NUREG/CP-0303 EPRI 3002005205

Volume 1 of 5

Methods for Applying Risk

Analysis to Fire

Scenarios (MARIAFIRES)-2012

Volume 1

Module 1: Probabilistic Risk Analysis (PRA)

Based on the Joint

NRC-RES/EPRI Training Workshops

Conducted in 2012

Weeks of July 16 and September 24, 2012

Bethesda, MD

U.S. Nuclear Regulatory Commission Electric Power Research Institute

Office of Nuclear Regulatory Research 3420 Hillview Avenue

Washington, DC 20555-0001 Palo Alto, CA 94304
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Commission.
Methods for Applying Risk Analysis to Fire

Scenarios (MARIAFIRES)-2012

NRC-RES/EPRI Fire PRA Workshop

Volume 1: Module 1: PRA

NUREG/CP-0303 EPRI 3002005205

Volume 1 of 5

Manuscript Completed: July 2015

Date Published: April 2016

U.S. Nuclear Regulatory Commission Electric Power Research Institute (EPRI)

Office of Nuclear Regulatory Research (RES) 3420 Hillview Avenue

Washington, DC 20555-0001 Palo Alto, CA 94304

U.S. NRC-RES Project Manager EPRI Project Manager

M. H. Salley A. Lindeman
DISCLAIMER OF WARRANTIES AND LIMITATION OF LIABILITIES

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

OF WORK SPONSORED OR COSPONSORED BY THE ELECTRIC POWER RESEARCH INSTITUTE,

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

ORGANIZATION(S) BELOW, OR ANY PERSON ACTING ON BEHALF OF ANY OF THEM:

(A) MAKES ANY WARRANTY OR REPRESENTATION WHATSOEVER, EXPRESS OR IMPLIED, 

(I) WITH RESPECT TO THE USE OF ANY INFORMATION, APPARATUS, METHOD, PROCESS, OR

SIMILAR ITEM DISCLOSED IN THIS DOCUMENT, INCLUDING MERCHANTABILITY AND FITNESS

FOR A PARTICULAR PURPOSE, OR (II) THAT SUCH USE DOES NOT INFRINGE ON OR INTERFERE

WITH PRIVATELY OWNED RIGHTS, INCLUDING ANY PARTY’S INTELLECTUAL PROPERTY, OR (III)

THAT THIS DOCUMENT IS SUITABLE TO ANY PARTICULAR USER’S CIRCUMSTANCE; OR

(B) ASSUMES RESPONSIBILITY FOR ANY DAMAGES OR OTHER LIABILITY WHATSOEVER

(INCLUDING ANY CONSEQUENTIAL DAMAGES, EVEN IF EPRI OR ANY EPRI REPRESENTATIVE

HAS BEEN ADVISED OF THE POSSIBILITY OF SUCH DAMAGES) RESULTING FROM YOUR

SELECTION OR USE OF THIS DOCUMENT OR ANY INFORMATION, APPARATUS, METHOD,

PROCESS, OR SIMILAR ITEM DISCLOSED IN THIS DOCUMENT.

ORGANIZATION THAT PREPARED THIS DOCUMENT:

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

THE TECHNICAL CONTENTS OF THIS DOCUMENT WERE NOT PREPARED IN ACCORDANCE WITH THE

EPRI NUCLEAR QUALITY ASSURANCE PROGRAM MANUAL THAT FULFILLS THE REQUIREMENTS OF

10 CFR 50, APPENDIX B AND 10 CFR PART 21, ANSI N45.2-1977 AND/OR THE INTENT OF ISO-9001

(1994). USE OF THE CONTENTS OF THIS DOCUMENT IN NUCLEAR SAFETY OR NUCLEAR QUALITY

APPLICATIONS REQUIRES ADDITIONAL ACTIONS BY USER PURSUANT TO THEIR INTERNAL

PROCEDURES.

NOTE

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

800.313.3774 or

e-mail askepri@epri.com.

