 2: SYSTEMS IN APPENDIX R

System Description Additional Details Address Comments

in Fire

PRA

Model?

(Y or N)

RCS Reactor Coolant PORV to prevent

System spurious opening and

consequential small

LOCA.

CVCS Chemical Volume Normal charging and

and Control letdown functions are

System credited.

RHR Residual heat Shutdown cooling is

Removal System credited

AFW Auxiliary Trains A&C are credited

Feedwater

System

MS Main Steam Secondary relief valves

System and MSIVs are included

to prevent spurious

opening causing

uncontrolled secondary

depressurization.

CCW Component

Cooling Water

System

SW Service Water

System

AC AC Power (all But certain buses not

voltage levels) credited (especially nonsafety) if loads not

3-29
TABLE 2: SYSTEMS IN APPENDIX R

System Description Additional Details Address Comments

in Fire

PRA

Model?

(Y or N)

otherwise required for

safe shutdown

DG Emergency

Diesel

Generators

DC DC Power But certain buses not

credited (especially nonsafety) if loads not

otherwise required for

safe shutdown

RCS RCS pressure, Required for safe

Instruments temperature, shutdown monitoring.

nuclear

instrumentation,

etc

IA Instrument Air Required for PORV and

other valves. Backup

nitrogen is provided for

PORV.

Secondary Steam Generator Required for safe

Instruments level, Streamline shutdown monitoring.

pressure, etc.

HVAC-HPI HVAC in HPI HVAC is required during

Pump Room 72-hr Appendix R

mission

HVAC-AFW HVAC in AFW HVAC is required during

Pump Room 72-hr Appendix R

3-30
TABLE 2: SYSTEMS IN APPENDIX R

System Description Additional Details Address Comments

in Fire

PRA

Model?

(Y or N)

mission

3-31
TABLE 3: LIST OF BASIC EVENTS IN MODEL

Basic Event Description

%12 Interfacing Systems LOCA at RCS/RHR Interface (2 MOVs in series)

%13 Interfacing Systems LOCA at RCS/CCW interface (Reactor Coolant Pump Cooler rupture)

%T15 LOSS OF COMPONENT COOLING WATER (CCW)

%T23 PARTIAL LOAD REJECTION

%T25 REACTOR TRIP WITH PORV OPENING

%T3 TURBINE TRIP

%T4 LOSS OF MAIN FEEDWATER

%T1 REACTOR TRIP

AFWA-FTR AFWA fails to run

AFWA-FTS AFWA fails to start

AFWB-FTR AFWB fails to run

AFWB-FTS AFWB fails to start

AOV-1_FTC PORV AOV-1 fails to CLOSE

AOV-1_FTO PORV AOV-1 fails to open

AOV-3_FTC AOV-3 FAILS TO CLOSE

EPS-120VBUSAF 120V BUS A FAULT

EPS-120VBUSAINVF FAILURE OF 120V BUS A INVERTER

EPS-125VDCBUSAF FAULT ON 125V DC BUS A

EPS-125VDCBUSBF FAULT ON 125V DC BUS B

EPS-125VDCPNLAF FAULT ON 125V DC PANEL A

EPS-125VDCPNLBF FAULT ON 125V DC PANEL B

EPS-480VLCAF 480V LOAD CENTER A FAULT

EPS-480VLCAXTF 480V LOAD CENTER A TRANSFORMER FAILS

EPS-480VLCBF 480V LOAD CENTER B FAULT

3-32
TABLE 3: LIST OF BASIC EVENTS IN MODEL

Basic Event Description

EPS-480VLCBXTF 480V LOAD CENTER B TRANSFORMER FAILS

EPS-480VMCCA1 F 480V MCC Al FAULT

EPS-480VMCCBlF 480V MCC B1 FAULT

EPS-4VBUSAF 4KV BUS A FAULT

EPS-4VBUSBF 4KV BUS B FAULT

EPS-BATA FAILURE OF BATTERY A

EPS-BATB FAILURE OF BATTERY B

EPS-BCAF FAILURE OF BATTERY CHARGER A

EPS-BCBF FAILURE OF BATTERY CHARGER B

EPS-DGAF FAILURE OF DIESEL GENERATOR A

EPS-DGBF FAILURE OF DIESEL GENERATOR B

HPIAFTR HPIA fails to run

HPIAFTS HPIA fails to start

HPIBFTR HPIB fails to run

HPIBFTS HPIB fails to start

MFWFAIL, MAIN FEEDWATER SYSTEM FAILURE AFTER REACTOR TRIP

MOV-10_FTO MOV-10 fails to open

MOV-1 1_FTO MOV-1 1 fails to open

MOV-14_FTO MOV-14 FAILS TO OPEN

MOV-15_FTO MOV-15 FAILS TO OPEN

MOV-1_FTO MOV-1 FAILS TO OPEN

MOV-2_FTC MOV-2 fails to close

MOV-3_FTO MOV-3 fails to open

MOV-4_FTO MOV-4 fails to open

3-33
TABLE 3: LIST OF BASIC EVENTS IN MODEL

Basic Event Description

MOV-5_FTC MOV-5 fails to close

MOV-5_FTO MOV-5 fails to open

MOV-6_FTC MOV-6 fails to close

MOV-6_FTO MOV-6 fails to open

MOV-9_FTO MOV-9 FAILS TO OPEN

OPER-1 Operator fails to switch over to recirculation

OPER-4 Operator fails to establish feed an bleed cooling

OPER-7 OPERATOR FAILS TO TRIP REACTOR COOLANT PUMP

RCPSEAL RCP SEAL LOCA GIVEN LOSS OF CCW AND SUCCESSFUL RCP TRIP

SUTF FAILURE OF START-UP TRANSFORMER (SUT)

UATF FAILURE OF UNIT AUXILIARY TRANSFORMER (UAT)

3-34
Continuation of Step2 and Including Steps 4 thru 6:

TABLE 4: DISPOSTION OF COMPONENTS IN PRA AND APPENDIX R (USE TABLES 1-3 AND

FIGURES 1-3)

In PRA In Add to Fire PRA Comments

Equipment ID Equipment Power Model? Appendix Equipment

Description Supply (Y or N) R? (Y or List? (Y or N)

N)

High pressure Y

HPI-A safety injection

Bus A

pump A

High pressure Y

HPI-B safety injection

Bus B

pump B

4.16kV Y

RHR-B RHR pump Bus B

Bus B

Instrument air 480 V Y

COMP-1

compressor LC 1

Motor driven 4.16kV Y AFW-A

AFW pump A Bus A

Steam driven N

AFW-B N/A

AFW Pump B

4.16 kV Y

AFW-C AFW Pump C

Bus 2

AOV-1 Pressure 

Y (only to

operated relief 120VAC ensure

(SOY-1) valve Bus A remains

closed)

AOV-2 Letdown 125 Y(for

VDC normal

(SOV-2) isolation valve

Bus B letdown)

A0V3 Charging pump 125 Y (for

injection valve VDC normal

3-35
TABLE 4: DISPOSTION OF COMPONENTS IN PRA AND APPENDIX R (USE TABLES 1-3 AND

FIGURES 1-3)

In PRA In Add to.Fire PRA Comments

Equipment ID Equipment Power Model? Appendix Equipment

Description Supply (Y or N) R? (Y or List?,(Y or N)

N)

(SOV-3) Bus B charging)

480V N

MOV-1 HPI valve

MCC Al

Y (for

VCT isolation 480V normal

MOV-2

valve MCC B1 suction to

charging)

Cont. sump 480V N

MOV-3

recirc valve MCC Al

Cont. sump 480V N

MOV-4

recirc valve MCC B1

RWST isolation 480V N

MOV-5

valve MCC Al

RWST isolation 480V N

MOV-6

valve MCC B1

Y (for

RHR inboard 480V

MOV-7 shutdown

suction valve MCC Al

cooling)

Y (for

RHR outboard 480V

MOV-8 shutdown

suction Valve MCC B1

cooling)

.480V N

MOV-9 HPI valve

MCC B1

AFW discharge 480V Y

MOV-10

valve MCC Al

MOV-1l AFW discharge 125 N

3-36
TABLE 4: DISPOSTION OF COMPONENTS IN PRA AND APPENDIX R (USE TABLES 1-3 AND

FIGURES 1-3)

In PRA In Add to Fire PRA Comments

Equipment ID Equipment Power Model? Appendix Equipment

Description Supply (Y or N) R? (Y or List? (Y or N)

N)

valve VDC

Bus B

PORV block 480V Y

MOV-13

valve MCC Al

AFW turbine 125 N

MOV-14 steam line VDC

isolation valve Bus B

AFW steam 125 N

MOV-1 5 inlet throttle VDC

valve Bus B

AFW test line 480V N

MOV-1 6

isolation valve MCC Al

AFW test line 480V N

MOV-1 7

isolation valve MCC B1

AFW C Pump 480 V Y MOV-18

Discharge MCC-2

AFW test line 480 V N

MOV-19

isolation valve MCC-2

Y (for

admin

CST isolation

V-12 N/A purposes valve.

to ensure

open)

120VAC Y

LI-1 RWST level

Bus A

LI-2 RWST level 120VAC Y

3-37
TABLE 4: DISPOSTION OF COMPONENTS IN PRA AND APPENDIX R (USE TABLES 1-3 AND

FIGURES 1-3)

In PRA In Add to Fire PRA Comments

Equipment ID Equipment Power Model? Appendix Equipment

Description Supply (Y or N) R? (Y or List? (Y or N)

N)

Bus B

Cont. sump 120VAC Y LI-3

level Bus A

Cont. sump 120VAC Y LI-4

level Bus B

Letdown heat N

120VAC

TI-1 exchanger Bus A

outlet temp

120VAC N

PT-1 RCS pressure Bus B

AFW motor 120VAC N

A-1

high temp Bus A

Train A 4160 V SUT-1 Y

SWGR-A

Bus EDG-A

Train B 4160 V SUT-1 Y

SWGR-B

Bus EDG-B

SWGR-i Non-Safety UAT-1

4160V Bus SUT-1

Non-Safety UAT-1 Y

SWGR-2

4160 V Bus SUT-1

SUT-1 Startup OSP

Transformer

Train A Y

EDG-A PNL-A

Emergency

3-38
TABLE 4: DISPOSTION OF COMPONENTS IN PRA AND APPENDIX R (USE TABLES 1-3 AND

FIGURES 1-3)

In PRA In Add to Fire PRA Comments

Equipment ID Equipment Power Model? Appendix Equipment

Description Supply (Y or N) R? (Y or List? (Y or N)

N)

Diesel

Generator

Train B Y

Emergency EDG-B PNL-B

Diesel

Generator

Non-Safety 480 N LC-1 SST-1

V Load Center

Non-Safety 480 Y LC-2 SST-2

V Load Center

Train A 480 V Y

LC-A SST-A

Load Center

Train B 480 V N

LC-B SST-B Load Center

Non-Safety N SWGRSST-1 Station Service 1

Transformer

Non-Safety SWGR- Y

SST-2 Station Service

2

Transformer

Train A Station Y

SWGRSST-A Service

A

Transformer

Train B Station N

SWGRSST-B Service

B

Transformer

3-39
TABLE 4: DISPOSTION OF COMPONENTS IN PRA AND APPENDIX R (USE TABLES 1-3 AND

FIGURES 1-3)

In PRA In Add to Fire PRA Comments

Equipment Power Model? Appendix Equipment Equipment ID

Description Supply (Y or N) R? (Y or List? (Y or N)

N)

Non-Safety 480 N

MCC-1 V Motor Control LC-1

Center

Non-Safety 480 Y

MCC-2 V Motor Control LC-2

Center

Train A 480 V Y

MCC-A1 Motor Control LC-A

Center

Train B 480 V N

MCC-B1 Motor Control LC-B

Center

Non-Safety MCC-1 N

BC-1 Swing Battery

Charger MCC-2

BCA Train A Battery MC-lN

Charger

Train B Battery N

Charger

Non-Safety N BAT-i N/A

Battery

BAT-A Train A Battery N/A Y

BAT-B Train B Battery N/A Y

CNon-Safety 125 BC-1 N

DC BUS-It BC-I VDC Bus BAT-I

3-40
TABLE 4: DISPOSTION OF COMPONENTS IN PRA AND APPENDIX R (USE TABLES 1-3 AND

FIGURES 1-3)

In PRA In Add to Fire PRA Comments

Equipment Power Model? Appendix Equipment Description Supply (Y or N) R? (Y or List? (Y or N)

N)

Train A 125 BC-A Y

DC BUS-A

VDC Bus BAT-A

Train B 125 BC-B Y

DC BUS-B

VDC Bus BAT-B

DC Y

INV-A Train A Inverter

BUS-A

DC Y

INV-B Train B Inverter

BUS-B

Train A 120 Y

VITAL-A INV-A

VAC Vital Bus

Train B 120 Y

VITAL-B INV-B

VAC Vital Bus

Train A 125 DC Y

PNL-A

VDC Panel BUS-A

Train B 125 DC Y

PNL-B

VDC Panel BUS-B

3-41
Step 3: Identify Fire-Induced Initiating Events Based on Equipment

Affected

Will need to examine each fire compartment / analysis unit and determine based on the

equipment and cables located there, which of the initiators (from Step 1 of Task 2) can be

caused by a fire in that compartment / analysis unit. If any new initiators are identified, include in

the Fire PRA. The Fire PRA will then include fires mapped to initiating events in the model.

Each compartment / analysis unit should have a disposition with regard to the initiating event(s)

that occur as a result of a fire in each location (even if "none"). Hold discussion with instructor.

3-42
Step 4: Identify Equipment with Potential Spurious Actuations that May Challenge the Mitigating

Capability to be Credited

Considered Description PRA System or Function Comments/Disposition

spurious equipment Possibly Affected

operations

MOV-16 AFW test line isolation valve

MOV-17 AFW test line isolation valve

MOV-19 AFW test line isolation valve

MOV-3 Cont. sump recirc. valve

MOV-4 Cont. sump recirc. valve

AOV-2 Letdown isolation valve

RHR RHR pump

3-43
Step 5: Identify Additional Mitigating, Instrumentation, and Diagnostic Equipment Important to

Human Response

Instrumentation Description Potentially Affected Comments/Disposition

• required to Human Failure Event

perform human in PRA model

actions

LI-1 RWST level

LI-2 RWST level

LI-3 Cont. sump level

LI-4 Cont. sump level

Letdown heat exchanger outlet temp

PT-1 RCS pressure

A-1 AFW motor high temp

Step 6: Include Potentially High Consequence Related Equipment

Equipment or combinations of Description of high consequence Comments/Disposition

equipment that can cause high event

consequence event

3-44
Step 7: Assemble Fire PRA Equipment List

Table 1: Fire PRA Equipment List Information (For Instructors)

Equipment PRA Event PRA Event Normal Desired Failed Failed Air Equipment ID Equipment Description Type Location Power Supply Identifier Description Position/ Position/ Electrical Position Status Status Position

HPIA FTS HPIA fails to

HPIWA1 High pressure safety Pump Aux Bldg. SWGR-A

injection pump A El. 0 Ft

HPIA_FTR HPIA fails to run

HPIB FTS HPIB fails to

HPI-B High pressure safety Aux Bldg. SWGR-t

injection pump B El. 0 Ft

HPIB_FTR HPIB fails to run

Motor driven AFW Turbine Bldg. AFWA fails to AFW-A pump A Pump El. SWGR-A AFWA-FTS start

AFW-A Motor dven AFW Pump Turbine Bldg. Turbine Bldg. AFWA-FTR AFWA fails to

Mtrdiepump A El. 0 Ft El. 0 Ft run

Steam driven AFW Turbine Bldg. N/A AFWB-FTS AFWB fails to AFW-B pump B Pump El. 0 Ft N A start

AFW-B Steam driven AFW

pump B Pump Turbine Bldg.