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

registered service marks of the Electric Power Research Institute, Inc.
COURSE TRAINING INSTRUCTORS AND PROJECT MANAGERS

U.S. Nuclear Regulatory Commission

Office of Nuclear Regulatory Research

Washington, DC 20555-0001

Susan Cooper – Module 4 HRA

Gabriel Taylor – Module 2 Electrical Analysis

Sandia National Laboratories

1515 Eubank SE

Albuquerque, NM 87185

Steven Nowlen – Module 3 Fire Analysis 

Jeff LaChance – Module 1 PRA

National Institute of Standards and Technology

100 Bureau Dr.

Gaithersburg, MD 20899

Kevin McGrattan – Module 5 Adv. Fire Modeling

Electric Power Research Institute

3420 Hillview Avenue

Palo Alto, CA 94303

Stuart Lewis – Module 4 HRA

Science Applications International Corp.

1671 Dell Ave, Suite 100

Campbell, CA 95008

Bijan Najafi

Francisco Joglar-Biloch – Module 3 Fire Analysis

Dan Funk (EDAN Engineering) – Module 2 Electrical

Richard Anoba (Anoba Consulting Services)–Module 1

PRA

Erin Collins –Module 4 HRA

Fred Mowrer (California Polytechnic State University) –

Module 5 Adv. Fire Modeling

Kendra Wright /Nicholas Melly

U.S. NRC-RES Project Managers

Ashley Lindeman

EPRI Project Manager
CITATIONS

This report was prepared by:

U.S. Nuclear Regulatory Commission

Office of Nuclear Regulatory Research

Washington, DC 20555-0001

Principal Investigators:

F. Gonzalez

K. Hamburger

T. Rivera

P. Smith

D. Stroup

K. Wright

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

Commission, Office of Nuclear Regulatory Research 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)-2012, NRC-RES/EPRI Fire

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

Commission, Office of Nuclear Regulatory Research, Washington, DC 20555-0001, and Electric

Power Research Institute, Palo Alto, CA, NUREG/CP-0303 and EPRI 3002005205.
NUREG/CP-0303, Volume 1 and EPRI 3002005205

has been reproduced from the best available copy

iii

ABSTRACT

The U.S. Nuclear Regulatory Commission (NRC) Office of Nuclear Regulatory Research (RES)

and the Electric Power Research Institute (EPRI) working under a memorandum of 

understanding (MOU) jointly conducted two sessions of the NRC– RES/EPRI Fire Probabilistic

Risk Assessment (PRA) Workshop on July 16–20, 2012, and September 24–28, 2012, at the 

Bethesda Marriott in Bethesda, MD. The purpose of the workshop was to provide detailed,

hands-on training on the fire PRA methodology described in the technical document, 

NUREG/CR-6850 (EPRI 1011989) entitled “EPRI/NRC-RES Fire PRA Methodology for

Nuclear Power Facilities.” This fire PRA methodology document supports implementation of

the risk-informed, performance-based rule in Title 10 of the Code of Federal Regulations

(10 CFR) 50.48(c) endorsing National Fire Protection Association (NFPA) Standard 805, as

well as other applications such as exemptions or deviations to the agency’s current regulations

and fire protection significance determination process (SDP) phase 3 applications.

RES and EPRI provided training in five subject areas related to fire PRA, namely: fire PRA,

electrical analysis, fire analysis, fire human reliability analysis (HRA), and advanced fire

modeling. Participants selected one of these subject areas and spent the duration of the

course in that module. The HRA module reviewed guidance provided in NUREG-1921 (EPRI

1023001), “EPRI/NRC-RES Fire Human Reliability Analysis Guidelines,” while the fire

modeling module reviewed the fire modeling guidance provided in NUREG-1934 (EPRI

1019195), “Nuclear Power Plant Fire Modeling Application Guide.” For each technical area,

the workshop also included a 1-day module introducing the fundamentals of the subject. The

purpose of the fundamentals modules was to assist students without an extensive background

in the technical area in understanding the in-depth training modules that followed. Attendance

in the fundamentals modules was optional. The workshop’s format allowed for in-depth

presentations and practical examples directed toward the participant’s area of interest.

This NUREG/CP documents both of the two sessions of the NRC-RES/EPRI Fire PRA

Workshop delivered in 2012 and includes the slides and handout materials delivered in each

module of the course as well as video recordings of the training that was delivered. This

NUREG/CP can be used 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 could also be useful preparatory material for

those planning to attend future sessions.

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’s conclusions or regulatory positions, 

and they should not be interpreted as such.

v

CONTENTS

ABSTRACT ............................................................................................................................iii

CONTENTS ............................................................................................................................ v

FIGURES...............................................................................................................................vii

ACKNOWLEDGMENTS .........................................................................................................ix

ACRONYMS...........................................................................................................................xi

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

1.1 About this Text....................................................................................................... 1-3

References..................................................................................................................... 1-8

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

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

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

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

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

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

2.2.4 Auxiliary Feedwater System..................................................................... 2-4

2.2.5 Electrical System ..................................................................................... 2-5

2.2.6 Other Systems......................................................................................... 2-5

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

2.4 SNPP Drawings..................................................................................................... 2-7

 PRINCIPLES OF PRA .................................................................................................. 3-1

Basics of Nuclear Power Plant Probabilistic Risk Assessment ....................................... 3-2

Overview of PRA............................................................................................................ 3-3

Initiating Event Analysis................................................................................................ 3-12

Accident Sequence Analysis ........................................................................................ 3-19

Systems Analysis ......................................................................................................... 3-34

Human Reliability Analysis ........................................................................................... 3-57

Data Analysis ............................................................................................................... 3-72

Accident Sequence Quantification................................................................................ 3-89

LEVEL 2/LERF Analysis............................................................................................. 3-106

 MODULE 1 FIRE PROBABILISTIC RISK ASSESMENT.............................................. 4-1

Introduction and Overview: The Scope and Structure of PRA/Systems Analysis Module 4-3

Sample Plant Description ............................................................................................. 4-13

Task 2 – Fire PRA Component Selection ..................................................................... 4-27

Task 5 – Fire-Induced Risk Model Development .......................................................... 4-51
vi

Task 4 - Qualitative Screening................................................................................... 4-71

Task 7 - Quantitative Screening ................................................................................ 4-76

Task 14 – Fire Risk Quantification................................................................................ 4-85

Task 15 – Uncertainty and Sensitivity Analysis............................................................. 4-95
vii

FIGURES

Figure 1-1 Relationship of Technical Tasks in NUREG/CR 6850 Volume 2 ........................ 1-6

Figure 1-2 Note: "B" is from Task 7B (Previous Page) ........................................................ 1-7

Figure 2-1 General Plant Layout ......................................................................................... 2-8

Figure 2-2 Plant Layout Section AA .................................................................................... 2-9

Figure 2-3 Auxiliary Building - Elevation 20 Ft................................................................... 2-10

Figure 2-4 Auxiliary Building – Elevation 0 Ft .................................................................... 2-11

Figure 2-5 Auxiliary Building – Elevation +20 Ft. ............................................................... 2-12

Figure 2-6 Auxiliary Building – Elevation +40 Ft. ............................................................... 2-13

Figure 2-7 Auxiliary Building Main Control Room .............................................................. 2-14

Figure 2-8 Turbine Building – Elevation 0 Ft...................................................................... 2-15

Figure 2-9 Main Control Board .......................................................................................... 2-16

Figure 2-10 Primary System P&ID ...................................................................................... 2-17

Figure 2-11 Secondary System P&ID.................................................................................. 2-18

Figure 2-12 Electrical One-Line Diagram ............................................................................ 2-19

ix

ACKNOWLEDGMENTS

The authors of this report greatly appreciate the contributions made by instructors and students 

at the 2012 NRC-RES/EPRI Fire PRA Workshop.