El. 0 Ft N/A AFWB-FTR AFWB fails torun

Turbine Bldg. WGAFW-C Motor driven AFW Pump el SWGR-2 pumpC El. 0 Ft

Motor driven AFW Turbine Bldg. SWGR-2 pump d PumP El. 0 Ft

RCP Reactor coolant pump Pump Containment SWGR-1

Instrument air C r r El.0F COMP-1 compressoressor LC

I t

I t

i i i

3-45
PORV AOV-1 AOV-I_FTO fails to open

RWST isolation valve

RWST isolation valve

RHR inboard suction

valve

RHR outboard suction

valve

3-46
3-47
Equipment ID Equipment Description Equipment

Type Location Power Supply PRA Event

Identifier

PRA Event Normal

Position/ Descdption Status

Desired

Position/

Status

Failed

Electrical

Position

Failed Air

Position

A-1 AFW motor high Annunciator SWG Access VITAL-A temperature Room

_ i_ i i

SUT-1

DC BUS-A

SWGR-A1' Train A 4160 V

switchgear Switchgear Switchgear

Room A

EDG-A

DC BUS-A

PNL-A

EPS-4VBUSAF2

4KV BUS A

FAULT

SUT-1 PNL-B KBUA

DC BUS-B EP-VUB- FAULT

SWGR-B Train B 4160 V Switchgear Switchgear 1

switchgear Room B EDG-B PS-BUSBF 4K BUS A

DC BUS-B 2P-VB~ FAULT

Non-safety 4160 V Turbine Bldg. UAT-1 SWR1 switchgear Swther El. Oft SUT-1

SWGR-2 Non-safety 4160 V Swtchgear Turbine Bldg. UAT-1 SWGR-2 El. Oft SUT-1

FAILURE OF

SUT-1 Startup transformer Transformer Yard OSP SUTF TAR MU

R (SLIT)

Trai DeselFAILURE A eergecy OF

EDG-A Train A emergency Diesel DG Bldg. DC BUS-A EPS-DGAF DIESEL diesel generator Generator GENERATOR A

EDG-B Train B emergency

diesel generator

Diesel

Generator

FAILURE OF

DG Bldg. DC BUS-B EPS-DGBF I•t F• I:1

LC-1 Non-safety 480 V load Load Center Turbine Bldg. SST-1 center El. 0 ft

LC-2 Non-safety 480 V load Load Center Turbine Bldg. SST-2 center El. 0 ft

3-48
E tPRA Event PRA Event Normal Desired Failed Failed Air

Equipment ID Equipment Description Type Location Power Supply Identifier Description Position/ Position/ Electrical Position

Type_____________ IdetiieDecrpton Status Status Position 48VLOAD

LC-A Train A 480 V load Load Cetr Switchgear SST-A EPS-480VLCAF CENTER A

center Room A PNL-A CENTER FAULT A

LC-B Train B 480 V load Load Center Switchgear SST-B 480V LOAD

center Room B PNL-B EPS-480VLCBF CENTER B FAUL

Non-safety station Transformer bine Bldg. SST-1 TurbineoBldg SWGR-1 service transformer El. 0 F

SST-2 station Transformer Turbine Bldg. SWGR-2 service transformer El. 0 F

480V LOA

SST-A Train A station service T or Switchger SWGR-A E CENTER A

transformer Room A 480VLCAXTF T SFORME

R FAILS

480V LOAD

SST-B Train B station service Transformer Switchgear SWGR-B EPS- CENTERMB transformer Room B 480VLCBXTF TRANSFORME

R FAILS

Non-safety 480 V Motor Control Turbine Bldg. P6, A

motor control center Center El. 0 Ft

MCC-2 Non-safety 480 V Motor Control Turbine Bldg. LC-2 motor control center Center El. 0 Ft

Train A 480 V motor Motor Control SWG Access LC-A EPS- 480V MCC Al MCC-AI control center Center Room 480VMCCAIF FAULT

MCC-B1 Train B 480 V motor Motor Control SWG Access LC-B EPS- 480V MCC B1

control center Center Room 480VMCCB1 F FAULT

ATS-1 Automatic transfer SWG Access MCC-1

switch ATSRoom MCC-2

Non-safety swing Battery Turbine Bldg. ATS-1 BC-1 battery charger Charger El. 0 Ft

BC-A Train A battery Battery Switchgear MFC-A1 EPS-BCAF BATTERY

charger Charger Room A CHARGERA

I I I ICHARGER A

3-49
Equipment PRA Event PRA Event Normal Desired Failed Failed Air Equipment ID Equipment Description Typ Location Power Supply Identifier Description Position/ Position/ Electrical Position ype Status Status Position

BC-B Train B battery Battery Switchgear MCC-Bi EPS-BCB F BATTERY

charger Charger Room B CHATTERY

Turbine Bldg. BAT-1 Non-safety battery Battery El. 0 FtN/A El Ft- EP-0T RE

BAT-A Train A battery Battery BRoy N/A EPS-BATA FAILURE OF _____ ________ _____ Room ABATRA

B battery Battery Battery N/A EP-BATB FAILURE OF

BAT-B Train BbaRoom B N E"- FBATTERY B

DC BUS-i Non-safety 125 VDC DC Bus Turbine Bldg. BC-1

bus El. 0 Ft BAT-1

DCBUS-A TrainA125VDCbus DCBus Switchgear BC-A EPS- FAULT ON

DC_ US- Trin12 _DCbus DBus Room A BAT-A 125VDCBUSAF 125V DC BUS A

DC BUS-B Train B 125 VDC bus DC Bus Swltchgear BC-B EPS- FAULT ON

Room B BAT-B 125VDCBUSBF 125V DC BUS B

PNL-A Train A 125 VDC Panel board t DC BUS-A EPS- FAULT ON PNL- Pael bard 'chgar 25V125V DC

panel Room A I25VDCPNLAF PANEL A

PNL-B Train B 125 VDC

panel Panel board Switchgear

Room B DC BUS-B EPS125VDCPNLBF

FAULT ON

125V DC

4 4 4

INV-A Train A inverter Inverter Switchgear

Room A DC BUS-A

INV-B Train B inverter Inverter Switchgear Room B DB BUS-B

VITAL-A Train A 120 VAC vital 120VAC Bus SWG Access INWA bus Room

VITAL_-B Train B 120 VAC vital 120VAC Bus SWG Access INV-B bus Room

1 -t -t

Notes:

3-50
Legend

Appendix R

Components/failures added

to PRA for Fire PRA Model

1 HPI pumps A & B are in Appendix R for normal charging function; not for HPI. 2 Closed to prevent LOCA /Open for feed and bleed. Appendix R has valve only to ensure remains closed.

3 AOV-2 in Appendix R only for normal letdown function.

4 AOV-3 in Appendix R only for normal charging function.

5 MOV-2 in Appendix R only for normal suction to charging.

6 MOV-3 and MOV-4 need to be closed for when using RWST water supply for HPI / Open for recirculation mode of injection. MOVs 5 & 6 are correspondingly open for RWST use and

closed for recirc.

7 Valve electrically blocked closed. Control power fuses are supposed to be removed. MOV-7 & 8 in Appendix R for both normal isolation and shutdown cooling functions. 8 If RWST level indication fails high the operator will fail to establish recirculation. If the RWST level fails low and containment sump level fails high, the operator will establish suction

to dry sump and fail the HPI the pumps due to insufficient suction. Need to resolve failure mode.

9 If letdown temperature indication fails low, the operator will fail to isolate the letdown line on loss of CCW. This will cause HPI pumps to cavitate due to high temperature in the

suction line.

10 If RCS pressure instrument fails high, the low pressure signal to initiate emergency safeguards actuation during a lire-induced LOCA will not be initiated

1 For SWGR-A or SWGR-B, only one basic event is used in the model. However, the circuit analysis impact to the switchgear is a function of the power supply

3-51
Session 2: Fire Induced Risk Model Development

3-52
Slide 1 Notes:

fP_- _ ~~IIEL RESEARCH ECTRIC 00WER InStiTUE

EPRI/NRC-RES FIRE PRA

METHODOLOGY

Task 5 - Fire-Induced Risk Model

Development

Joint RES/EPRI Fire PRA Course

September and November 2008

Bethesda, MD

I A Co&.boration of U.$ NRC 0ff,• o& Nud.ar R-gl-tory Research (RES) & EIctri Po.we R e...rh lnstitute (EPRI)

Slide 2 Notes:

!

Fire PRA Risk Model

Scope

* Task 5: Fire-Induced Risk Model Development

- Constructing the PRA model

Fire PRO Wor*shop, Bftosa MD. 2008 Task 5. R4 0nd RisO ModU . t.opono S~d. 2 A Coa &bora U. S. WC Om O d Nuk gOtory .ori, IRES) &Ele , . 0 IEURO

3-53
Slide 3 Notes:

! Fire PRA Risk Model General Comment/Observation

Task 5 does not represent any changes from past

practice, but what is modeled is largely based on Task 2

with HRA input from Task 12

Bottom line -just "tweaking" your Internal Events PRA is

probably NOT sufficient

F-, PRO W oW. 8.1 ý,0dD, 2008 s . A C baUOe - 0ouS. N120 002me . NýRe1,ft

MSo 5.Rn4OduOed sk odel12M D .,,8 , _n , Re (RES) A 80 00 R0 - (00 PR)O

Slide 4 Notes:

I Task 5: Fire Risk Model Development General Objectives

Purpose: Configure the Internal Events PRA to provide fire

risk metrics of interest (primarily CDF and LERF).

* Based on standard state-of-the-art PRA practices

* Intended to be applicable for any PRA methodology or

software

* Allows user to quantify CDF and LERF, or conditional

metrics CCDP and CLERP

* Conceptually, nothing "new" here - need to "build the PRA

model" reflecting fire induced initiators, equipment and

failure modes, and human actions of interest

FE PRO Worshop, SeP.00a MD, 2008 M A 000o08o0,80 S NRC 04008 RapeCmy Tas S. Firs00.8nue RhO04k Mode . fieopm~nt 0000.08(RE0) EOlt0..000000 R (o= C020I•A (0000

3-54
Slide 5 Notes:

! Task 5: Fire Risk Model Development Inputs/Outputs

Task inputs and outputs:

Inputs from other tasks: [Note: inclusion of spatial

information requires cable locations from Task 3]

- Sequence considerations, initiating event considerations, and

components from Task 2 (Fire PRA Component Selection),

- Unscreened fire compartments from Task 4 (Qualitative Screening),

- HRA events from Task 12 (Post-Fire HRA)

* Output to Task 7 (Quantitative Screening) which will further

modify the model development.

* Can always iterate back to refine aspects of the model

RM PRA Wo,.c8hp, B e.esdR AID, 2008 0 A C2 0 d 0S,.N 0NR Oc80 8o 8R2g8

ra.• 0-FOr.4nd.,8d PJR Model Fopme,, F4mc8 (00S) 8200,8 , 802 POO

Slide 6 Notes:

! Task 5: Fire Risk Model Development Steps in Procedure

Two major steps:

* Step 1: Develop CDF/CCDP model

* Step 2: Develop LERF/CLERP model

Fim PRA Wwo*8P Be0heda MD, 2008 0 Co&0b0gioo (R 0 00R Of-•e ou• tory0la 

Tsk 5 . Flmýnduc.2 Risk Mode) Deelopment Reýr0h (RES) & 0000r00 .h ,erR (P00

3-55
Slide 7 Notes:

Task Steps 5: Fire Risk Model Development in Procedure/Details

Step 1 (2): Develop CDF/CCDP (LERF/CLERP) models

Step 1.1 (2.1): Select fire-induced initiators and sequences and

incorporate into the model

Each fire-induced initiator is mapped to an internal events initiator that

mimics the effect on the plant of the fire initiator

Internal events sequences form bulk of sequences for Fire PRA, but a

search for new sequences should be made (see Task 2). Some new

sequences may require new logic to be added to the PRA model

FhT PRA W*kShop, BbesO MD, 200 " ] A C US-C S Woatlao O &t.ýo Nu008w 08 W oY

Tý 55 Sn Risk Mode DeO-P R-h (-ES) &. E P 0.0. - R-S - u-(EP-#

Slide 8 Notes:,

Task Steps 5: Fire Risk Model Development in Procedure/Details

Step 1.1 (2.1) - continued

Plants that use fire emergency procedures (FEPs) may

need special models to address unique fire-related actions

(e.g., pre-defined fire response actions and MCR

abandonment).

* Some human actions may induce new sequences not

covered in Internal Events PRA and can "fail" components

- Example: SISBO, or partial SISBO

Fire PRA WaOdhop, Be000d- MD, 2008 side CdM-a8500t°S US. N NS-k t- ROguOSSY

Task 5 o FInducd RIsk Modl Denbopm. nt 5R-ft - (RES) A05800 Power Research u(0eEPRI)

3-56
Slide 9 Notes:

! Task 5: Fire Risk Model Development Steps in Procedure/Details

Loss ofraw watoe

as initiator

Example of new logic with a fireinduced loss of raw water initiating

event

Fire PRA Wor*OW. 0.50.5al MO, 200W 8 , CooM -ine OO US NRC OA oN2Rioy

Th.k 5. F1r.Aod.e.d R10 Uodel D 0toR & 0e00r05 Power R 22lt (EP1lI

Slide 10 Notes:

! Task 5: Fire Risk Model Development Steps in Procedure/Details

Step 1.2 (2.2): Incorporate fire-induced equipment failures

* Fire PRA database documents list of potentially failed

equipment for each fire compartment

* Basic events for fire-induced spurious operations are

defined and added to the PRA model

* Inclusion of spatial information requires equipment and

cable locations

- May be an integral part of model logic, or handled with manipulation

of a cable location database, etc.

FiR. PRA Wor000op, Bethesda MD, 2008 [ A conbort800 U.S. N O 000e N -C 06 R 1u t

Task 5 .fr.Anducd Risk Model Oeweop-nf R -snh (RES) & E-0t0c P-e R-e-o0 osblA, (EFR0

3-57
Slide 11

Task 5: Fire Risk Model Development

Steps in Procedure/Details

Loss of high

Original logic pressure injection

..Suppose fire in

fails to sta fails ta .s to open could fail pump A

because pump A is in Li

and cable for pump A is

in L2

Loss o fhigh

Possible temporary pressure injection

change to model to run

CCDPs for LI and L2

Set to

TRUE

Fire PRA Workshop, Bethesda MD, 2008 . .A Collaboration of U.S. NRC Office of Nuclear Regulatory

Task 5 - Fire-Induced Risk Model Development I Slide Research (RES) & Electric Power Research Institute (EPRI)

Notes:

3-58
Slide 12

Notes:

3-59
Slide 13 Notes:

! Task 5: Fire Risk Model Development Steps in Procedure/Details

Step 1.3 (2.3): Incorporate fire-induced human failures

* New fire-specific HFEs may have to be added to the model

to address actions specified in FEPs [Note: all HFEs will be

set at screening values at first, using Task 12 guidance]

* Successful operator actions may temporarily disable ("fail")

components

FR. PPA ý WNdthp. B5,hesa MdD M, 2X0 2M A Cotb ,, 57U8. NRC 054.f l 

AMA 5-fl,.InducdRIskMod• 5 . FlR.RC •walopmýn R (DES) R k ý . (ý (&RO E

3-60
Slide 14

i Task 5: Fire Risk Model Development

Steps in Procedure/Details

Suppose a proceduralized manual action

carried out for fires in compartments AA & BB

defeats Pump A operation by de-energizing the

pump (opening its breaker drawer)...

Fire PRA Workshop, Bethesda MD, 2008

Task 5 - FIre-Induced Risk Model Development

Slide 14I A Collaboration of U.S. NRC Office of Nuclear Regulatory .............................. Research (RES) & Electric Power Research Institute (EPRI)

Notes:

3-61
Slide 15 Notes:

! Sample Problem Exercise for Task 5

* Distribute blank handout for Task 5, Steps 1 and 2

* Distribute completed handout for Task 5, Steps 1 and 2

* Question and Answer Session

Fire PRA Workshop, Be&hesda MD, 2008 CAto.. of UA U NRC 0006 NuC0eA R y

T 5-fýlýnducd Rs. RikModIDeopmWnt R-00 (RES) & ElecOc P RR 6 0 00-t- (EPRO

3-62
Session 3: Qualitative/Quantitative Screening

3-63
Slide 1 Notes:

•..•,o REEARCH INSnuTE

EPRI/NRC-RES FIRE PRA

METHODOLOGY

Task 4 - Qualitative Screening

Task 7 - Quantitative Screening

Joint RES/EPRI Fire PRA Workshop

September and November 2008

Bethesda, MD

A Collborabtion of U.S. NRC Office of Niuclear Regltory Rsesh (RES) & Elecoic Power Reseanrh Institute (EPR)

Slide 2 Notes:

! Qualitative / Quantitative Screening Scope

* Task 4: Qualitative Screening

- First chance to identify very low risk compartments

* Task 7: Quantitative Screening

- Running the Fire PRA model to iteratively screen / maintain

modeled sequences at different levels of detail

Fim PRA WEok1oi, BeOeoaa MD, 2008 r A c o o . 00 O2c dNc ere

Trsk 4£?- Quaiite~ve/Qulnb'ette Scrniog ReSOrch 0RE) & E7ecri P- ReO COCOS (0/0R0

3-64
Slide 3 Notes:

Task 4: Qualitative Screening

Objectives and Scope

* The objective of Task 4 is to identify those fire

compartments that can be shown to have a negligible risk

contribution without quantitative analysis

- This is where you exclude the office building inside the protected

area

* Task 4 only considers fire compartments as individual

contributors

- Multi-compartment scenarios are covered in Task 11(b)

- Compartments that screen out qualitatively need to be reconsidered as potential Exposing Compartments in the multicompartment analysis (but not as the Exposed Compartment)

Re PRA WO, fhehel MD. 2008 [ ] A CdkboRtk doS NRC ORC. o RRdk

fTi 4 & 7 - OfltufiA84OuatitN S-re tnr R008h (REC) & CI0C Tower ReSesh CStOuf (EPA)

Slide 4 Notes:

Task 4: Qualitative Screening Required Input and Task Output

* To complete Task 4 you need the following input:

- List of fire compartments from Task 1

- List of Fire PRA equipment from Task 2 including location mapping

results

- List of Fire PRA cables from Task 3 including location mapping

results

* Task Output: A list of fire compartments that will be

screened out (no further analysis) based on qualitative

criteria

- Unscreened fire compartments are used in Task 6 and further

screened in Task 7

Fir P04 Wo,*rsOp, Bethewet MO, 2008 4 A Colbraio doS. NRC 0&US 0 o NU 0 9080',

Tak 4 & 7 - OuaktokN/Quaool3btlV Sc,.,ng I - R - (RES) & E POOR• p R fh -8t0u (EPRI)

3-65
Slide 5 Notes:

! Task 4: Qualitative Screening A Note....

Qualitative Screening is OPTIONAL!

- You may choose to retain any number of potentially low-risk fire

compartments (from one to all) without formally conducting the

Qualitative Screening Assessment for the compartment

- However, to eliminate a compartment, you must exercise the

screening process for the compartment

- Example 1: Many areas will never pass qualitative screening, so

simply keep them

- Example 2: If you are dealing with an application with limited scope

(e.g. NFPA 805 Change Evaluation) a formalized Qualitative

Screening may be pointless

Rke PRA WwkshW, B.nd. MD. 2008 s A CotO,. 60 U.S. N 07 W7&cbON 0.gMda).

TSA~ 44& 7 - Qullt~itnQuan~tahn Snenlfl Renwv, (WSE) & , P i ft. -1.n.c7 0 (EM2I)

Slide 6 Notes:

g Task Screening 4: Qualitative Criteria Screening

A Fire Compartment may be screened out** if: - No Fire PRA equipment or cables are located in the compartment, and

- No fire that remains confined to the compartment could lead to:

* An automatic plant trip, or * A manual trip as specified by plant procedures, or * A near-term manual shutdown due to violation of plant Technical

Specifications*

*In the case of tech spec shutdown, consideration of the time

window is appropriate - No firm time window is specified in the procedure - rule of thumb:

consistent with the time window of the lire itself - Analyst must choose and justify the maximum time window

considered

(**Note: screened compartments are re-considered as fire source compartments in the multi-compartment analysis - Task 1 ic)

Firm PRA Workshop, Be1h0.00 MD, 2008 [ , A Cd)80,O0 dOS NRC 070 of Nu.20*, Reguatoty

Task 4 & 7 - Quaiita /Quaibtiv. SornnI - -*81h (-ES) I -)0,00 -*w0 -S h -,2)0u. (-70I)

3-66
Slide 7 Notes:

Task General 7: Objectives Quantitative Screening

Purpose: allow (i.e., optional) screening of fire compartments and scenarios based on contribution to fire risk. Screening is

primarily compartment-based (Tasks 7A/B). Scenario-based

screening (Tasks 7C/D) is a further refinement (optional).