In addition, we want to extend our gratitude to Tojuana Fortune-Grasty (NRC’s publications

analyst) and the NRC’s printing specialist’s team whose invaluable support and expertise

were critical to ensuring the published report’s quality. We also extend a special thanks and

appreciation to Anita Aikins-Afful (RES/DRA administrative assistant) for providing the technical 

edit of this report.

xi

ACRONYMS

ACB Air-cooled Circuit Breaker

ACRS Advisory Committee on Reactor Safeguards

AEP Abnormal Event Procedure

AFW Auxiliary Feedwater

AGS Assistant General Supervisor

AOP Abnormal Operating Procedure

AOV Air Operated Valve

ASEP Accident Sequence Evaluation Program

ATHEANA A Technique for Human Event Analysis

ATS Automatic Transfer Switch

ATWS Anticipated Transient Without Scram

BAT Boric Acid Tank

BNL Brookhaven National Laboratory

BWR Boiling-Water Reactor

CBDT Cause-Based Decision Tree

CCDP Conditional Core Damage Probability

CF Cable (Configuration) Factors

CCPS Center for Chemical Process Safety

CCW Component Cooling Water

CDF Core Damage Frequency

CFD Computational Fluid Dynamics

CFR Code of Federal Regulations

CLERP Conditional Large Early Release Probability

CM Corrective Maintenance

CR Control Room

CRS Cable and Raceway (Database) System

CST Condensate Storage Tank

CVCS Chemical and Volume Control System

CWP Circulating Water Pump

DC Direct Current

ECCS Emergency Core Cooling System

EDG Emergency Diesel Generator

EDS Electrical Distribution System

EF Error Factor

EI Erroneous Status Indicator

EOP Emergency Operating Procedure

EPR Ethylene-Propylene Rubber

EPRI Electric Power Research Institute

ET Event Tree

FEDB Fire Events Database

FEP Fire Emergency Procedure
xii

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

HCR Human Cognitive Reliability

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

INPO 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

LP/SD Low Power and Shutdown

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
xiii

NPP Nuclear Power Plant

NPSH Net Positive Suction Head

NQ cable Non-Qualified (IEEE-383) cable

NRC U.S. Nuclear Regulatory Commission

ORE Operator Reliability Experiments

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

SPAR-H Standardized Plant Analysis Risk HRA

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

T-H Thermal Hydraulic

THERP Technique for Human Error Rate Prediction
xiv

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)

VVER The Soviet (now Russian Federation) designation for light-water pressurized 

reactor

XLPE Cross-Linked Polyethylene

ZOI Zone of Influence
INTRODUCTION

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

performance- based alternative regulation in Title 10 of the Code of Federal Regulations

(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’s Office of Nuclear Regulatory Research (RES) 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 state-of-the art methods, tools, 

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

power plant (NPP) application. This report is intended to serve the needs of a fire risk analysis

team by providing a general framework for conducting 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 (IPEEE), 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 the NRC’s participation in

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

analysis contained in that document. The analyses and methods documented in that 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. However, the results from that 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.

However, it is not sufficient to merely develop a potentially useful method, such as

NUREG/CR- 6850, and announce its availability. It is also necessary to teach potential users

how to properly use the method. To meet this need RES and EPRI have collaboratively

conducted the NRC-RES/EPRI Fire PRA Workshops to train interested parties in the application 

of this methodology since 2005. The course is provided in five parallel modules covering tasks 

from NUREG/CR-6850 “EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities” 

Reference [1].

These five training modules are:
1-2

• Module 1: PRA/Systems Analysis – This module covers the technical tasks for

development of the system response to a fire including human failure events. 

Specifically, this module covers Tasks/Sections 2, 4, 5, 7, 14, and 15 of Reference [1].

• Module 2: Electrical Analysis – This module covers the technical tasks for analysis of

electrical failures as the result of a fire. Specifically, this module covers

Tasks/Sections 3, 9, and 10 of Reference [1].

• Module 3: Fire Analysis – This module covers technical tasks involved in development of 

fire scenarios from initiation to target (e.g., cable) impact. Specifically, this module 

covers Tasks/Sections 1, 6, 8, 11, and 13 of Reference [1]. 