* Screening criteria not the same as acceptance criteria for

regulatory applications (e.g., R.G. 1.174)

* Screening does not mean "throw away" - screened

compartments/scenarios will be quantified (recognized to be

conservative) and carried through to Task 14 as a measure

of the residual fire risk

Fri PRA Workshop, 88the.da MO. 20 M S CoIbbora A &S D NRC Oll•a oM w0u Re oy

T 4 & 7 - Ou.UWtefatbOn SO.h ftDm.Rs (RS) & •0 - Rh -P8w st-in (E')

Slide 8 Notes:

Task 7: Quantitative Screening Inputs/Outputs

Inputs from other tasks for compartment-based screening

(7A/B):

- Fire ignition frequencies from Task 6,

- Task 5 (Fire-Induced Risk Model),

- Task 12 (Post-Fire HRA Screening), and

- Task 8 (Scoping Fire Modeling) (7B only)

Fim PRA Wo"shop, ý8he8d. MO, 2000 A 00o05, ofS U5$ N850I00N R"g-My

Task Qu.15v•/Qulatvn Scnsnlng F-?- Sese (SF e 08,85 0ht8,f (PR5

3-67
=. .

Task 7: Quantita

Inputs/Outputs (cont

Inputs from other task

include inputs listed at

- Task 9 (Detailed Circui

- Task 11 (Detailed Fire

- Task 12 (Post-Fire HPR

- Task 10 (Circuit Failure

Fre PRA Worthop, Bethesda MD. 200M

Ts 4 A 7 - O•u.UvWQulmntttýv ScýInmng

Slide 9 Notes:

tive Screening

'd)

s for scenario-based screening (7CMD)

bove plus:

t Failure Analysis) and/or

Modeling) and/or

A Detailed), and

Mode Likelihood Analysis) (7D only)

S~.9 A Colabora00,o UOOS. NRC Offi. ol u., 00~hR•

Rumn (RES) ( Dirn Pee Ri Ons0 (0P80

Slide 10 Notes:

Task 7: Quantitative Screening

Inputs/Outputs (cont'd)

Outputs to other tasks:

- Unscreened fire compartments from Task 7A go to Task 8 (Scoping

Fire Modeling),

- Unscreened fire compartments from Task 7B go to Task 9 (Detailed

Circuit Failure Analysis) and/or Task 11 (Detailed Fire Modeling)

and/or Task 12 (detailed Post-Fire HRA),

- Unscreened fire scenarios from Task 7C/D go to Task 14 (Fire Risk

Quantification) for best-estimate risk calculation

Fim PRA WofhSop, Beobsi.e MD, 2008 Cd~h DI A=A U.S. NRC 0•OE ON N RO• RI)

T 0 4 & 7 -.- 0,0000Snlg

3-68
Slide 11

Task 7: Quantitative Screening

Overview of the Process

Make

Unscreened compartment Make

or scenario based on

calculated

CDF/CCDP/LERF/CLERP

more realistic via

ircuit analysis

Perform any one,

two, or all three

more realistic via based on where

fire modeling you will get more

realistic results

for the least

resources more realistic via

re detailed HRA

Screens?

If NO, iterate as

necessary

wŽ1 Make

mo

Fire PRA Workshop, Bethesda MD, 2008

Task 4 & 7 - Qualitative/Quantitative Screening

Slide 11 A Collaboration of U.S. NRC Office of Nuclear Regulatory

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

Notes:

3-69
Slide 12 Notes:

Task 7: Quantitative Screening Steps in Procedure

Three major steps in the procedure:

* Step 1: Quantify CDF/CCDP model

* Step 2: Quantify LERF/CLERP model

* Step 3: Quantitative screening

RM PPA Wo*08op M , MD, 200 Sh08 • A7 C , 77 D R

Tas 4 & 7 - Qqalt.Uv.Vtrnnltfr. screeng Re (RE8) & S. P.,S. hnlef (708RI

Slide 13 Notes:

! Task 7: Quantitative Screening Steps in Procedure/Details

Step 1: Quantify CDF/CCDP models.

* Step 1.1: Quantify CCDP model

- Fire-induced initiators are set to TRUE (1.0) for each fire

compartment, CCDP calculated for each compartment

- This step can be bypassed, if desired, by using fire frequencies in

the model directly and calculating CDF

Fik PRO Wýrn'h, 8080084d MO, 2008 W6 A 080ab8ra0.n 7777 NRC 0808 aNsk.,77RQMt8, TaSk 4 4 ?- Qua77t5wve0uanUtkfa .Screning • R ES ) E ~ctaP~e ch IR tc n sn.0804e 770R0

3-70
Task 7: Quantitativ

Steps in Procedure/D

Step 1: Quantify CDF/C

- Step 1.2: Quantify CD

- Compartment fire-induc

compartment CCDPs fr

- Step 1.3: Quantify ICE

- ICDP includes unavaila

routinely

- Recommend this be do

management

RPe PRA WOr*0hop, Bethaoda MD. 2008

fT, 4 & 7 - Oof.I 8. .. fft0thVe So,,.nIng

Slide 14 Notes:

,e Screening

etails

'CDP models.

3F

ced initiator frequencies combined with

rom Step 1.1 to obtain compartment CDFs

DP (optional)

bility of equipment removed from service

ne if will use PRA for configuration

Sli ded 1A boI4nt U S, NRC (fl .8 N080. R.ouoatr

RSlidh e8ES) & 60w4 R.4.0*h nOft. (EPRl) Notes:c 

Slide 15 Notes:

! Task 7: Quantitative Screening Steps in Procedure/Details

Step 2: Develop LERF/CLERP models.

* Exactly analogous to Step 1 but now for LERF and CLERP

Ftm PRA Worsop, 8000,0al MD, 2008 0 A Co800o0001 00.0.0NR 08 Of Nud.., R15u2.10

Task 4 & 7- Ou.Otattae'8urnfiftfv. S-neng - eSe IR- S) & 04 dl1 P * . Rs t, I (EPRR 

3-71
Slide 16 Notes:

Task 7: Quantitative Screening

Screening Criteria for Single Fire Compartment

Step 3: Quantitative screening, Table 7.2 from NUREG/CR-6850

Quantiflcation Type CDF and LERF ICDP and ILERP

Compartment Screening Compartment Screening

Criteria Criteria (Optional)

Fire Compartment CDF CDF < 1.OE-7/yr

Fire Compartment CDF ICOP a 1.OE-7

With Intact Trains/Systems

Unavailable

Fire Compartment LERF LERF < 1.OE-8/yr

Fire Compartment LERF ILERP a 1.OE-8

With Intact Trains/Systems

Unavailable

Fre PP. Workshop, aeshaWa MD, 200a [ 16 A voflabotic,,fdat NRC oNucldnc.esndarRe

raSx 4 &7- OQmaaeamtmaa•uaeea sMt••...,rrc (RES) &a Deas aP R-aah faace (EPRIs

Slide 17 Notes:

! Task 7: Quantitative Screening Screening Criteria For All Screened Compartments

Step 3: Quantitative screening, Table 7.3 from NUREG/CR-6850

Quantification Type Screening Criteria

Sum of CDF for all screened-out fire compartments < 0.1 * (internal event average CDF)

Sum of LERF for all screnned-out fire compartments < 0.1 * (internal event average LERF)

Sum of ICDP for all screened-out fire compartments < I.0E-6

Sum of ILERP for all screened-out fire < 1.OE-7 compartme nts

Fire PRA Workshop, Bethea, Mo , 2008 v A Ceerarflaon. rUNR Office osNceerRenccacc, Task 4 & 7 - Oufiatoe/Quam V = Ocroenena Rssv -ena nR a F.Oetr• PM (.Evn cdcR fl/t)

3-72
Task 7: Quantitati'

Bases for Values

Bases for quantitative s

NUREG/CR-6850

* Premise is that most

* Increase in CDF less

increase in R.G. 1.17'

* Sum of CDF from scr

limited to 10% of total

* Individual compartme

CDF

Ft. PRA Wo,*ash, Bethesda MD, 2008

Ta•k 4 & 7 - Ouf VR-atla. Sca.0ln

Slide 18 Notes:

ve Screening

screening criteria provided in App. D to

CDFs are -1.OE-5/yr

than 1.OE-6/yr is defined as very small

4

eened-out compartments therefore

CDF

nt limit set at 1.OE-7/yr, or 1% of total

A Cc°•bý,a' /U NRC ýOake Nd - Re0aRu"

Reac-h (RES) & Eý cPOl, Rema-h q e(EPRo

Slide 19 Notes:

! Task 7: Quantitative Screening Bases for Values (cont'd)

Basis for LERF values same as for CDF, but factor of 10

lower

SICDP screening criterion of 1.OE-6 based on temporary

change risk criterion in EPRI PSA Applications Guide,

EPRI-TR-105396

Similar basis for ICLERP criterion of 1.OE-7

Fim PRA Workshop, Bethesda MD, 2008 • A - -Nbst US NRC 0ffice ° 0 N g0/tbv

Task 4 & 7 - Ou•,uls uar-Q-a5N Scmns/g -05s0 (-ES) &. B -/ R actP / P-ROj

3-73
Slide 20 Notes:

Task 7: Quantitative Screening A word of caution...

* Note that the quantitative screening criteria are a point of

ongoing discussion, e.g.:

- Should fire criteria be based on internal events totals

or based on fire totals?

* ASME/ANS standard provide guidance regarding

quantitative screening

- Anticipate that changes may be forthcoming

Fft PRA Wo,*.•op, BetEda MD. 2008 2 A Co#.bn0anofUS NRC 08RA o0NcM.,RqM!o0

Tlk 4 A 7 - Qu-I.Uv.VuanfltN Sw.. g Rflnwh (RES) & EIfl POferRfl0 08M12 (E000

3-74
Session 4: Screening Post-Fire 

HRA

3-75
Slide 1 Notes:

EEI'I•IL~c.,0ow.a RESEARCH INSTITUTE

EPRIINRC-RES FIRE PRA

METHODOLOGY

Task 12a - Screening Post-Fire HRA

Joint RES/EPRI Fire PRA Workshop

2007

Palo Alto, CA

I A Collabortion of U.S. NRC Office of Nulear Reglatory Research (RES) & Electrc Power Research Institute (EPRO

Slide 2 Notes:

! HRA Screening Scope

Task 12a: Post-fire HRA (screening)

- Identifying applicable post-fire human failure events and

establishing screening values used during the running of the Fire

PRA model

Fim PRA Workwop, 2007, Pah .Ao, CA j A C -ioOfUS NRC OffNklso`Nw1RnSGty

Task 12a. Screning Postirse KRA R.• h (R-S) IEldc P- - r R - s-N e (U-1)PRI)

3-76
Slide 3 Notes:

r Task General 12a: Objectives Post-Fire HRA (Screening)

Purpose: identify reasonable and feasible human actions and

resulting HFEs to include in Fire PRA, and assign screening HEPs to simplify the model and focus analysis resources appropriately.

* Addresses screening values based on: - Whether a prior analyzed Internal Events HFE vs. a new fire-related HFE

- Potential effects of fire scenario for which Internal Events HFE is applied - Timing considerations for new fire-related HFEs

* Accounts for fire-scenario-induced changes in assumptions,

model structure, and performance shaping factors

* Addresses need to use procedures (e.g., FEPs) beyond those

modeled in the Internal Events PRA

* Does not address pre-initiator HFEs that are handled within the

data used in Tasks 6, 8, and 11

Fft PRA Ktkshop, 2007. Aro Afc 55 CARa _ Coteo• oPU5 •y

r40It-Sh a wd.. Pw pjpq k n. (RýtS) A lhpnRSe ~ e (EPR

Slide 4 Notes:

Task 12a: Post-Fire HRA (Screening)

Inputs/Outputs

Inputs from other tasks:

- Mitigating equipment and diagnostic indications from Task 2 (Fire

PRA Component Selection),

- Human actions already in PRA (because of internal events

modeling) from Task 5 (Fire-Induced Risk Model),

- Information may be used for identifying equipment failures, spurious

operations and indications from Tasks 3 (Fire PRA Cable Selection),

9 (Detailed Circuit Failure Analysis), 10 (Circuit Failure Mode

Likelihood Analysis), 8 (Scoping Fire Modeling), and 11 (Detailed

Fire Modeling) as available, so as to determine proper screening

criteria to be used

FPh PRA W000,hop, 2W07, Pa0. , C A C5borh055S NRC Oawaohd.WRegs.Ry,

.as fs.. -.Se. gPost.hn NRA R 0 - (RES( (-) A 1 EACPo,. Ref0ý (EPR3-77
Slide 5 Notes:

! Task 12a: Post-Fire HRA (Screening) Inputs/Outputs (continued)

- Outputs to other tasks:

- May identify human actions implying other equipment and

indications to be added in Task 2 (Fire PRA Component Selection)

and thus modeling additions in Task 5 (Fire-Induced Risk Model)

- Provides screening HEPs for Task 7 (Quantitative Screening)

Fn. PRA Wo*kMvo, 2007. Pao Alt, CA A bo•n U S NRC Offlc -lNUCe RVeOtatoey

T..k 12. Scfeenng poSt•Fcle HRA R-.r (RSS) & E-.Ctrcp - eSflh s-- .- (EPRI)

Slide 6 Notes:

! Task 12a: Post-Fire HRA (Screening) Steps In Procedure

Two major steps:

* Step 1: Modify and add HFEs to the model

* Step 2: Assign quantitative screening HEPs

Fire PRA Workeop, 2007, Pa Afto CA e A C -oeee 1e RI U.S NRC 0RN dNkRecR00

Task 12.. Seee*nlng Pr-Wl.P HRA ResOh (RES) & EleHRctri Re4sa nSRA, lEtS)

3-78
Slide 7 Notes:

! Task Steps 12a: Post-Fire HRA (Screening) In Procedure/Details

Step 1: Modify and add HFEs to the model.

- Step 1.1: Review existing Internal Events HFEs and modify

as necessary

- Many existing HFEs will remain as is except for screening value

- Some existing HFEs may need to change such as due to the use of

different procedures, possible fire environmental effects, or different

scenario timing due to fire

- Step 1.2: Add new fire-unique HFEs

- Primarily from fire-specific procedures

- Actions taken in response to spurious (erroneous) indications

Ffl PRA Wowksh•o. 2007, Pao Afto, CA 7 A C-ob0OAn &US N ON/ic of N00.,r WC RWatoPY

lsk 12, . ScýnM g P0*F/n HRA Research (-/N) & ElefI/ pI.- , -.- 1,-,tu0 (EPRI)

Slide 8 Notes:

Task 12a: Post-Fire HRA (Screening) Steps In Procedure/Details

The following are important elements of the identification

process:

* Expected steps taken in response to fires in specific

compartments

* Comparison of fire response actions to EOP actions

* Consider fire-specific training, if information is available and

relevant

* Role of each crew member during fire scenario

* Fire-specific informal rules that are part of crew knowledge

Ff. PRA Wo0/00 p, W120 , Plo A0to, CA A Ga7aON~ U. NRC OffiA-e ANuO& R00'0N,

T'.. 10.. S..n.ng f0t.-F00 HRA R__h WROS) . 81.c- Pow-r R-00.h -ns10uf (-PRI)

3-79
Task 12a: Post-Fir

Steps In Procedure/I

Step 2: Assign quantitati

specific basis)

Four sets of screening

- Set 1: multiply internal

potential fire brigade in

workload/distraction iss

- Set 2 (spurious events

safety-related train/divi

10 times original value

across scenario.

- Set 3: applies generall

meeting Set 1 or 2. Us

hour of fire initiation. U

- Set 4: applies to new H

screening value of.1 .0.

Fft PRA Wo.r*w, 2007. PaS At CA

rA 0120. - S-,.g Po0-r*P HRA

Slide 9 Notes:

re HRA (Screening)

Details

ve screening HEPs (on a fire scenario

criteria :

events HEP by 10 to account for effects of

teraction and other minor increased

sues. Examine dependencies across scenario.

could have impact but to only one critical

sion): increase internal events HEP to 0.1, or

whichever is greater. Examine dependencies

y to new HFEs but also to existing HFEs not

a 1.0 if action has to be performed within one

Ise 0.1 otherwise.

lFEs associated with MCR abandonment. Use

Slide 10 Notes:

Task 12a: Post-Fire HRA (Screening)

Bases for Screening Values

Values have no direct empirical bases. Bases are:

* Experience with range of screening values used and

accepted in HRA

* Experience in quantifying HEPs for events in nuclear power

plant HRAs

* Experience applying range of HRA methods and values

associated with those methods

* Experience performing HRA for Fire PRAs, including pilots

* Peer comments

* Not so low so as to miss potential dependencies among

HFEs

FRM PRA Wor*sh•p. 2001, Pao Afto, CA A CA0*cl.t . NRC 0f7. -f0.. 77 Nod•r Ropat~oy

T.sk ,2a. sneonrng Pos~t-PO. HRn 00700000 ) & El0ti p., 00s.0 IfstA. (0p00

3-80
Session 5: Post-Fire HRA Detailed Analysis

3-81
Slide 1 Notes:

jt~~I LECIR IC POWER. C raII RESEARCH iNSTIUTE

EPRI/NRC-RES FIRE PRA

METHODOLOGY

Task 12b - Post-Fire HRA Detailed

Analysis

Joint RES/EPRI Fire PRA Course

September and November 2008

Bethesda, MD

A CollIboretion of U.S. NRC Office of Nuclear Regulatoy Re.00,h (RES) & Electric Poe- Reseemrh Institute (EPRO

Slide 2 Notes:

! Post-Fire Scope HRA Detailed Analysis

Task 12b: Post-Fire Human Reliability Analysis (Detailed

Analysis)

- Obtaining more realistic human error probabilities (i.e., not

screening values)

JoWnt FiO PRA Cou-e, Sept. & No,. 2008 A Cdlbrstkoe of UW S NRC OtteeWNu(RMgeN Re toWV

7.0 12b - Pea-Fi. NRA DPtelld Analyels Reersah (RES) & 6 oft Fs R wereR. - res Iate EPRI)

3-82
Slide 3 Notes:

Task 12b: Post-Fire HRA Detailed Analysis General Objectives

Purpose: assign best-estimate HEPs to allow more realistic

estimate of fire risk.

- Current procedures do not specify an HRA method to use

- There are too many methods analysts might use (THERP, ASEP,

CBDT...) and each is unique in what it treats and how to determine

HEPs

- Hence, procedure outlines what should be addressed but not how to

specifically incorporate into existing HRA methods because there are

too many of them

* Addresses fire-scenario-induced changes in assumptions,

model structure, and performance shaping factors

-Addresses need to use procedures (e.g., FEPs) beyond

those modeled in the Internal Events PRA

JoFir* PRA & Mc2M Cooonioi a NRC 00. RAg-- "foy

UMsk 12hb - Po HRA ONObd Abnt F I (RE S) & EpI o. ý (ý

Slide 4 Notes:

Task 12b: Post-Fire HRA Detailed Analysis Inputs/Outputs

Task inputs and outputs:

* Inputs from other tasks: feedback from Task 7 (Quantitative

Screening) identifying HFEs needing detailed analysis

* Outputs to other tasks: best-estimate HEPs for Task 14 (Fire

Risk Quantification)

Joint Fm PRA CouS, Sepf & Nft. 2008 a C o--; Uoi ofUS. NRC O(EioRNM R y IaRC 

Tas 12b - fost-Fm NRA DHtýWd A-"ys1s Roi.h0 (RES) & P000C pR RSrch 00800 (ERI)

3-83
Slide 5 Notes:

Task 12b: Post-Fire HRA Detailed Analysis ASME PRA Standard & Possible Team Addition

Overall approach is not new

- Continue to follow the basic HRA approach addressed in

ASME/ANS PRA Standard (RA-S-200X) including both Part 3 (fire)

and cited requirements from Part 2 (internal events)

- Recommends individual with experience in human behavior during

fires (firefighter trainers, etc.) be involved in quantification IF useful

for safe shutdown considerations (e.g., for local actions)

- But need to recognize the difference between operator safe

shutdown actions generally in the MCR vs. fire-fighting actions in the

vicinity of the fire

Jo400 FP PRA Corn., Sept. & No-. 2008 F 5 * 0000000 - o0 .. NR0 -t-1- 000. 000,u.0 RA 

natt 12h - PoW•F O e0 W48 APntlP. ft-.0 r(E) s I0 •. - -- - (00Rq

Slide 6 Notes:

Task 12b: Post-Fire HRA Detailed Analysis PSFs and Fire Effects to Consider

Guidance focuses on identification of fire-relevant

performance shaping factors (PSFs) and potential interactions among the PSFs (fire conditions could make

PSFs different than those for internal events):

-Available staffing resources - Fire situation may need more staffing than responding to an internal

event

- "Nominal" staffing for internal event could be less than adequate for

fire

* Applicability and suitability of training/experience

- Extent of familiarity/training may be less for fire than for internal

events for both in-MCR and local actions

more...

Joint FiP PRA Co-s. Sept. & Nov. 2008 [M~e ] A Co/aJoratono U.0 NRC Office 0f Nud&ea Regatory

T&O 12b - P0t.-FPh HRFA DMta(ihd Anspal 0.R0.00 (RES) & E00. p-•. R.0r.h 00*00. (5PR0)

3-84
Slide 7 Notes:

Task PSFs and 12b: Fire Post-Fire Effects to HRA Consider Detailed Analysis

Suitability of relevant procedures and administrative

controls

- Fires may require multiple procedures be used at the same time

(e.g., EOPs and Fire Procedures) that may be more burdensome

and together take more time to implement

- There may be less detailed or no procedures available for some

actions (e.g., local action steps are not spelled out but require more

skill-of-the-craft or memory)

-Availability and clarity of instrumentation

- Possibility of spurious or failed indications more likely for fires than

for internal events

Job,! Fe FAA Coun, S.7 , 8 Nov. 2008 7 0 Cd8000000 AbS NRC Offfb a OU MOY

Tas 12b5M'!, 12b- P000-F/r ARA HM ýftlh A- oyobs (RES) . FMbcPoe R ... -r - (iF0R)

Slide 8 Notes:

Task 12b: Post-Fire HRA Detailed Analysis PSFs and Fire Effects to Consider

Time available and needed to complete action, including

impact of concurrent and competing activities

- Timing of scenario could be different from comparable internal events

scenario due to spurious events and introduction of new/different

procedures and actions

- Actions themselves may have different execution times (e.g., have to

disable before reposition, may require more sequencing of actions,

etc.)

* Environment in which action is to be performed

- Fires can introduce new considerations (smoke, heat, chemicals,

toxic gases...)

* Accessibility and operability of equipment

- Fires can eliminate or delay the ability to take actions due to

accessibility or damage considerations

Joint Fir. PRA Cou.-, Spt. & Nov., 2008 A WlMaboOn 0f S ORb Ol• obeaiR.5bM01o

Task 12b -PosFFim HRA Detae Analyis -e~mh (RES) & E-Otb Po RSt -0t0ute (EPRI)

3-85
Task 12b: Post-Fir PSFs and Fire Effects

* Need for special tools - Fires may increase the

protective clothing, ladi

- Need to ensure access

execution of similar act

* Communications

- Could be greater dema

runners)

* Team/crew dynamics

- Potential for different re

timing of plant status ch

* Special fitness needs - Ensure no new fitness

Jot FF0 PRA Goo COP?. W, & Nov. 2008

Task 12b - Post-Fr HRA Deiiled A-&y.1s

Slide 9 Notes:

e HRA Detailed Analysis to Consider

and clothing se requirements (e.g., breathing gear,

ders, keys...)

to these, and consider potential increase in

ions than that estimated for internal events

nd and potential need for different form (e.g.,

and crew characteristics

les/responsibilities, less frequent or different

hecks (could affect recoveries), etc.

needs required (especially for local actions)

RSlidh tES) 1 eNArO p. RNotes 00100?. lEPR))

Slide 10 Notes:

Task 12b: Post-Fire HRA Detailed Analysis

MCR Abandonment

Should consider as part of the PSF evaluations:

- Procedural/training approach and explicitness/clarity of

criteria for abandoning MCR

- Could be confusion about need to evacuate MCR

- Impact of crew reluctance to abandon MCR

- Timeliness of decision and problems associated with delays in

abandoning MCR

- Inappropriate abandonment of MCR (e.g., premature or less viable

option)

Joinl Fhe PRA Co-rn, S.0 t. & NO. 2008 1 A COJI-o, dVoofUS, NRC Ol1eOf -u1 Rn&,uatoy

tea 02b - POSt.iMe H0A 0Detaile Aalyis -.. h (RES) I E--ti P00r - R00h Ist,- - (-IER)

3-86
Task 12b: Post-Fi MCR Abandonment (

" Effects of crew no lone

indications and the inf

* Number and complexi

out subsequent activit

* Number of different lo.

" Extent to which multip

sequentially performe

-Ability to communicate

* Need to wear breathin

-Adequacy of human-n

and local panels

Jort Firn PRA Co.ur-e Se. & No, 2008

Slide 11 Notes:

re HRA Detailed Analysis

cont'd)

ger having access to complete MCR

ormation they provide

ty of actions to shift control and carry

ies

cations to be visited

le actions need to be coordinated or

d

e between different locations

ig apparatus or special clothing

iachine interface at remote shutdown

0 A0 * CoiCorat W o,oV S. NRC Offo &u , R0,gud*Wy

e (RES) A Noteos

Slide 12 Notes:

Task 12b: Post-Fire HRA Detailed Analysis

Cases Where Little or No Credit Should be Allowed

* Tasks needing significant interaction/communication

between individuals wearing SCBAs unless can be justified

as not a problem

* Fire causes numerous spurious actuations (or stops) of

equipment including instruments

* Actions performed in fire areas or requiring travel through

fire areas

* Actions requiring use of damaged equipment

* Actions without procedural direction or training, lacking

necessary tools, or with inadequate time available

Joint Fo. P94 Cou,,e, Sept & N-p 2008 [-S;;l. 12 04 . S. NRC OC3AN00O N u R R O• 0Ay

T- 12b - plsfat, H"A OeooWood Anmrsl$s 6. WRS) 000 tr00 006.- 6•r3h 04 X0 0060

3-87
Slide 13 Notes:

! Task 12b: Post-Fire HRA Detailed Analysis Documentation

Product of this task is a calculation package, which should

contain (per the ASME/ANS PRA Std. Part 3):

-All human actions and HFEs considered, including

descriptions in context of fire scenarios

* Quantification approach and method/tools used

* HEP results and bases for HEP calculations, including

dependencies, PSFs, and uncertainty

* Important sensitivities

JoWi Fkv PRA Couws, Sept. & No,. 2008 [iE ] A C (REbS)An ofU. NRC -ft-hl 1-Nu Rg)ataV

rlAO 12b - PolFl HRA DotbMd AW•Ilys iRES Res.h & E&000pows Refl.h Inste (EPR/)

3-88
Session 6: Fire Risk Quantification

3-89
Slide 1 Notes:

ELECTI CPOWER

RESEARCM INSTITUTE

EPRI/NRC-RES FIRE PRA

METHODOLOGY

Task 14 - Fire Risk Quantification

Joint RES/EPRI Fire PRA Course

September and November 2008

Bethesda, MD

A Collabortion of U.S. NRC Office of Nucle Regulatory Research (RES) & EIoct,* Powe Research MOtWitte (EPRO

Slide 2 Notes:

! Fire Risk Quantification Scope

Task 14: Fire Risk Quantification

- Obtaining best-estimate quantification of fire risk

Jo, Fitr PRA Cou-, Sept. No' 2008 , A Co#SbOmtlO ofi1S NRC 000. ofN0.i R~gulatOy

Trsk 14 - Fin Risk QU..nRoakon R.esearh (RES) 1EltPower Re0.h ini0ute (-90

3-90
Slide 3 Notes:

Task 14: Fire Risk Quantification General Objectives

Purpose: perform final (best-estimate) quantification of fire

risk

* Calculate CDF/LERF as the primary risk metrics

* Include uncertainty analysis / sensitivity results (see Task

15)

* Identify significant contributors to fire risk

* Carry along insights from Task 13 to documentation but this

is not an explicit part of "quantifying" the Fire PRA model

* Carry along residual risk from screened compartments and

scenarios (Task 7); both (final fire risk and residual risk) are

documented in Task 16 to provide total risk perspective

, fle mot PRA Co-r SW1. & fto. 2008 * C o nOt UG NRC0Oo It&t0r0 Ts* 14 -_ Rr o Quanftifaton Reserc P f ,00R*. .RE IPPRO El.,8i 

Slide 4 Notes:

Task 14: Fire Risk Quantification

Inputs/Outputs

Task inputs:

- Inputs from other tasks:

- Task 5 (Fire-Induced Risk Model) as modified/run thru Task 7

(Quantitative Screening),

- Task 10 (Circuit Failure Mode Likelihood Analysis),

- Task 11 (Detailed Fire Modeling), and

- Task 12 (Post-Fire HRA Detailed Analysis)

Joint Ffm PRO Cou-, Sept & Nof. 2W00 Ste4 ] A ofUS. UrO.0n WRC 000. i Rn.AlMoV

Tas. 14-. 8ire RsQntfic.ton R1&Et ROearth -~* (REH. _ o - (EP-0

3-91
Slide 5 Notes:

Task 14: Fire Risk Quantification Inputs/Outputs

Output is the quantified fire risk results including the

uncertainty and sensitivity analyses directed by Task 15

(Uncertainty and Sensitivity Analysis), all of which is

documented per Task 16 (Fire PRA Documentation)

JoInt e PRA Co-rse, Sept. & 2008 j A - Co n 1 - RC - e e uO/ R"g.a/R

Task 14 - Frse Risk Qufleffon R --- (-) & -- - - (-I)

Slide 6 Notes:

! Task 14: Fire Risk Quantification Steps in Procedure

Four major steps in the procedure*:

- Step 1: Quantify CDF

- Step 2: Quantify LERF

* Step 3: Perform uncertainty analyses including propagation

of uncertainty bounds as directed under step 4 of Task 15

* Step 4: Perform sensitivity analyses as directed under step

4 of Task 15

In each case, significant contributors are also identified

Joi Fr PRA Coure C Sept Nov, 2008 R AC Co od1SNRC ORCceN/ssR/eeRoRWry

Task 14 - Fire Risk Cuenttfcaton M - Reý /RER/ & Sec/Cý pR ReR hf CC//C/e (ERR))

3-92
Slide 7 Notes:

! Task 14: Fire Risk Quantification Quantification Process

Characteristics of the quantification process:

* Procedure is "general"; i.e., not tied to a specific method

(event tree with boundary conditions, fault tree linking...)

* Can calculate CDF/LERF directly by explicitly including fire

scenario frequencies or first calculate CCDP/CLERP and

then combine with fire scenario frequencies

* Quantification is to.be done in conformance with relevant

ASME PRA Standard (RA-S-200X) requirements and

supporting requirements - Fire section (Part 3) references internal events section (Part 2) for

most aspect of risk quantification

Jort F PRA C o , S ef, & Nov. 2008 • A C o ata U 000U . NR O OfN i

0 0 7 4 -- Risk o m ultiun R rES) . E-COCP .. s 6v0*k 0200

3-93
Session 7: Uncertainty and Sensitivity Analysis

3-94
Slide 1 Notes:

/-> 1E1(21 CliINSPITUTE

EPRIINRC-RES FIRE PRA

METHODOLOGY

Task 15 - Uncertainty and Sensitivity

Analysis

Joint RES/EPRI Fire PRA Workshop

September and November, 2008

Bethesda, MD

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

Slide 2 Notes:

Task 15:Uncertainty and Sensitivity Analysis

Purpose

Purpose: Provide a process for identifying and treating

uncertainties in the Fire PRA, and identifying sensitivity

analysis cases

* Many of the inputs to the Fire PRA are uncertain

* Important to identify sources of uncertainty and assumptions

that have the strongest influence on the final results

* Fire risk can be quantified without explicit quantification of

uncertainties, but the risk results cannot be considered as

complete without it

* Sensitivity analysis is an important tool in uncertainty

assessment

Fir, PRA WOS-op. hoptsda MD, 2008 A CO0tioO 0(U0 NRC Offic• NueRegulatory

T-k 15 - Uncertaly and SonsitiAnta-Jysis Reah (RES) &fEdt Power Re..hfln0080e (fPR0

3-95
Slide 3 Notes:

Task Scope 15:Uncertainty and Sensitivity Analysis

Scope of Task 15 includes:

-Background information on uncertainty

-Classification of the types of uncertainty

-A general approach on treating

uncertainties in Fire PRA

FRe PRA Wo*Sog. aeftWda MO, 2008 _ A &o NM.0 RVg-aoY

T i$. 13 - n00.n S-fJifwi Anay•as 000.5,00 (_ES) * Elet Power R ,Oa (Ep_ 0

Slide 4 Notes:

Task 15:Uncertainty and Sensitivity Analysis Types of Uncertainty

Distinction between aleatory and epistemic uncertainty:

- "Aleatory" - from the Latin alea (dice), of or relating to

random or stochastic phenomena. Also called "random uncertainty or variability."

- Reflected in the Fire PRA models as a set of

interacting random processes involving a fireinduced transient, response of mitigating systems,

and corresponding human actions

- "Epistemic" - of, relating to, or involving knowledge;

cognitive. [From Greek episteme, knowledge]. Also

called "state-of-knowledge uncertainty."

- Reflects uncertainty in the parameter values and

models (including completeness) used in the Fire

PRA - addressed in this Task FR. PRO Wo , 800h0.a MO,2008 e 4 CoaOýon o, us. NDC 02W8 a, ua y

TSk 15. U•rltj ard Sana)v• yAnafIs R.08h (RCS) & E-tokiP p s R 0 ,50.h SPt )

3-96
Slide 5 Notes:

Task 15:Uncertainty and Sensitivity Analysis Inputs and Outputs

Inputs from other Tasks:

- Identification of sources of epistemic uncertainties from Tasks 1 through

13 worthy of uncertainty/sensitivity analysis (i.e., key uncertainties)

- Quantification results from Task 14 including risk drivers used to help

determine key uncertainties

- Proposed approach for addressing each of the identified uncertainties

including sensitivity analyses

Outputs to other Tasks:

- Sensitivity analyses performed in Task 14

- Results of uncertainty and sensitivity analysis are reflected in

documentation of Fire PRA (Task 16)

FkI p/A Workshop Balh/ . O. / 2008 5 A C fRSo , 4lliS NRC 0f10 -N h tnV&UR tu

Tmak 15. /Uceainly and 5nsai//f5 Ana"Is Resa,00 fRESt &Eflctc R/ a n~uf h l (EPRO

Slide 6 Notes:

I Task 15:Uncertainty and Sensitivity Analysis General Procedure

Addresses a process to be followed rather than a pre-defined

list of epistemic, uncertainties and sensitivity analyses, since

these could be plant specific

*Step 1: Identify uncertainties associated with each task

-Step 2: Develop strategies for addressing uncertainties

-Step 3: Review uncertainties to decide which uncertainties

to address and how

-Step 4: Perform uncertainty and sensitivity analyses

-Step 5: Include results of uncertainty and sensitivity

analyses in Fire PRA documentation

Fi. PRA WPA*shop, Belhesda MD, 2008 6 A Conabatlkn of U.S. NRC 00 0 N/Jea Reg•palty

Task .5 -Uncertiy and Senslvaply Analysls Res h (RES) , Ee00p - a E• adhta (EPRO

3-97
Slide 7 Notes:

Task 15:Uncertainty and Sensitivity Analysis Steps in Procedure/Details

See Appendix U to NUREG/CR-6850 for background on uncertainty analysis. See Appendix V for details for each

task.