• Module 4: Fire Human Reliability Analysis – This module covers the technical tasks

associated with identifying and analyzing operator actions and performance during a

postulated fire scenario. Specifically, this module covers Task 12 as outlined in

Reference [1] based on the application of the approaches documented in Reference [2].

• Module 5: Advanced Fire Modeling – This module was added to the training in 2011. It

covers the fundamentals of fire science and provides practical implementation guidance

for the application of fire modeling in support of a fire PRA. Module 5 covers fire

modeling applications for Tasks 8 and 11 as outlined in Reference [1] based on the

material presented in Reference [3].

The first three modules are based directly on the “EPRI/NRC-RES Fire PRA Methodology for

Nuclear Power Facilities,” EPRI 1011989, and NUREG/CR-6850 [1]. However, that document

did not cover fire human reliability analysis (HRA) methods in detail. In 2010, the training

materials were enhanced to include a fourth module based on a more recent EPRI/RES

collaboration and the then draft guidance document, EPRI 1019196, NUREG-1921 [2] published 

in late 2009. The training materials are based on this draft document including the

consideration of public comments received on the draft report and the team’s responses to

those comments. In 2011 a fifth training module on Advanced Fire Modeling techniques and

concepts was added to the course. This module is based on another joint RES/EPRI

collaboration and a draft guidance published in January 2010, NUREG-1934 EPRI 1019195 [3].

In 2012 an additional first day of training was included in the NRC-RES/EPRI Fire PRA

Workshop to cover principal elements of each technical area covered in the Fire PRA course,

i.e., PRA, HRA, Electrical Analysis, and Fire Analysis. This introductory module was intended to

assist in preparing the students to understand the in-depth fire PRA training modules that

followed. The introductory modules were not intended to be a substitute for education and/or

training in the subject matter. The intent was that they would serve as a primer for those

individuals who lacked such training or those who were cross-training in an area other than their

primary area of expertise.

The four introductory modules listed below (referred to as Module 0) were offered in parallel on

the first day of the workshop.

Module 0a: Principles of PRA

Module 0b: Principles of Electrical Analysis

Module 0c: Principles of Fire Science and Modeling

Module 0d: Principles of HRA
1-3

These sub-modules are included in the text and on the accompanying DVDs as a part of their 

related module. 

1.1 About this Text

“Methods for Applying Risk Analysis to Fire Scenarios (MARIAFIRES) – 2012”, is a collection of

the materials that were presented at the two sessions of the NRC-RES/EPRI Fire PRA

conducted July 16–20, 2012, and September 24-28, 2012.

The 2012 workshop was video recorded and adapted as an alternative training method for those

who were unable to physically attend the training sessions. This NUREG/CP is comprised of the

materials supporting those videos and includes the five volumes below (the videos are enclosed

on DVD in the published paper copies of this NUREG/CP). This material can also serve as a

refresher for those who attended one or more of the training sessions, and would be useful 

preparatory material for those planning to attend a session.

MARIAFIRES is comprised of 5 volumes.

Volume 1 – Module 0a Principles of PRA and Module 1: PRA/Systems Analysis 

Volume 2 – Module 0b Principles of Electrical Analysis and Module 2: Electrical Analysis 

Volume 3 – Module 0c Principles of Fire Science and Modeling and Module 3: Fire Analysis 

Volume 4 – Module 0d Principles of HRA and Module 4: Fire Human Reliability Analysis

Volume 5 – Module 5: Advanced Fire Modeling 

Integral to Modules 1, 2 and 3 is a set of hands-on problems based on a conceptual generic

nuclear power plant (NPP) developed for training purposes. This generic plant is referred to in

this text and in classroom examples as SNPP (Simplified Nuclear Power Plant). The same

generic NPP is used in all three modules. Chapter 2 of this document provides the background

information for the problem sets of each module, including a general description of the sample

power plant and the internal events PRA needed as input to the fire PRA. The generic NPP

defined for this training is an extremely simplified one that in many cases does not meet any

regulatory requirements or good engineering practices. For training purposes, the design

features presented highlight the various aspects of the fire PRA methodology.

For Module 4 and 5, independent sets of examples are used to illustrate key points of the 

analysis procedures. The examples for these two modules are not tied to the simplified plant. 

Module 4 uses examples that were derived largely from pilot applications of the proposed fire 

HRA methods and on independent work of the EPRI and RES HRA teams. The examples for

Module 5 were taken directly from Reference [3] and represent a range of typical NPP fire 

scenarios across a range of complexity and that highlight some of the computation challenges 

associated with the NPP fire PRA fire modeling applications.

A short description of the Fire PRA technical tasks is provided below. For further details, refer to 

the individual task descriptions in EPRI 1011989, NUREG/CR-6850, Volume 2. The figure 

presented at the end of this chapter provides a simplified flow chart for the analysis process and 

indicates which training module covers each of the analysis tasks.

Plant Boundary Definition and Partitioning (Task 1). The first step in 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.
1-4

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 Part 50, 

“Domestic Licensing of Production and Utilization Facilities,” Appendix R, “Fire Protection 

Program for Nuclear Power Facilities Operating prior to January 1, 1979,” 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 because of concern for 

multiple spurious actuations that may threaten the credited functions and components, as well

as from concerns about fire effects on instrumentation used by the plant crew to respond to the 

event.

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

considerations associated with identifying cables supporting those components selected in

Task 2 above. In previous fire PRA methods (such as EPRI Fire-Induced Vulnerability 

Evaluation (FIVE) and Fire PRA Implementation Guide), this task was relegated to the 

SSD analysis and its associated databases. NUREG/CR-6850 (EPRI 1011989) 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.

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 plans. These procedures may impact availability of

functions and components or include fire-specific operator actions (e.g., self- induced station

blackout).

Fire Ignition Frequency (Task 6). This task describes the approach to develop frequency

estimates for fire compartments and scenarios. Significant changes from the EPRI FIVE method

have been made in this task. The changes generally relate to the use of challenging events,

considerations associated with data quality, and increased use of a fully component-based

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

example, in FIVE).

Quantitative Screening (Task 7). A fire PRA allows the screening of fire compartments and

scenarios based on their contribution to fire risk. This approach considers the cumulative risk

associated with the screened compartments (i.e., the ones not retained for detailed analysis) to

ensure that a true estimate of fire risk profile (as opposed to vulnerability) is obtained.

Scoping Fire Modeling (Task 8). This step provides simple rules to define and screen fire

ignition sources (and therefore fire scenarios) in an unscreened fire compartment.

Detailed Circuit Failure Analysis (Task 9). This task provides an approach and technical

considerations for identifying how the failure of specific cables will impact the components

included in the fire PRA SSD plant response model.
1-5

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 

NUREG/CR-6850 (EPRI 1011989) benefits from the knowledge gained from the tests performed

in response to the circuit failure issue.

Detailed Fire Modeling (Task 11). This task describes the method to examine the

consequences of a fire. This includes consideration of scenarios involving single compartments,

multiple fire compartments, and the main control room. Factors considered include initial fire

characteristics; fire growth in a fire compartment or across fire compartments; detection and

suppression; electrical raceway fire barrier systems, and damage from heat and smoke. Special

consideration is given to turbine generator (T/G) fires, hydrogen fires, high-energy arcing faults

(HEAF), 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.

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

of the fire risk results.

Uncertainty and Sensitivity Analyses (Task 15). This task describes the approach to follow 

for identifying and treating uncertainties throughout the fire PRA process. The treatment may 

vary from quantitative estimation and propagation of uncertainties where possible (e.g., in fire 

frequency and non-suppression probability) to identification of sources without quantitative 

estimation. The treatment may also include one-at-a-time variation of individual parameter 

values or modeling approaches to determine the effect on the overall fire risk (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. 
1-6

 

 

 

 