Step 1: Identify epistemic uncertainties for each task

* Initial assessment of uncertainties to be treated is provided

in Appendix V to NUREG/CR-6850 (but consider plant

specific analysis for other uncertainties such as specific

assumptions)

* From a practical standpoint, characterize uncertainties as

modeling and data uncertainties

* Outcome is a list of issues, by task, leading to potentially

important uncertainties (both modeling and data uncertainty)

FR . PRA l00 0 ° .00 80 B M h0 MD. 2008 A A "°00rO,0 n8U S NMRC 8tt 8.0,N 0 0• °

Thfl is - AUcfrn dap sI tAnatys s Refe •(ES) & dEtt Pým.SoDeRO

Slide 8 Notes:

! Task 15:Uncertainty and Sensitivity Analysis Steps in Procedure/Details

Step 2: Develop strategies for addressing uncertainties

* Strategy can range from no action to explicit quantitative

modeling

* Each task analyst is expected to provide suggested

strategies

* Possible strategies include propagation of data

uncertainties, developing multiple models, addressing

uncertainties qualitatively, quality review process, and basis

for excluding some uncertainties

* Basis for strategy should be noted and may include

importance of uncertainty on overall results, effects on

future applications, resource and schedule constraints

Fie PRA Woahep. Beheede MD, 2008 a coqae, eoUD NRC Office olNu de Iey T {1.5* -5 Uncertiny and Seernlt~y Anatyes Re~eamhS (RED) 8 EMOetn Power Resiear, Er,•fue (EPRN

3-98
Slide 9 Notes:

Task Steps 15:Uncertainty in Procedure/Details and Sensitivity Analysis

Step 3: Review uncertainties to decide which uncertainties to

address and how

* Review carried out by team of analysts familiar with issues,

perhaps meeting more than once

* Review has multiple objectives:

- Identify uncertainties that will not be addressed, and reasons why

- Identify uncertainties to be addressed, and strategies to be used

- Identify uncertainties to be grouped into single assessment

- Identify issues to be treated via sensitivity analysis

- Instructions to task analysts to perform the analyses

Fan PRA Wo*shoP, Betheasda MD. 2W&8 J ACbo "o NC Offc.ohNctaaR Teakt 15. - U ABnc.ertait S.,p.diappyAnalysis BP~srh S) BF& F P.Bri P• ~ e to sCt~daiPaPd iPP

Slide 10 Notes:

! Task 15:Uncertainty and Sensitivity Analysis Sensitivity Analysis

Sensitivity analysis can provide a perspective that

cannot be obtained from a review of significant risk

contributors.

- Each task analyst can provide a list of parameters that had the

strongest influence in their part of the analysis

- Experiment with modified parameter to demonstrate impact on the final risk results

- Modeling uncertainties can be demonstrated through sensitivity

analysis

- Sensitivities should be performed for individual uncertainties as well as for appropriate logical groups of uncertainties

Fire PRA Wof,•,h Bethesda MD, 200P [li. 1 A CPAPWboronofU.S. NRC Oifire o(NPDePRSuSaGW

Task 15. Uncertainty and Senttvilty An~aysSs -esa, (PES) & 000t0 P1e- R -eh fNsIPP (EP-)

3-99
Task 15:Uncertain Steps in Procedure/D

Step 4: Perform uncert

* Uncertainty analyses - Quantitative sampling - Manipulation of model - Qualitative evaluation

* Following items shoul - Uncertainties being ad - Strategy being followe - Specific methods, refe

(to allow traceability)

- Results of analyses, in

of Fire PRA - Potential impacts on a

Fir PRA WOMo, Bes•esa l MD, 2008

na,* 15 - UIcwfrmnf and S•. )y•v ma s

Slide 11 Notes:

ty and Sensitivity Analysis

)etails

ainty and sensitivity analyses

may involve:

of parameter distributions s to perform sensitivity analyses

of uncertainty

d be made explicit:

dressed

d

rences, computer programs, etc. being used

cluding conclusions relative to overall results

nticipated applications of results

A -BP) d .S. NtPMWR a dl 00 Ra

Slide 12 Notes:

! Task 15:Uncertainty and Sensitivity Analysis Steps in Procedure/Details

Step 5: Include results in PRA documentation

- Adequate documentation of uncertainties and sensitivities is

as important as documentation of baseline results

- Adequate documentation leads to improved decision-making

- Documentation covered more fully under Task 16

Ff. PRA Wor*shop, Bethesda MD, 2008 dE 2 A C (REOSao of& U. NRC Offlct N, RBG -d oy

Taak 15 - Unceranty and Senslty Anabmis Re~atrB (RES) & eCO*POwerReserh It PR)

3-100
Task 15:Uncertain

Expectations

Minimum set of uncerta

treatment:

- Fire PRA model structur

to how fires could result

outcomes (Tasks 5/7)

- Uncertainty in each signi

- Uncertainty in each signi

- Uncertainty in each signi

- Heat release rate

- Suppression failure mod

- Position of the target se

- Uncertainty in each signi

- Uncertainty in each sequ

frequency based on the

significant equipment fail

F-n P-A Wo*rshop, eeUlesda MD, 2O8

Task 15. Ufnceinty and S-nsýkMvl AnýlwJ'

Slide 13 Notes:

ty and Sensitivity Analysis

inties expected to have a formal

e itself, representing the uncertainty with regard

in core damage and/or large early release

ficant fire ignition frequency (Task 6)

ficant circuit failure mode probability (Task 10)

ficant target failure probability (Task 11)

tel and failure rate

t vs. ignition sources

ficant human error probability (Task 12)

ence core damage and large early release

above inputs as well as uncertainties for other

ures/modes (Task 14)

S ide A Cdo Lk±ofUS NRC ONot RnAtty es:l 

Slide 14 Notes:

Task Expectations 15:Uncertainty and Sensitivity Analysis

* Other uncertainties may be relevant to address

* Sensitivity analyses should be performed where

important to show robustness in results (i.e., demonstrate

where results are / are not sensitive to reasonable

changes in the inputs)

* While not really a source of uncertainty, per se, technical

quality issues and recommended reviews are also

addressed

Fire PRA Workso, Behesda MD, 2008 1 A Co•t'.rof URS. NRCO N u RegkItv

Task 15. MnerftW ln Rnd S.sbOIIW YAnayihs Reseac (RES) . EkPcrc (-F0

3-101

Appendices

Appendix A: Questions Asked in Module 1 Sessions

NRC Disclaimer: Appendix A is intended solely for use as part of a training tool. No portion

represents NRC Conclusions or Regulatory Positions, and should not be interpreted as such.

Session la: Fire PRA Component Selection

QUESTION

Given that spurious opening is a high-consequence event, would doubles and triples be

considered in the PRA?

RESPONSE

The relevant guidance addresses doubles, but provides no numerical upper limit for

consideration.

QUESTION

For consideration in the PRA, why aren't small LOCAs considered?

RESPONSE

They are! We're not concerned about pipe-break LOCAs, but we are concerned about any

other event, like a spurious opening of the PORV, that can result in the same thing, like a smallbreak LOCA, or a medium-break LOCA, or a large-break LOCA.

QUESTION

Should we include the auxiliary feedwater pumps' test-line isolation valves in the fire PRA

equipment list?

RESPONSE

No. The system description says that they are not a significant diversion of flow.

ASIDE: DUAL FAILURE MODES

Both failure modes of a particular component, like failure to open and spuriously opening, must

be considered in the model and potentially each of those failure modes will involve different

cables. The circuit analysts will need that information, so when identifying information for the

PRA equipment list, identify each failure mode.

QUESTION

Are most components required for the PRA included in the fire PRA?

A-1
RESPONSE

In most cases, this is correct, but additional information is needed. In addition, there may be

certain situations on which you make a decision that you don't want to include in the fire PRA

model.

QUESTION

On the electrical drawings, it says that MOV-7 and -8 are racked out. Are these included

anyway because they are high consequence?

RESPONSE

Yes. The circuit analysts will consider the potential for a three-phase smart short, even if the

valve is racked out. Include it in the analysis, but you will ultimately determine that the

probability of this short is very low.

QUESTION

Could you clarify the meaning of the "normal position" column in table I in the fire PRA

equipment list?

RESPONSE

"Normal position" here refers to the position of the component prior to failure. For example,

HPI-A in the first line of the table initiates from standby, turning off in the event of an electrical

failure and on as the desired position.

Session 2: Fire-Induced Risk Model

Development

QUESTION

Slide 5: How do you define what LERF means in the context of a PRA?

RESPONSE

It's the same definition that you use in the PRA right now. There's no change in the definition of

these risk metrics. What will change is the things that give you core damage and LERF, like the

fire damage and impacts, the fire initiators in your model. The intent of 6850 was to make it

generic enough that it would work with any type of software package, rather than being

designed to focus on any one methodology or software package. The advantage of this

approach is that it is flexible, but the downside is that it might be too high-level or generic in

some cases. For this reason, we present here a guide on how to get there, but once you get it,

you try to incorporate these fire risk models into your plant. You'll have to tailor it into your own

packages. Nothing new here, just new fire-induced initiators and failure modes. The human

actions will have to be reassessed; in particular, the post-initiator actions that must be taken

A-2
after a plant trip. You must account for the fact that there are higher stress conditions and more

confusion during a fire. In addition, there might be some new operator actions in the fire

emergency procedures that must be assessed for their impacts on core damage. In appendix

R, the reason you shut off equipment (thereby risking damage to it) is so that you minimize the

risk of spurious equipment functioning. For this reason, you pretty much isolate the off-site

power and try to guarantee that you can get power from the diesels.

QUESTION

Slide 7: How do you know which initiating events correspond to which system failure right

away? Do you have to look at all possible scenarios and screen from there, or can you

immediately only look at certain subsets?

RESPONSE

If it doesn't affect another of your mitigation systems and it's just a reactor trip, most of the time

it's going to screen. The fire actually has to be affecting something in your mitigation system for

you to get that impact. Take, for example, a seal LOCA in a PWR. You have to lose CCW,

thermal barrier cooling, and seal injection. You might think that the fire causes an RCP seal

LOCA. When trying to figure out how to do this, you need to remember that this is all one big

AND gate, so you're always multiplying by some initiator. What initiator can I use to capture the

effect? If I map to a reactor trip and my cables are hooked up to a CCW and a charging pump, I

could conceivably capture that effect through cable propagation. I don't need to make a special

seal LOCA, initiator or CCWs because I would have captured that effect through propagation

through my model. In this example, as in the real situation, you don't want to map every fire

zone to every initiator, you want instead to find a set that seem to match the plant response. If

you have a fire that causes a LOCA, you want to go under the LOCA tree. If you have

something that effectively causes a steam line break, you will want to go under the steam line

break tree. This is where you will need to exercise some judgment on where to map the fire

initiating events, and sometimes you will have to look at the cable impacts, if you have them.

I want to mention that this task 5 is another iteration one. First shot through, you may or may

not have the cables. Although you want them, if you don't have them, you must assume failure

of that system. So, by definition, if you don't have the cables, you take a hit on that system.

Once you do have the cables, then you can back off that system and the cable impacts will

come through the system rather quickly.

QUESTION

I'm still confused: can the core damage frequencies inform the event trees?

RESPONSE

Yes. Here is an example: you come in with some transient initiating event in your event tree.

(The event tree is a series of events used to model the steps necessary to core damage.) Then

you ask yourself about the RCS integrity. Each one of these branches has a fault tree under it,

a number of different ways to arrive at an RCS integrity failure. You're looking for those fire

events that can make PORVs go open, cause seal LOCAs, cause vent valves to open, anything

that can give you a hole in your system that's equivalent to a small LOCA. The next step is to

figure out how to mitigate a break in your coolant system, and this is a simplified model. You

can inject water into the system for injection, in order to mitigate the loss of coolant. Let's go

A-3
back to the basics: up means it did occur, down means it did not. When you go down on this

branch here, it means you have some kind of LOCA on your hands. Success is always up,

failure is always down. You then ask yourself, "what do I need to mitigate this LOCA?". But

then you put water into the reactor coolant system, and it's coming out of the PORV, but it's now

just draining to the bottom of containment. The water in the tank won't last, so this is where the

recirculation pumps come into the picture. You first ask yourself whether or not you have water

to put in. In this simplistic model, if you don't have the water, that's core damage.

Another option is to go up on the coolant injection branch. Yes, you did have injection. But,

remember now, you're running on water out of the tank. Now you have to realign the suction

from the sump to get suction from the pump so that you can go into recirculation and maintain

core cooling. If you fail recirculation, that's core damage. If you make it through recirculation,

you're done.

QUESTION

So where does the CDF come from?

RESPONSE

CDF is the sum of the probabilities of every instance here where you do obtain core damage.

You can fail injection and have core damage, or you can fail recirculation and also have core

damage.

QUESTION

Slide 9: When you say you're going to map it to the different trees, I guess that there are

enough inputs on the different trees that are different, because of the HRAs and other things,

that wouldn't you really just copy that event tree structure and have a separate initiator and use

that same structure?

RESPONSE

That's one way of doing it. You could have an event tree for each initiating event, but what

we've found in these initiating events is that the structure is rather similar. The plant responds

in pretty much the same way to different events, it's just about what's being affected

underneath. If you have a hole in your system, you will always need injection.

QUESTION

We've got deterministic fire houses, and we've got actions in some zones that require the

operator to just start shutting stuff down. Operators hate that. That's where we are, and

hopefully 805 will get us out of that. The procedures will take us to where we want to be. When

you're building this pump, and you're building in access to the operator,...

RESPONSE

You can either put it in up-front, or you can build a model without those actions, unless you

know it will degrade something, and then as you run through the model... Take, for example, an

MOV. The procedure might say, "the MOV spuriously opened, send a guy there locally and go

close it." Initially, I wouldn't put that operator action in the model, I would just let the fire fail that

A-4
MOV. Then what I'll do is add a recovery on that system. I can add a solution and then

simulate the operator recovering from that failure, with some number, and then I'll see what the

impact of that is. It's an iterative process.

Now, the ones where they actually turn off equipment, like a HPSI pump, can be initially

mapped as a failure of that particular pump. I'll then see if there are any associated recoveries.

You have to capture the detrimental effect, and in most cases, you can capture them within the

framework of the PRA model. There are some cases, however, that are so severe that the

structure of the model doesn't work, and you have to build a special response model for the

FEPs, which hopefully you won't have to do. An example I gave is an SISBO, or a self-induced

station blackout. When they perceive a fire, the first thing they'll do is cut off the off-site power

and bring the power from the diesels. All of the unnecessary equipment, except the safe

shutdown equipment, will be turned off. It's pretty hard to actually determine this structure, and

it will be largely determined by HRA. The necessary human actions are what will bring you into

a stable condition. If you have a LOCA on your hands, which hopefully you won't, it might not

be the best thing to do. If the off-site power is there, you may wish to consider it. The point

here is that there are some situations where the PRA model may not be adequate. You have to

see what your FEPs are doing and see whether you can operate within the framework of the

existing PRA.

QUESTION

Slide 9: Given a known cause of an event with multiple causes, how do you model the particular

initiating event in the fault tree?

RESPONSE

This is your internal event initiator model in here. You have determined that fire in this

compartment will give you a loss of all water pumps. In order to model this effect, you have to

look at what initiating events are getting you there. One way is to find the loss of water initiating

event in your model, embedded somewhere in your fault trees, then make an OR gate, and slap

this fire initiating model right underneath of it. So now when you solve that model, a fire in that

compartment will have the same impact as a loss of all water.

QUESTION

I'm modeling, over here, the fire initiator. Is this meant to be just one of a number of acceptable

ways to do this? In terms of adding an initiator, I'm looking at the raw water initiator and adding

under there. One could also go in and look at the unavailability term, and calculate a conditional

core damage probability, and multiply by the fire probability term. There are different ways to

approach it.

RESPONSE

Yes. If there were more than one fire initiator, they would all be under the OR gate. There are

different ways to approach it. This method here gives you a CDF. If I multiplied these two

frequencies outside the model, all I want is a CCDP. So instead of putting an initiator under

there, I would put something else, like when I calculate a core damage probability for this

compartment, I would put some logic element there that would fail that gate. If you fail that gate,

it will propagate all up through the front line systems, at which point you will get a core damage

probability without considering the probability of the initiator. The initiator has been set to 1. But

A-5
then you've put some kind of flag in there to fail that gate to simulate the loss of raw water. So,

like I said, there are a couple of ways to do it.

QUESTION

Does the way you approach this affect what capability or category you would be? Is there a

preferred method with the NRC for this? (CCDP or CDF)

RESPONSE

Not that I know of, but people have done it both ways. Any way that you can show a core

damage probability. The example shows here that if you wanted to integrate these fire initiators

into your model for an A4 application, this might be the way you want to go. And when you

solve this model, you will get cutsets, or combinations of fire initiators. You will actually see the

fire initiator in the solution. Whereas, if you were to do CCDP, you're setting all your initiators to

1.

QUESTION

Slide 14: Shouldn't that operator action be a 1? Otherwise, you're taking credit for a failure.

RESPONSE

Yes, you're right. Usually it is 1. In fact, a lot of times, I don't even add the AND gate, I just

consider it passed. I'm just showing how, if you wanted to model it explicitly, and someone

could give you a model for that. But usually it's pretty high, because you're counting on the

operator to follow standard procedures. It's kind of a weird thing in this fire thing where, if you

follow procedures, you degrade your equipment.

Session 3: Qualitative/Quantitative Screening

QUESTION

What is the screening process for internal HRA events?

RESPONSE

We follow a similar screening process for internal events and HRA. Generally, we'll put in a

very high failure probability for all the human error events, something like .3, and then we just let

it propagate through the initial evaluation model. The combinations that come up through the

cut sets, you look at them and you have to evaluate those combinations with regard to the

possibility of dependencies between them. Sometimes they may even be related to the same

procedural steps: you might have failure to initiate one system and failure to initiate another

system, and if they're the same functions its addresses in the same step of the procedure, its

really the same human error. That's why, when you do this initially, you put a high screening

value in because potentially a lot of these could be deep, highly dependent almost the same

event. If you start to put in 1OA-2, they disappear and you miss out on important insights, so

A-6
that's why we use high screening probabilities and I think its sort of the similar concept here is

you need to be very careful in eliminating human error events.

QUESTION

In the KFW example, you're saying that you're applying 7-3 to all of these actions in the control

room. Why or what's the thought process on doing that for recovery actions that don't have any

interactions with KFW?

RESPONSE

Simply because there exist other performance-shaping factors besides what directly relates to

the systems, the crew will be distracted by other things going on. They have a lot of workload to

address, so its more then just addressing the aux feedwater system itself, you have to address

all the actions that take place.

QUESTION

What do you mean by compartment, is that the same as a room?

RESPONSE

In the fire modeling section of 6850 they define, they have criteria for defining what a fire

compartment is. It could correspond to a fire area, it may not. In some people's plants the fire

areas are very large and it's just too much to handle in just one shot, so they actually divide

these things into smaller divisions, which they call compartments. These are the subdivisions of

the plant that were determined in the other task from a fire standpoint, it's a fire criteria to define

what the boundaries are, what the criteria is. So from this stand point it's an input to us we say

ok that's a compartment, we believe you.

QUESTION

How do you screen HRA events before you know their interaction with hardware?

RESPONSE

When we say screening, I'm getting it below a threshold, but later on I put the HRA in and a lot

of those things pop up so I'm not really getting rid of it I'm doing it on step at a time. I'm running

the model with normal HRAs to get a feel for what's getting me directly by the fire independent

of the HRA. From experience you put them all together, and you don't know if it's the HRA or

the hardware. When we first wrote it we said put it all together and see what happens, but in

practice, you can't see the forest from the trees. You ask yourself, "is it the HRA that's getting

me or is it the hardware?" So what I do in the first step is forget about the HRA for a while and

take a look at the hardware impacts because of the fire and then start bringing in the HRA to

see what HRAs we have to work on.

A-7
Session 6: Fire Risk Quantification

QUESTION

When we started using the PRA tool, it was really to try to come up with the best estimate of

risk, but in the last two or three years, especially with the development integrating fire into it, that

approach has really changed, and we're going to make conservative assumptions in applied

PRA techniques. Why the change from what we had for several years?

RESPONSE

Part of the reason is when we do these applications and these STPs, people ask us the impact

of fire or seismic events. So now, from a risk perspective, and while realizing that maybe the

technology is not as evolved as what we have in the internal events in the other applications,

you have to really the limitations of the model.

QUESTION

How much data do we have to support the fire ignition frequency numbers? Where did we get

that data from?

RESPONSE

Well as part of that effort, the fire database was generated, and it included information that

came from multiple sources, including some from reportable incidents which the NRC keeps

record of. And some instances aren't necessarily recordable, so all the information that could

be obtained was obtained with regards to fire incidents and put into a database. And that data

essentially was parsed, used to look at the frequencies of fires related to specific components

for example the frequency of fires related to transient combustibles. This is covered in the fire

task.

QUESTION

Should steps 3-6, the analysis of spurious events, be included in steps 1&2?

RESPONSE

Some are included in the fire safe shutdown analysis, but not all of them. The intent here was

to work your way down from a high level. The first step is kind of looking at your PRA at a high

level and looking at the initiators and sequencers in your model, asking yourself if they can be

evolved in a fire. From a fire perspective, you find out at a high level which trees can be

screened out with proper justification, and you're stuck with maybe one or two event trees that

you really have to deal with. After understanding and screening the internal events model, you

wonder how it can be impacted with fire and its scope. Now you're looking at it from the

standpoint of your appendix R and safe shutdown analysis, and there'll be some commonality,

but the reason why we do this is because supposedly appendix R has the components for the

system that have already been analyzed for circuits. You have cable routing, and you have

circuit analysis, so you want to take as much credit as you can for that appendix R when you're

able to. So we compare, for example, a safety injection pump and you look like you got a match

A-8
there, and some of the circuits match. However, if you look in your PRA model, the automatic

actuations model, the appendix R manual operations presume that you don't have the circuitry

for automatic actuations. When we do the comparison, when we get down to the component

level, you can reflect and go back to your high level thinking and get an idea whether you won't

be comparing oranges and apples; you'll know the differences. So we start from a high level, of

sequences and event trees, and then compare appendix R and at some point you're going to

end up at the component level. You'll have this big database with components. Some

components are in appendix R, and some components are in PRA, and by this point you should

know why.

QUESTION

What percentage do you add when you do steps 3-6?

RESPONSE

There are some plants that I would say are light on their internal events, so by definition, they

are missing a bunch of variables. For the ones that have pretty robust PRA models, it's mostly

adding more failure models than components. It's hard to give a number because it varies.

A-9

Appendix B:

Sample Exercise Solutions (Module 1, Task 2)

Step 1: Identify Internal Events PRA Sequences to be Included (and those to be excluded) in the Fire PRA Model

INITIATING EVENTS IN THE PRA MODEL

Initiator Average Description Accident Address in Fire Comments

Frequency Sequence PRA Model?

(per yr) or Event (Y or N)

Tree Model

%T1 7.23E-01 Reactor Trip Transient

Event Tree

%T2 9.33E-02 Loss of Condenser Transient

Vacuum Event Tree

%T3 4.13E-01 Turbine trip Transient

Event Tree

%T4 3.73E-02 Loss of Main Feedwater Transient

Event Tree

%T5P 4.25E-02 Loss of Offsite Power Transient

(Plant-Centered) Event Tree

%T5C 1.02E-02 Loss of Off-Site Power Transient N Grid-related loss of offsite power due to

(Grid-Related) Event Tree internal fire is highly unlikely

%T5D 6.26E-03 Loss of Off-Site Power Transient N Weather~induced loss of offsite power due to

(Weather-Induced) Event Tree internal fire is highly unlikely

%T6 7.35E-03 Steamline/Feed line Main Y Fire-induced pipe break is not likely.

Break Upstream of Main Steamline However, stuck-open SG PORV/atmospheric

Steam Isolation valves Break Event relief valve could result in uncontrolled

B-1
INITIATING EVENTS IN THE PRA MODEL

Initiator Average Description Accident Address in Fire Comments

Frequency Sequence PRA Model?

(per yr) or Event (Y or N).

Tree Model

or Downstream of Tree cooldown.

Feedwater Isolation

Valves (Includes StuckOpen Secondary relief

valves)

%T7 5.44E-03 Steamline Break Main Y Fire-induced pipe break is not likely.

Downstream of Main Steamline However, condenser steam dump valve(s)

Steam isolation valves Break Event could result in uncontrolled cooldown.

(Includes Stuck-Open

Secondary relief valves)

%T8 2.94E-04 Loss of 4160 V Bus 1 Transient

Event Tree

%T9 2.94E-04 Loss of 416 OVBusA Transient

Event Tree

%T10 2.94E-04 Loss of 4160 V Bus B Transient

Event Tree

%T11 2.94E-04 Loss of 4160 V Bus 2 Transient

Event Tree

%T12 3.OOE-03 Loss of 125 VDC Bus A Transient

Event Tree

%T13 3.OOE-03 Loss of 125 VDC Bus B Transient Y

Event Tree

%T1 5 Fault Tree Loss of CCW System Transient

Model Event Tree

%T15-INIT

%T16 Fault Tree Loss of Service Water Transient

B-2
INITIATING EVENTS IN THE PRA MODEL

Initiator Average Description Accident Address in Fire Comments

Frequency Sequence PRA Model?

(per yr) or Event (Y or N)

Tree Model

Model System Event Tree

%T16-INIT

Y

%T1 7 Fault Tree Loss of Instrument Air Transient

Model Event Tree

%T17-INIT

%T21 3.41 E-02 Closure of MSIV (1 SG Transient

Loop) Event Tree

%T22 1.24E-02 Closure of both MSIVs Transient

Event Tree

%T23 1.78E-01 Partial Load Rejection Transient

Event Tree

%T24 5.79E-02 Spurious Steam Gen. Transient

Isolation Signal Event Tree

%T25 7.23E-02 Reactor Trip With PORV Transient

Opening/Demand Event Tree

%T26 Fault Tree Loss of Power from*120 Transient Y

Model VAC Buses A & B Event Tree

%T26-1NIT

See comment Fire-induced pipe break or passive RCP seal break L Small CA (peal Small LrA failure is unlikely. However, portion of small

breaks and RCP seal Event Tree LOCA event tree likely needed to treat

LOCA) transient-induced small LOCA from Transient

Event Tree such as the pressurizer PORV

demanded and stuck-open or RCP seal

LOCA via loss of RCP seal cooling.

%M 9.60E-06 Medium LOCA (pipe Medium See comment Fire-induced pipe break is unlikely. For our simplified plant, any transient-induced LOCA

B-3
INITIATING EVENTS IN THE PRA MODEL

Initiator Average Description Accident Address in Fire Comments

Frequency Sequence PRA Model?

(per yr) or Event (Y or N)

Tree Model

breaks) LOCA Event is considered small. If in another plant, a

Tree transient-induced medium LOCA could occur

based on one or more events (e.g., a single

but large size PORV or multiple PORVs

spuriously opening), then a portion of the

medium LOCA event tree may be needed.

See comment Fire-induced pipe break is unlikely. For our %Lrgeaks) LA (irge Lsimplified plant, any transient-induced LOCA

breaks) Event Tree is considered small. If in another plant, a

transient-induced large LOCA could occur

based on one or more events (e.g., a single

but large size PORV or multiple PORVs

spuriously opening), then a portion of the

large LOCA event tree may be needed.

%R N Fire-induced rupture of SG tubes is highly 7.93E-03 Steam Generator Tube SGTR Eventuniey

Rupture Tree unlikely.

%12 2.OOOE-07 Interfacing Systems ISLRHR Y Spurious opening of both RHR suction valves

LOCA at RCS/RHR Sequence could result in ISLOCA. Since power to the

Interfat (2 in (sinleqvene inboard valve is racked out, a 3-phase hot Interface (2 MOVs in (single event short would be required to open that valve.

series) model) Consideration of 3-phase hot shorts is

required for high-consequence lines.

%13 Fault Tree Interfacing Systems ISLCCW N Fire-induced ISLOCA highly unlikely via a

Model LOCA at RCS/CCW Sequence passive rupture of the RCP Cooler

13QINIT interface (Reactor

Coolant Pump Cooler

rupture)

%VR N Fire-induced rupture of the reactor vessel is 2.70E-07 Reactor Vessel Rupture Single Eventhilyuiky

in Master highly unlikely

Fault Tree

B-4
Accident Sequence

or Event Tree Model

TRA

ACCIDENT SEQUENCE OR EVENT-TREE MODELS IN THE PRA

Description Additional Details Address in Comments

Fire PRA

Model?

(Y or N)

Transient Includes transient-induced Y

LOCAs such as stuck-open

PORV and RCP seal LOCA

SLOCA Small LOCA Pipe breaks & RCP seal LOCA See comment

MLOCA Medium LOCA Pipe breaks See comment

Fire-induced pipe break or passive

RCP seal failure is unlikely. However,

portion of small LOCA event tree

likely needed to treat transientinduced small LOCA from Transient

Event Tree such as the pressurizer

PORV demanded and stuck-open or

RCP seal LOCA via loss of RCP seal

cooling.

Fire-induced pipe break is unlikely.

For our simplified plant, any transientinduced LOCA is considered small. If

in another plant, a transient-induced

medium LOCA could occur based on

one or more events (e.g., a single but

large size PORV or multiple PORVs

spuriously opening), then a portion of

the medium LOCA event tree may be

needed.

Fire-induced pipe break is unlikely.

For our simplified plant, any transientinduced LOCA is considered small. If

in another plant, a transient-induced

large LOCA could occur based on

one or more events (e.g., a single but

large size PORV or multiple PORVs

spuriously opening), then a portion of

the large LOCA event tree may be

needed.

Fire-induced failure of the reactor

LLOCA Large LOCA Pipe breaks See comment

ATWS Anticipated Transients Reactor Protection System fails N

B-5
SGTR

MSLB

ISLCCW

ISLRHR

Without Scram

Steam Generator

Tube Rupture

Main Steamline Break

Interfacing Systems

LOCA at RCS/CCW

interface

Interfacing Systems

LOCA at RCS/RHR

Interface

Spurious Safety

Injection with HPI

Feedwater ramp-up or

AFW spurious start

RWST drain down

event

safe on loss of power. Trip circuits

are highly redundant and

confirmed to be physically

separated.

Includes spurious opening of

secondary relief valves.

Rupture of Reactor Coolant Pump

Cooler

Fire-induced opening of RHR

suction valves

Requires multiple spurious valve

openings and possibly 2°d pump start

May require multiple spurious

events (need to check)

Requires combination of either or

both MOV-5,6 with either or both

MOV-3,4 spuriously opening.

N

Y

N

Y

New

New

See comment

See comment

See comment

protection system is highly unlikely.

Fire-induced rupture of SG tubes is

highly unlikely.

Fire-induced pipe break is not likely.

However, stuck-open SG

PORV/atmospheric relief valve or

condenser steam dump valve(s) could

result in uncontrolled cooldown.

Requires passive failure of RCP

cooler - not likely coincident with fire.

Spurious opening of both RHR

suction valves could result in ISLOCA

Need to check on potential to cause

such an event'and the number of

spurious events required. Likely to

cause reactor trip (manual or

automatic on high pressure) if is not

or cannot be terminated by operator.

Could cause subsequent LOCA.

Need to check on potential to cause

such an event and the number of

spurious. events required. Likely to

cause reactor trip (manual or

automatic such as on steam/feed

mismatch) if is not or cannot be

terminated by operator. Could cause

damage to AFW-B pump.

Need to check on potential to cause

such an event and the number of

spurious events required. Likely to

cause procedure-driven manual reactor

trip due to loss of initial safety injection

water supply / LCO condition for HPI.

New

B-6
Step 2: Review of the Internal Events PRA Against the Fire Safe Shutdown Analysis

TABLE 1: SYSTEMS IN PRA MODEL

System Description Additional Details Address in

Fire PRA

Model?

Comments

(Y or N)

RCS

CVCS

HPI

Reactor Coolant

System

Chemical and

Volume Control

System

High Pressure

Injection System

Residual Heat

Removal System

Auxiliary Feedwater

System

Main Feed Water

PORV for pressure relief and feed & bleed.

Stuck-open PORV causes small LOCA.

Normal charging and letdown functions are

not modeled. However, components required

to isolate charging and letdown are modeled

for HPI mode.

The charging pumps in the CVCS also

function as safety injection pumps.

Y Only includes isolation of letdown and

charging for operation in HPI mode

Y

Y

RHR

AFW

MFW

Shutdown cooling is not modeled

Only Trains A and B are modeled

N Will be the same as Internal events

model

Y

Would take considerable effort to get cables

involved and their locations

Stuck-open secondary relief valves could

N Although credited in the PRA model,

the cost of cable routing out weighs

the risk benefit.

MS Main Steam System Y

B-7
System Description

TABLE 1: SYSTEMS IN PRA MODEL

Additional Details Address in

Fire PRA

Model?

(Y or N)

Comments

CS

CF

Cl

ESFAS

CCW

SW

AC

DG

DC

cause equivalent of mainsteam line break.

Containment Spray Required for recirculation during LOCA

Containment Fan Required for recirculation during LOCA

Coolers

Containment Isolation Modeled in LERF

Emergency

Safeguards Actuation

System

Component Cooling

Water System

Service Water

System

AC Power (all voltage To extent power is needed to support

levels) equipment in the PRA.

Emergency Diesel

Generators

DC Power To extent power is needed to support

Y

Y

Y

Y

Y

Y

Y To extent power is needed to support

equipment in the PRA.

Y

Y To extent power is needed to support

B-8
TABLE 1: SYSTEMS IN PRA MODEL

System Description Additional Details Address in

Fire PRA

Model?

Comments

(Y or N)

equipment in the PRA. equipment in the PRA.

IA Instrument Air

HVAC-HPI HVAC in HPI Pump

Room

Required for PORV and other valves. Backup

nitrogen is provided for PORV and is what is

credited.

HVAC is required during 24-hr PRA mission

Y

Y

B-9
TABLE 2: SYSTEMS IN APPENDIX R

System Description Additional Details Address in

Fire PRA

Model?

Comments

(Y or N)

RCS Reactor Coolant System

Chemical Volume and

Control System

PORV to prevent spurious opening

and consequential small LOCA.

Normal charging and letdown

functions are credited.

CVCS

RHR

AFW

Residual heat Removal

System

Auxiliary Feedwater

System

Main Steam System

Shutdown cooling is credited

Trains A&C are credited

Secondary relief valves and MSIVs

are included to prevent spurious

opening causing uncontrolled

secondary depressurization.

Y But need in fire PRA for more than just

ensuring closure for Appendix R.

N Normal charging and letdown functions are

not credited in PRA. However isolation of

letdown and normal charging is to be

modeled for HPI mode.

N Shutdown cooling not credited in PRA

Y Especially since will not credit MFW,

crediting all trains of AFW in the Fire PRA

may be important for 'realistic' risk

estimates.

MS Y Review from PRA perspective

CCW Component Cooling

Water System

Y

B-10
TABLE 2: SYSTEMS IN APPENDIX R

System Description Additional Details Address in

Fire PRA

Model?

Comments

(Y or N)

SW

AC

Service Water System

AC Power (all voltage

levels)

Y

But certain buses not credited

(especially non-safety) if loads not

otherwise required for safe

shutdown

Y Need to ensure all buses are included that

are needed to support equipment to be

modeled in the Fire PRA.

DG

DC

Emergency Diesel

Generators

Y

DC Power But certain buses not credited

(especially non-safety) if loads not

otherwise required for safe

shutdown

Required for PORV and other

valves. Backup nitrogen is provided

for PORV.

Y Need to ensure all buses are included that

are needed to support equipment to be

modeled in the Fire PRA.

IA Instrument Air Y

RCS Instruments RCS pressure,

temperature, nuclear

instrumentation, etc

Steam Generator level,

Required for safe shutdown

monitoring.

Required for safe shutdown

See. Include in model if instruments impact

comment equipment or human failure events in the

PRA. See Step 5.

Secondary See Include in model if secondary instruments

B-11
TABLE 2: SYSTEMS IN APPENDIX R

System Description Additional Details Address in

Fire PRA

Model?

Comments

Instruments

HVAC-HPI

HVAC-AFW

Streamline pressure, etc. monitoring.

HVAC in HPI Pump

Room

HVAC in AFW Pump

Room

HVAC is required during 72-hr

Appendix R mission

HVAC is required during 72-hr

Appendix R mission

(Y or N)

comment impact equipment or human failure events

in the PRA. See Step 5.

Y Required for PRA during 24-hr mission. 72-

hr mission is for cold shutdown and not part

of PRA.

N 72-hr mission is for cold shutdown and not

part of PRA. Not required for PRA mission.