TASK 1: Plant Boundary & 

Partitioning

TASK 2: Fire PRA Comp

Selection

TASK 3: Fire PRA Cable 

Selection 

TASK 4: Qualitative Screening

TASK 6: Fire Ignition 

Frequencies

TASK 5: Fire-Induced 

Model

TASK 7A: Quantitative 

Screening - I

TASK 8: Scoping Fire Modeling

SUPPORT TASK A: Plant 

Walk Downs

SUPPORT TASK B: Fire PRA 

Database

TASK 7B: Quantitative 

Screening - II

TASK 12A: Post-Fire 

Screening

B

Fire Analysis Mod

PRA/System Mod

Circuits Module

HRA Module

Fire Analysis and 

Modeling Module

Figure 0-1 Relationship of Technical Tasks in NUREG/CR 6850 Volume 2
1-7

 

 

 

 

Detailed Fire Scenario Analysis

B

TASK 11: Detailed Fire Modeling 

A. Single Compartment

B. Multi-Compartment 

C. Main Control Room 

TASK 9: Detailed Circuit Failure 

Analysis

TASK 10: Circuit Failure Mode & 

Likelihood Analysis

TASK 14: Fire Risk Quantification

TASK 15: Uncertainty & 

Sensitivity Analyses

TASK 16: Fire PRA 

Documentation

TASK 12B: Post fire HRA

Detailed & recovery TASK 13: Seismic-Fire 

Interactions

Fire Analysis Modu

PRA/System Modu

Circuits Module

HRA Module

Fire Analysis and F

Modeling Modules

Figure 0-2 Note: "B" is from Task 7B (Previous Page)
1-8

References

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

Nuclear Power Facilities, September 2005.

2. NUREG-1921, EPRI 1023001, EPRI/NRC-RES Fire Human Reliability Analysis Guidelines, May

2012.

3. NUREG-1934, EPRI 1023259, Nuclear Power Plant Fire Modeling Application Guide,

November 20121

.

 1 At the time of the 2012 NRC-RES/EPRI Fire PRA Workshop, this final report had not yet been 

published. A draft for public comment was used to conduct the training. 
2-1

EXAMPLE CASE PLANT-GENERAL INFORMATION

2.1 Overall Plant Description

This chapter provides background information about the generic plant used in the hands-on 

problem sets of Modules 1, 2 and 3. Note that the examples used in Module 4 (HRA) are not 

based on the example case plant. The following notes generally describe the example case 

plant, including its layout:

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

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

chemical volume control system and multiple train High Pressure Injection system, as well 

as a single train residual heat removal system interface with the primary system

2. The secondary side of the plant contains a main steam and feedwater loop associated with 

the single steam generator, and a multiple train auxiliary feedwater system to provide 

decay heat removal.

3. The operating conditions and parameters of this plant are similar to that of a typical PWR.

For example, the primary side runs at about 2,200 psi pressure. The steam generator can 

reject the decay heat after a reactor trip. There is a possibility for feed and bleed.

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

5. The plant is laid out in accordance with Figures 1 through 9. The plant consists of a 

containment building, auxiliary building, turbine building, diesel generator building and the 

yard. All other buildings and plant areas are shown but no details are provided.

2.2 Systems Description

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

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

2.2.1 Primary Coolant System

The following notes and Figure 10 define the primary coolant system:

1. The primary coolant loop consists of the reactor vessel, one reactor coolant pump, and one 

steam generator and the pressurizer, along with associated piping.

2. 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.
2-2

3. The pressure transmitter (PT-1) on the pressurizer provides the pressure indication for the

 primary coolant system and is used to signal a switch from chemical volume control system 

(CVCS) to high pressure injection (HPI) configuration. That is, PT-1 provides the automatic 

signal for high pressure injection on low RCS pressure. It also provides the automatic 

signal to open the PORV on high RCS pressure.

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

loss of plant air but does not have sufficient capacity to support long-term operation.