B-12
TABLE 3: LIST OF BASIC EVENTS IN MODEL

Basic Event Description

%12 Interfacing Systems LOCA at RCS/RHR Interface (2 MOVs in series)

%13 Interfacing Systems LOCA at RCS/CCW interface (Reactor Coolant Pump Cooler rupture)

%T15 LOSS OF COMPONENT COOLING WATER (CCW)

%T23 PARTIAL LOAD REJECTION

%T25 REACTOR TRIP WITH PORV OPENING

%T3 TURBINE TRIP

%T4 LOSS OF MAIN FEEDWATER

%T1 REACTOR TRIP

AFWA-FTR AFWA fails to run

AFWA-FTS AFWA fails to start

AFWB-FTR AFWB fails to run

AFWB-FTS AFWB fails to start

AOV-1_FTC PORV AOV-1 fails to CLOSE

AOV-1_FTO PORV AOV-1 fails to open

AOV-3_FTC AOV-3 FAILS TO CLOSE

EPS-120VBUSAF 120V BUS A FAULT

B-13
TABLE 3: LIST OF BASIC EVENTS IN MODEL

Basic Event Description

EPS-120VBUSAINVF FAILURE OF 120V BUS A INVERTER

EPS-125VDCBUSAF FAULT ON 125V DC BUS A

EPS-125VDCBUSBF FAULT ON 125V DC BUS B

EPS-125VDCPNLAF FAULT ON 125V DC PANEL A

EPS-125VDCPNLBF FAULT ON 125V DC PANEL B

EPS-480VLCAF 480V LOAD CENTER A FAULT

EPS-480VLCAXTF 480V LOAD CENTER A TRANSFORMER FAILS

EPS-480VLCBF 480V LOAD CENTER B FAULT

EPS-480VLCBXTF 480V LOAD CENTER B TRANSFORMER FAILS

EPS-480VMCCAl F 480V MCC Al FAULT

EPS-480VMCCBI F 480V MCC B1 FAULT

EPS-4VBUSAF 4KV BUS A FAULT

EPS-4VBUSBF 4KV BUS B FAULT

EPS-BATA FAILURE OF BATTERY A

EPS-BATB FAILURE OF BATTERY B

EPS-BCAF FAILURE OF BATTERY CHARGER A

EPS-BCBF FAILURE OF BATTERY CHARGER B

B-14
TABLE 3: LIST OF BASIC EVENTS IN MODEL

Basic Event Description

EPS-DGAF FAILURE OF DIESEL GENERATOR A

EPS-DGBF FAILURE OF DIESEL GENERATOR B

HPIAFTR HPIA fails to run

HPIAFTS HPIA fails to start

HPIBFTR HPIB fails to run

HPIBFTS HPIB fails to start

MFWFAIL MAIN FEEDWATER SYSTEM FAILURE AFTER REACTOR TRIP

MOV-10_FTO MOV-10 fails to open

MOV-1 1_FTO MOV-1 1 fails to open

MOV-14_FTO MOV-14 FAILS TO OPEN

MOV-15_FTO MOV-15 FAILS TO OPEN

MOV-1_FTO MOV-1 FAILS TO OPEN

MOV-2_FTC MOV-2 fails to close

MOV-3_FTO MOV-3 fails to open

MOV-4_FTO MOV-4 fails to open

MOV-5_FTC MOV-5 fails to close

MOV-5_FTO MOV-5 fails to open

B-1 5
TABLE 3: LIST OF BASIC EVENTS IN MODEL

Basic Event Description

MOV-6_FTC MOV-6 fails to close

MOV-6_FTO MOV-6 fails to open

MOV-9_FTO MOV-9 FAILS TO OPEN

OPER-1 Operator fails to switch over to recirculation

OPER-4 Operator fails to establish feed and bleed cooling

OPER-7 OPERATOR FAILS TO TRIP REACTOR COOLANT PUMP

RCPSEAL RCP SEAL LOCA GIVEN LOSS OF CCW AND SUCCESSFUL RCP TRIP

SUTF FAILURE OF START-UP TRANSFORMER (SUT)

UATF FAILURE OF UNIT AUXILIARY TRANSFORMER (UAT)

B-16
Continuation of Step2 and Including Steps 4 thru 6:

TABLE 4: DISPOSTION OF COMPONENTS IN PRA AND APPENDIX R (USE TABLES 1-3 AND FIGURES 1-3)

Equipment ID

HPI-A

HPI-B

Equipment Description

High pressure safety

injection pump A

High pressure safety

injection pump B

Power

Supply

In PRA

Model? (Y or

N)

In Appendix

R? (Y or N)

Add to Fire PRA

Equipment List? (Y or N)

4.16kV Bus A

4.16kV Bus B

Y

Y

N

Y

Y

Y

Comments

Y

Y

N

RHR-B RHR pump 4.16kV Bus B

Not modeling shutdown

cooling and spurious

operation benign to credited

functions (See Step 4)

To ensure long-term PORV

operation

COMP-1

AFW-A

AFW-B

AFW-C

AOV-1

Instrument air

compressor

Motor driven AFW pump

A

Steam driven AFW Pump

B

AFW Pump C

Pressure operated relief

480 V LC 1

4.16kV Bus A

N/A

4.16 kV Bus 2

120VAC Bus

Y

Y

Y

N

Y

Y

Y

N

Y

Y

Y

Y

Y

Y

Credited in Appendix R. Will

be credited for the fire PRA

Y (only to Need for both ensuring

B-17
TABLE 4: DISPOSTION OF COMPONENTS IN PRA AND APPENDIX R (USE TABLES 1-3 AND FIGURES 1-3)

Equipment ID

(Sov-1)

Equipment Description

valve

Letdown isolation valve

Charging pump injection

valve

Power

Supply

In PRA

Model? (Y or

N)

In Appendix

R? (Y or N)

Add to Fire PRA

Equipment List? (Y or N)

A ensure

remains

closed)

AOV-2

(SOV-2)

AOV-3

(SOV-3)

MOV-1

MOV-2

Comments

125 VDC Bus

B

125 VDC Bus

B

480V MCC Al

480V MCC B1

N

Y

Y (for normal

letdown)

Y (for normal

charging)

Y

Y

closure (does not spuriously

open) and to open when

needed for feed and bleed.

Especially for isolation when

needed. See Step 4.

Especially for isolation when

needed.

HPI valve Y

Y

N Y

Y

VCT isolation valve

Y (for normal

suction to

charging)

MOV-3

MOV-4

MOV-5

MOV-6

Cont. sump recirc valve

Cont. sump recirc valve

RWST isolation valve

RWST isolation valve

480V MCC Al

480V MCC B1

480V MCC Al

480V MCC B1

Y

Y

Y

Y

N

N

N

N

Y

Y

Y

Y

Especially for isolation when

needed.

Also see Step 4 for spurious

operation concerns

Also see Step 4 for spurious

operation concerns

Need to ensure both open

and close when desirable

Need to ensure both open

B-18
TABLE 4: DISPOSTION OF COMPONENTS IN PRA AND APPENDIX R (USE TABLES 1-3 AND FIGURES 1-3)

Equipment ID Equipment Description

Power

Supply

In PRA

Model? (Y or

N)

In Appendix

R? (Y or N)

Add to Fire PRA

Equipment List? (Y or N)

Comments

Y (for

ISLOCA)

MOV-7

RHR inboard suction

valve

480V MCC Al

Y (for

shutdown

cooling)

Y (for

shutdown

cooling)

Y

Y

and close when desirable

PRA will not address

shutdown cooling but need

to address for fire-induced

ISLOCA and possible high

consequence event (see

Step 6)

PRA will not address

shutdown cooling but need

to address for fire-induced

ISLOCA and possible high

consequence event (see

Step 6)

Y (for

ISLOCA)

MOV-8

RHR outboard suction

valve

480V MCC B1

MOV-9

MOV-10

MOV-11

MOV-13

MOV-14

HPI valve

AFW discharge valve

AFW discharge valve

PORV block valve

AFW turbine steam line

480V MCC B1

480V MCC Al

125 VDC Bus

B

480V MCC Al

125 VDC Bus

Y

Y

Y

N

Y

N

Y

N

Y

N

Y

Y

Y

Y

Y

B-19
TABLE 4: DISPOSTION OF COMPONENTS IN PRA AND APPENDIX R (USE TABLES 1-3 AND FIGURES 1-3)

Equipment ID Equipment Description

Power

Supply

In PRA

Model? (Y or

N)

In Appendix

R? (Y or N)

Add to Fire PRA

Equipment List? (Y or N)

Comments

isolation valve B

MOV-15

AFW steam inlet throttle

valve

AFW test line isolation

valve

125 VDC Bus

B

480V MCC Al

Y

N

N

N

Y

N

MOV-16

Potential for significant flow

diversion is small.

See Step 4

Potential for significant flow

diversion is small.

N N N

MOV-17

MOV-18

MOV-19

AFW test line isolation

valve

AFW C Pump Discharge

AFW test line isolation

valve

480V MCC B1

480 V MCC-2

480 V MCC-2

See Step 4

N

N

Y

N

Y

N Potential for significant flow

diversion is small.

See Step 4.

N

V-12 CST isolation valve N/A

Y (for admin

purposes to

ensure open)

N Unlikely to be

closed/plugged coincident

with fire.

LI-1 RWST level

120VAC Bus

A

N Y Y See Step 5.

B-20
Equipment ID

LI-2

LI-3

LI-4

TI-1

PT-1

A-1

SWGR-A

SWGR-B

SWGR-1

TABLE 4: DISPOSTION OF COMPONENTS IN PRA AND APPENDIX R (USE TABLES .1-3 AND FIGURES 1-3)

In PRA In Appendix Add to Fire PRA Power

Equipment Description Model? (Y or R? (Y or N) Equipment List? (Y or N) Supply N

N)

120VAC Bus N Y Y RWST level

B

120VAC Bus N Y Y

Cont. sump level A

120VAC Bus N Y Y

Cont. sump level B

Letdown heat exchanger 120VAC Bus N N Y

outlet temp A

120VAC Bus N N Y

RCS pressure B

120VAC Bus N N Y

AFW motor high temp A

SUT-1 Y Y Y

Train A 4160 V Bus

EDG-A

SUT-1 Y Y Y

Train B 4160 V Bus

EDG-B

UAT-1 N N Y

Non-Safety 4160 V Bus

SUT-1

Comments

See Step 5.

See Step 5.

See Step 5.

See Step 5.

See Step 5.

See Step 5.

B-21
Equipment ID

SWGR-2

SUT-1

EDG-A

EDG-B

LC-1

LC-2

LC-A

LC-B

SST-1

SST-2

TABLE 4: DISPOSTION OF COMPONENTS IN PRA AND APPENDIX R (USE TABLES 1-3 AND FIGURES 1-3)

In PRA In Appendix Add to Fire PRA Power

Equipment Description Model? (Y or R? (Y or N) Equipment List? (Y or N) Supply N

N)

UAT-1 N Y Y

Comments

Non-Safety 4160 V Bus

Startup Transformer

Train A Emergency

Diesel Generator

Train B Emergency

Diesel Generator

Non-Safety 480 V Load

Center

Non-Safety 480 V Load

Center

Train A 480 V Load

Center

Train B 480 V Load

Center

Non-Safety Station

Service Transformer

Non-Safety Station

SUT-1

OSP

PNL-A

PNL-B

SST-1

SST-2

SST-A

SST-B

SWGR-1

SWGR-2

Y

Y

Y

N

Y

Y

N

N

Y

Y

Y

Y

Y

Y

Y

Y

Y

Y

B-22
TABLE 4: DISPOSTION OF COMPONENTS IN PRA AND APPENDIX R (USE TABLES 1-3 AND FIGURES 1-3)

In PRA In Appendix Add to Fire PRA Power

Equipment Description Model? (Y or R? (Y or N) Equipment List? (Y or N) SupplyN)

Comments

Equipment ID

SST-A

SST-B

MCC-1

MCC-2

MCC-Al

MCC-B1

BC-1

BC-A

BC-B

BAT-1

Service Transformer

Train A Station Service

Transformer

Train B Station Service

Transformer

Non-Safety 480 V Motor

Control Center

Non-Safety 480 V Motor

Control Center

Train A 480 V Motor

Control Center

Train B 480 V Motor

Control Center

Non-Safety Swing Battery

Charger

Train A Battery Charger

Train B Battery Charger

Non-Safety Battery

SWGR-A

SWGR-B

LC-1

LC-2

LC-A

LC-B

MCC-1

MCC-2

MCC-A1

MCC-B1

N/A

Y

N

N

Y

Y

N

N

Y

Y

Y

Y

Y

Y

Y

N

N

N

Y

Y

Y

B-23
Equipment ID

BAT-A

BAT-B

DC BUS-1

DC BUS-A

DC BUS-B

INV-A

INV-B

VITAL-A

VITAL-B

PNL-A

PNL-B

TABLE 4: DISPOSTION OF COMPONENTS IN PRA AND APPENDIX R (USE TABLES 1-3 AND FIGURES 1-3)

In PRA In Appendix Add to Fire PRA Power

Equipment Description Model? (Y or R? (Y or N) Equipment List? (Y or N) Supply N

N)

Train A Battery N/A Y Y Y

Train B Battery N/A Y Y Y

BC-1 N N Y

Comments

Non-Safety 125 VDC Bus

Train A 125 VDC Bus

Train B 125 VDC Bus

Train A Inverter

Train B Inverter

Train A 120 VAC Vital

Bus

Train B 120 VAC Vital

Bus

Train A 125 VDC Panel

Train B 125 VDC Panel

BAT-1

BC-A

BAT-A

BC-B

BAT-B

DC BUS-A

DC BUS-B

INV-A

INV-B

DC BUS-A

DC BUS-B

Y Y Y

Y Y Y

Y

N

Y

N

Y

Y

Y

Y

Y

Y

Y

Y

Y

Y

Y

Y

Y

Y

B-24
Step 3: Identify Fire-Induced Initiating Events Based. on Equipment Affected

Will need to examine each fire compartment / analysis unit and determine based on the equipment and cables located there, which

of the initiators (from Step 1 of Task 2) can be caused by a fire in that compartment / analysis unit. If any new initiators are identified,

include in the Fire PRA. The Fire PRA will then include fires mapped to initiating events in the model. Each compartment / analysis

unit should have a disposition with regard to the initiating event(s) that occur as a result of a fire in each location (even if "none").

Hold discussion with instructor.

B-25
Step 4: Identify Equipment with Potential Spurious Actuations that May Challenge the Mitigating Capability to be Credited

Considered

spurious equipment

operations

MOVA-16

MOV-1 7

MOV-19

MOV-3

MOV-4

AOV-2

RHR

Description

AFW test line

isolation valve

AFW test line

isolation valve

AFW test line

isolation valve

Cont. sump

recirc. valve

Cont. sump

recirc. valve

Letdown

isolation valve

RHR pump

PRA System or

Function Possibly

Affected

Failure of AFW-A due

to flow diversion

Failure of AFW-B due

to flow diversion

Failure of AFW-C due

to flow diversion

Failure of high

pressure injection

Failure of high

pressure injection

Failure of high

pressure injection

Spurious operation of

pump - investigate

effect on HPI & RHR

Comments/Disposition

Fire-induced spurious opening

of MOV-1 6 could occur but flow

diversion is not significant. Will

not be modeled.

Fire-induced spurious opening

of MOV-1 7 could occur but flow

diversion is not significant. Will

not be modeled.

Fire-induced spurious opening

of MOV-1 9 could occur but flow

diversion is not significant. Will

not be modeled.

Fire-induced spurious opening

of MOV-3 during HPI injection

mode will fail HPI - need to

model this failure.

Fire-induced spurious opening

of MOV-4 during HPI injection

mode will fail HPI - need to

model this failure.

Loss of CCW with failure to

isolate letdown (such as

spurious open signal) will result

in HPSI pump cavitation

Could fail RHR pump

(assuming no pump suction

protection) but RHR shutdown

cooling not modeled anyway.

B-26
No adverse effect on high

pressure injection. No need to

model.

B-27
Step 5: Identify Additional Mitigating, Instrumentation, and Diagnostic Equipment Important to Human Response

Instrumentation

required to perform

human actions

Description Potentially Affected

Human Failure Event

in PRA model

LI-1 RWST level

OPER-A

Failure to align

recirculation

OPER-1

Failure to align

recirculation

LI-2 RWST level

Comments/Disposition

Instrumentation required to

identify need to switch over to

recirculation. While lots of

redundancy with other Ll's, all

are in close proximity on MCR

board and so one fire could

affect multiple instruments.

Hence will model.

Instrumentation required to

identify need to switch over to

recirculation. While lots of

redundancy with other Ll's, all

are in close proximity on MCR

board and so one fire could

affect multiple instruments.

Hence will model.

Instrumentation required to

identify need to switch over to

recirculation. While lots of

redundancy with other LU's, all

are in close proximity on MCR

board and so one fire could

affect multiple instruments.

Hence will model.

Instrumentation required to

identify need to switch over to

recirculation. While lots of

redundancy with other LU's, all

are in close proximity on MCR

board and so one fire could

affect multiple instruments.

LI-3

LI-4

Cont. sump

level

Cont. sump

level

OPER-1

Failure to align

recirculation

OPER-1

Failure to align

recirculation

B-28
Hence will model.

TI-1

Letdown heat

exchanger

outlet temp

OPER-2

Failure to isolate

letdown on high

temperature due to

CCW loss

PT-1 RCS OPER-4

Failure to align feed

pressure and bleed

Instrumentation required to

identify need to isolate letdown

when CCW is lost

Instrumentation (indicator)

required to identify need for

feed and bleed

Spurious high temperature

alarm causes operator to shut

down AFW pump erroneously

N/A

A-1 AFW motor

high temp

B-29
Step 6: Include Potentially High Consequence Related Equipment

Equipment or combinations

of equipment that can cause

high consequence event

MOV-7 and MOV-8

Description of high

consequence event

Spurious opening of MOV-7 and

MOV-8 results in interfacing

system LOCA - meets high

consequence event definition.

Comments/Disposition

Need to model this combination

failure.

B-30
Step 7: Assemble Fire PRA Equipment List

Table 1: Fire PRA Equipment List Information (For Instructors)

Equipment PRA Event PRA Event Normal Desired Failed Failed Air Equipment ID Equipment Description Type Location Power Supply I Position/ Position/ Electrical Position

Type Identifier Description Status Status Position

HPIA FTS HPIA fails to Standby On Off N/A

HPI-A! High pressure safety Pump Aux Bldg. SWGR-A HP't injection pump A Pup El. 0 Ft SGHPIA FTR HPIA fails to On On Off N/A run

HPIB FTS HPIB fails to Standby On Off N/A

HPI-B High pressure safety Pump Aux Bldg. SWGR-B

P-B injection pump B El. 0 Ft

HPIB FTR HPIB fails to On On Off N/A run

AFW-A Motor driven AFW Pump Turbine Bldg. SWGR-A AFWA-FTS AFWA fails Standby On Off N/A

pump A El. 0 Ft start

AFW-A Motor driven AFW Pump Turbine Bldg. Turbine Bldg. AFWA-FTR AFWA fails to On On Off N/A AFW-A pump A El. 0 Ft El. 0 Ft run

AFW-B Steam driven AFW Pump Turbine Bldg. N/A AFVB-FTS AFWB fails to Standby On N/A N/A

pump B El. 0 Ft start

AFW-B Steam driven AFW

pump B Pump Turbine Bldg.

El. 0 Ft N/A AFWB-FTR AFWB fails to

run On On N/A N/A

AFW-C Motor driven AFW Pump Turbine Bldg. SWGR-2 AFW-C i pump C El. 0 Ft

AFW-C Motor driven AFW Pump Turbine Bldg. SWGR-2 AFW-C pump C El. 0 Ft

RCP Reactor coolant pump Pump Containment SWGR-1

COMP-1 Instrument air Compressor Turbine Bldg. LC-1

compressor El. 0 Ft

COMP-1 Instrument air Compressor Turbine Bldg. LC-1

compressor El. 0 Ft

Standby On Off N/A

Standby On Off N/A

On Off Off N/A

Cycle Cycle Off N/A

Cycle Cycle Off N/A

B-31
Equipment PRA Event PRA Event Normal Desired Failed Failed Air Equipment ID Equipment Description Type Location Power Supply Pd EventiPRr Evsnt Position/ Positionl Electrical Paitin Type Identifier Description Status Status Position

Closed Closed Closed Closed

AOV-1" Power operated relief AOV Containment VITAL-A (SOV-1) valveF AOV-1 FTO PfR AOV-t Closed Open Closed Closed

fails to open

AOV-2~ Letdown isolation AxBldg. D U-

(SOV-2) valve AOV El. 0 Ft DC BUS Open Closed Closed Closed

AOV- ChagingpumpAux ldg.I I AOV-ý3 FAILS

AOV-3' Charging pump Aux Bldg. DC BUS-B AOV-3 FTC OOpen Closed Closed Closed

(SOV-3) injection valve El. 0 Ft - TO CLOSE OC

MOV-1 HPI discharge valve MOV Aux Bldg. MCC-A1 MOV-1 FTO MOV-1 FAILS Closed Open As Is N/A El. 0 Ft - TO OPEN

MOV-2v VCT isolation valve MOV Aux Bldg. MCC-B1 MOV-2_FTC MOV-2 fails to Open Closed As Is N/A El. 0 Ft - close

MOV-3 FTO MOV-3 fails to Closed Open As Is N/A

MOV_3v' Cont. sump recirc MOV Aux Bldg. MCC-A1 open valve El. -20 Ft ••Closed Closed AslIs N/A

SMOV-4 FTO MOV-4 fails to Closed Open As Is N/A

Cont. sump recirc MOV Aux Bldg. open

valve El. -20 Ft

Closed Closed As Is N/A

MOV-5 RWST isolation valve MOV Aux Bldg. MCC-A l MOV-5 FTO MOV-5 fails to Closed Open/Close As Is N/A

El. 0 Ft open

MOV-6 RWST isolation valve MOV Aux Bldg. MCC-B1 MOV-6 FTO MOV-6 fails to Closed Open/Close As Is N/A

El. 0 Ft open d

MOV-r vlvRHR inboard suction MOV Containment MCC-A1 Closed Closed As Is N/A valveW

MOV-8 RHR vaveE. outboard suction MOV Aux 20F Bldg. MCC-B31 CClosed Closed As Is N/A

valve El. -20 Ft

MOV-9 HPI discharge valve MOV Aux Bldg. MCC-B1 MOV-9_FTO MOV-9 FAILS Closed Open As Is N/A El. 0 Ft - TO OPEN

MOV-10 AFW pump A MOV Turbine Bldg. MCC-A1 MOV-10 FTO MOV-10 fails to Closed Open As Is N/A discharge valve El. 0 Ft open

MOV-11 AFW pump B MOV Turbine Bldg. DC BUS-B MOV-11 FTO MOV-1 1 fails to Closed Open As is N/A

discharge valve El. 0 F open I I I I

B-32
Equipment PRA Event PRA Event Normal Desired Failed Failed Air Equipment ID Equipment Description Type Location Power Supply Identifier Description Position/ Position/ Electrical Position TpIdnii Derito Status Status Position

MOV-13 PORV block valve MOV Containment MCC-A1 Open OeAs is N/A ________ ~Closed _____

AFW pump B turbine Turbine Bldg. DC BUS-B MOV-14 FTO MOV-14 FAILS Closed Open As-Is N/A

MOV-14 steam line isolation MOV El. 0 Ft DCTO OPEN

I valve I I I II _II

MOV-15 AFW pump B steam

inlet throttle valve MOV Turbine Bldg.

El. 0 Ft DC BUS-B MOV-15_FTO MOV-15 FAILS Closed I I TO OPEN Throttled As-Is N/A

MOV-18 AFW pump C MOV Turbine Bldg. MCC-2 discharge valve El. 0 Ft

LI1-v"' RWST level Instrument Yard VITAL-A

LI-2 RWST level Instrument Yard VITAL-B

LI-3 Cont. sump level Instrument Containment VITAL-A

LI-4 Cont. sump level Instrument Containment VITAL-B

Tll1x Letdown heat Instrument Aux Bldg. VITAL-A exchanger outlet temp El. 0 Ft

PT-1x RCS pressure Instrument Containment VITAL-B

A-1 AFW motor high Annunciator SWG Access VITAL-A temperature Room

Closed Open As-Is N/A

Available Available Low N/A

Available Available High N/A

Available Available Low N/A

Available Available High N/A

Available Available High N/A

Available Available High N/A

Available Available Low N/A

Available Available High N/A

Available Non Unavailable N/A spurious

SWGR-Ak Train A 4160 V

switchgear Switchgear Switchgear

Room A

SUT-1 PS-A 4KV BUS A "erg"

DCEPS-4VBUSAF- from SUT1 1 1 1

Energized

from SUT-1 De-Energized N/A

B-33
Equipment PRA Event PRA Event Normal Desired Failed Failed Air Equipment ID Equipment Description Type Location Power Supply Identifier Descripton Position/ Position/ Electrical Position

TypeIdentifierDs___tion Status Status Position

PNL-AEnergized

EDG-A PNL-A 4KV BUS A Energized

DC BUS-A EPS-42BUSAF-2 FAULT from SUT- from EDG-A De-Energized N/A

B 4KV BUS A ergzed Energized DUT-1 EPS-4VBUSBF- FAULT VSUT fro SUT-1 De-Energized NWA

SWGR-B Train B 4160 V Swtcgw Swltchgear 1 1

switchgear SRoom EDG-B PNL-B 4KV BUS A Energized Energized D

DO BUS-B EPS&4V2USBF FAULT from SUT- from EDG-B e-Energized N/A

SWGR-1 Non-safety 4160 V Switchgear Turbine Bldg. UAT-1 Energized Energized De-Energized N/A switchgear Swicha Oftu B

SWGR-2 Non-safety 4160 V Swicner Turbine Bldg. UAT-1 Energized Energized De-Energized N/A

SUT-1 Startup transformer Transformer Yard OSP SUTF START-UP Energized Energized De-Energized N/A TRANSFORME

R (SUT)

FAILURE OF

Train A emergency Diesel EPSDGAF DIESEL EDG-A diesel generator Generator DG Bldg. DC BUS-A GENERATOR Standby On Off N/A

A

FAILURE OF

Train B emergency Diesel DG Bldg. DC BUS-B EPS-DGBF DIESEL Standby On Off N/A EDG-B diesel generator Generator GENERATOR

B

LC-1 centerNd LTad Center El. SST-1 Oft Energized Energized De-Energized N/A

LC-2 Non-safety 480 V load Load Center Turbine Bldg. SST-2 Energized Energized De-Energized N/A

center El. 0 ft

480V LOAD

LC-A Train A 480 V load Load Center Switchgear SST-A EPS-480VLCAF CENTER A Energized Energized De-Energized N/A Icenter II Room A PNL-AFAL II

LC-B Train B 480 V load

center Load Center Switchgear

Room B

SST-B

PNL-B

480V LOAD

EPS-480VLCBF CENTER B Energized :AI II T

Energized De-Energized N/A

SST-1 Non-safety station Transformer Turbine Bldg. SWGR-1 service transformer El. 0 F Energized Energized De-Energized N/A

B-34
Equipment PRA Event PRA Event Normal Desired Failed Failed Air Equipment ID Equipment Description Type Location Power Supply d ventiPRe Even Position/ Position/ Electrical Poitio Type Identifier Description Status Status Position Position

SST-2 Non-safety station Tr mer Turbine Bldg. SWGR-2 Energized Energized De-Energized N/A service transformer Ell. 0 F

48VLOAD

SST-A Train A station service Transformer Room A 480VLCATF TCNTSFORME Energized Energized De-Energized N/A

trnsforer Roomttea SWRA E80PS-T CANTER A

R FAILS

480V LOAD

Train B station service Switchgear SWGR-B EPS- CENTER B En SST-B transformer Transformer Room B 480VLCBXTF TRANSFORMEDe-Energized N/A

R FAILS

MCC-1 Non-safety 480 V Motor Control Turbine Bldg. LC-1 Energized Energized De-Energized N/A motor control center Center El. 0 Ft

MCC-2 Non-safety 480 V Motor Control Turbine Bldg. LC-2 Energized Energized De-Energized N/A motor control center Center El. 0 Ft

MCC-A1 Train A 480 V motor Motor Control SWG Access EPS- 480VI MC A

control center Center Room 480VMCCAl F FAULT E gnidn zd

MCC-B1 Train B 480 V motor Motor Control SWG Access LC-B EPS- 480V MCC B1 Energized Energized De-Energized N/A control center Center Room 480VMCCBlF FAULT

Automatic transfer ATS SWG Access MCC-1 Energized Energized ATS-l switch Room MCC-2 from from MCC-1 e-Energized N/A

BC-1 Non-safety swing Battery Turbine Bldg. Energized Energized De-Energized N/A battery charger Charger El. 0 Ft

Battry SitchearFAILURE OF

BC-A Train A battery charger Battery Switchgear MCC-A1 EPS-BCAF BATTERY Energized Energized De-Energized N/A Charger Room AA

Battry SitchearFAILURE OF

BC-B Train B battery charger Battery Switchgear MCC-B1 EPS-BCBF BATTERY Energized Energized De-Energized N/A Charger Room BB

Turbine Bldg.

BAT-i Non-safety battery Battery N/A Available Available Unavailable N/A El. 0 Ft

BAT-A Train a A btt ery Btt aery Room Battery A N/A EPS-BATA FAILURE BATTERY OF A Available I Available II Unavailable N/A

BAT-B Train B battery Battery Battery

Room B N/A EPS-BATB EPSBAT FAILURE BATTERY OF 8 Available Available Unavailable N/A

DC BUS-1 Non-safety 125 VDC DC Bus Turbine Bldg. BC-1 bus El. 0 Ft BAT-1 Energized Energized De-Energized N/A

B-35
Equipment PRA Event PRA Event Normal Desired Failed Failed Air Equipment ID Equipment Description Type Location Power Supply IdentifierEve n Position/ Position/ Electrical Faitir Type Identifier Description Status Status Position Position

Switchgear BC-A EPS- FAULT ON

DC BUS-A Train A 125 VDC bus DC Bus Room A BAT-A 125VDCBUSAF 125V DC BUS Energized Energized Do-Energized NIA

.A

sSwitchgear BC-B EPS- FAULT ON

DC BUS-B Train B 125 VDC DC Bus RoomBear B BAT-B 125VDCBUSBF 125V DC BUS Energized Energized De-Energized WA

NL ABTrain A 125 Pa eCB U -AT EPS- FAULT ON

panel Panel board Room A 125VDCPNLAF 125V DC Energized Energized Do-Energized N/A

_____________ ________ PANEL A_________

Train B 125 VDC Switchgear DC BUS-B EPS- FAULT ON

panel PRoom B 25VDCPNLBF 125V DC Energized Energized Do-Energized N/A

PNL-B___panel __Panel br _ Rm2 C L PANEL A

INV-A Train A inverter Inverter Switchgear

Room A DC BUS-A EPS120VBUSAINVF

rI-LUMM UJr

120V BUS A Energized Energized De-Energized N/A

INV-B Train B inverter Inverter SwiR Room t B DC BUS-B

VITAL-A Train A 120 VAC vitai 120VAC Bus SWG Access INV-A bus Room

VITAL-B Train B 120 VAC vital 120VAC Bus SWG Access INV-B bus Room

Energized Energized De-Energized N/A

120V BUS A

FAULT Energized Energized De-Energized N/A

Energized I Energized De-Energized N/A

Notes:

Legend

Appendix R

Components/failures added

to PRA for Fire PRA Model

HPI pumps A & B are in Appendix R for normal charging function; not for HPI.

i Closed to prevent LOCA /Open for feed and bleed. Appendix R has valve only to ensure remains closed.

'"AOV-2 in Appendix R only for normal letdown function.

iv AOV-3 in Appendix R only for normal charging function.

v MOV-2 in Appendix R only for normal suction to charging.

B-36
v' MOV-3 and MOV-4 need to be closed for when using RWST water supply for HPI / Open for recirculation mode of injection. MOVs 5 & 6 are correspondingly open for RWST use

and closed for recirc.

Vii Valve electrically blocked closed. Control power fuses are supposed to be removed. MOV-7 & 8 in Appendix R for both normal isolation and shutdown cooling functions.

vii If RWST level indication fails high the operator will fail to establish recirculation. If the RWST level fails low and containment sump level fails high, the operator will establish suction

to dry sump and fail the HPI the pumps due to insufficient suction. Need to resolve failure mode.

If letdown temperature indication fails low, the operator will fail to isolate the letdown line on loss of CCW. This will cause HPI pumps to cavitate due to high temperature in the

suction line.

X If RCS pressure instrument fails high, the low pressure signal to initiate emergency safeguards actuation during a lire-induced LOCA will not be initiated

For SWGR-A or SWGR-B, only one basic event is used in the model. However, the circuit analysis impact to the switchgear is a function of the power supply

B-37

NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER

(9-2004) (Assigned by NRC, Add Vol., Supp., Rev.,

NRCMD 3.7 and Addendum Numbers, if any.)

BIBLIOGRAPHIC DATA SHEET

(See instructions on the reverse) NUREG/CP-0194 Volume 1

2. TITLE AND SUBTITLE 3. DATE REPORT PUBLISHED

MONTH YEAR

Methods for Applying Risk Analysis to Fire Scenarios (MARIAFIRES) - 2008, July 2010

NRC-RES/EPRI Fire PRA Workshop, Volume 1 - Module 1: PRA/HRA

4. FIN OR GRANT NUMBER

5. AUTHOR(S) 6. TYPE OF REPORT

Conference Proceedings

Compiled and edited by D. Stroup, F. Gonzalez, and H. Woods (NRC) ConferenceProceedings

7. PERIOD COVERED (Inclusive Dates)

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (If NRC, provide Division, Office or Region, U.S. Nuclear Regulatory Commission, and mailing address; if contractor,

provide name and mailing address.)

U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research (RES), Washington, DC 20555-0001

Electric Power Research Institute (EPRI), 3420 Hillview Avenue, Palo Alto, CA 94303

9. SPONSORI NG ORGANIZATION - NAM E AND ADDRESS (IffNRC, type "Same as above",; if contractor, provide NRC Division, Office or Region, U.S. Nuclear Regulatory Commission,

and mailing address.)

U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research (RES), Washington, DC 20555-0001

Electric Power Research Institute (EPRI), 3420 Hillview Avenue, Palo Alto, CA 94303

10. SUPPLEMENTARY NOTES

NRC-RES/EPRI Fire PRA Workshops conducted Sept. 28 - Oct. 2, 2008 and Nov. 17-20, 2008 in Bethesda, MD

11. ABSTRACT (200 words or less)

The U.S. Nuclear Regulatory Commission (NRC) approved the risk-informed and performance-based alternative

regulation 10 CFR 50.48(c) in July 2004, which allows licensees the option of using fire protection requirements

contained in the National Fire Protection Association (NFPA) Standard 805, "Performance Based Standard for Fire

protection for Light-Water Reactor Electric Generating Plants, 2001 Edition," with certain exceptions. To support

licensees' use of that option, the NRC and the Electric Power Research Institute (EPRI) jointly issued NUREG/CR-6850

(EPRI 1011989) "Fire PRA Methodology for Nuclear Power Facilities," in September 2005. That report documents the

state-of-the-art methods, tools, and data for conducting a fire Probabilistic Risk Assessment (PRA) in a commercial

nuclear power plant (NPP) application. Since the release of NUREG/CR-6850 in 2005, the NRC-RES and EPRI have

conducted a number of joint public workshops to provide training in the use of the methodologies and tools contained in

the document. The workshops have attracted both domestic and international. The material in this NUREG/CP was

recorded during two workshops conducted in 2008. It was adapted by NRC-RES Fire Research Branch (FRB) members

for use as an alternative training method for those who were unable to physically attend the training sessions. This report

can also serve as a refresher for those who attended one or more training sessions, and would be useful preparatory

material for those planning to attend a session.

12. KEY WORDS/DESCRIPTORS (List words or phrases that will assist researchers in locating the report.) 13. AVAILABILITY STATEMENT

unlimited

fire, performance-based, risk-Informed regulation, fire hazard analysis (FHA), fire safety, 14. SECURITY CLASSIFICATION

fire protection, nuclear power plant, probabilistic risk assessment (PRA), fire modeling, circuit (This Page)

analysis Unclassified

(This Report)

unclassified

15. NUMBER OF PAGES

16. PRICE

NRC FORM 335 (9-2004) PRINTED ON RECYCLED PAPER

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