2.2.2 Chemical Volume Control and High Pressure Injection Systems

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

1. The CVCS normally operates during power generation.

2. Valve type and position information include:

Table 2-1 Chemical volume control and high pressure injection systems valve type and

position information

Valve Type Status on Loss of 

Power (Or Air as 

applicable)

Position During 

Normal 

Operation

Motor Power 

(hp)

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

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

MOV-1 Motor Operated 

Valve

Fail As Is Closed >5

MOV-2 Motor Operated 

Valve

Fail As Is Open <5

MOV-3 Motor Operated 

Valve

Fail As Is Closed <5

MOV-4 Motor Operated 

Valve

Fail As Is Closed <5

MOV-5 Motor Operated 

Valve

Fail As Is Closed <5

MOV-6 Motor Operated 

Valve

Fail As Is Closed >5

MOV-9 Motor Operated 

Valve

Fail As Is Closed >5

3. One of the two HPI pumps runs when the CVCS is operating.

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

postulated accident conditions.

5. HPI and CVCS use the same set of pumps.

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

• AOV-3 closes.
2-3

• MOV-5 and MOV-6 open.

• MOV-2 closes.

• Both HPI pumps receive start signal, the stand-by pump starts and the operating

pump continues operating.

• MOV-1 and MOV-9 open.

7. HPI supports feed and bleed cooling when all secondary heat removal is unavailable. When 

there is a low level indication on the steam generator, the operator will initiate feed and bleed

cooling by starting the HPI pumps and opening the PORV.

8. HPI is used for re-circulating sump water after successful injection in response to a 

loss-of-coolant accident (LOCA) or successful initiation of feed and bleed cooling. For 

recirculation, upon proper indication of low refueling water storage tank (RWST) level and

sufficient sump level, the operator manually opens MOV-3 and MOV-4, closes MOV-5 and 

MOV-6, starts the RHR pump, and aligns component cooling water (CCW) to the RHR heat 

exchanger.

9. RWST provides the necessary cooling water for the HPI pumps during injection. During the 

recirculation mode, HPI pump cooling is provided by the recirculation water.

10. 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.

11. The air compressor provides the motive power for the air-operated valves but the detailed 

connections to the various valves are not shown.

2.2.3 Residual Heat Removal System

The following notes and Figure 10 define the residual heat removal (RHR) system:

1. The design pressure of the RHR system downstream of MOV-8 is low.

2. Valve type and position information include the following:

Table 2-2 Residual Heat Removal System Valve Type and Position Information

Valve Type Status on Loss 

of Power

Position During 

Normal 

Operation

Motor Power 

(hp)

MOV-7 Motor Operated 

Valve

Fail As Is Closed (breaker 

racked out)

>5

MOV-8 Motor Operated 

Valve

Fail As Is Closed >5

MOV-20 Motor Operated 

Valve

Fail As Is Closed >5
2-4

3. Operators have to align the system for shutdown cooling, after reactor vessel

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

on and establish cooling in the RHR heat exchanger.

2.2.4 Auxiliary Feedwater System

The following notes and Figure 11 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-A is motor-driven, AFW-B is steam turbine-driven, and AFW-C is diesel-driven.

3. Valve type and position information include the following:

Table 2-3 Auxiliary Feedwater System Valve Type and Position Information

Valve Type Status on Loss 

of Power

Position During 

Normal 

Operation

Motor Power 

(hp)

MOV-10 Motor Operated 

Valve

Fail As Is Closed >5

MOV-11 Motor Operated 

Valve

Fail As Is Closed >5

MOV-14 Motor Operated 

Valve

Fail As Is Closed <5

MOV-15 Motor Operated 

Valve

Fail As Is Closed <5

MOV-16 Motor Operated 

Valve

Fail As Is Closed <5

MOV-17 Motor Operated 

Valve

Fail As Is Closed <5

MOV-18 Motor Operated 

Valve

Fail As Is Closed >5

MOV-19 Motor Operated 

Valve

Fail As Is Closed <5

4. Upon a plant trip, main feedwater isolates and AFW automatically initiates by starting AFW-A

and AFW-C pumps, opening the steam valves MOV-14 and MOV-15 to operate the AFW-B

steam-driven pump, and opening valves MOV-10, MOV-11, and MOV-18.

5. The 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.
2-5

8. The atmospheric relief valve opens, as needed, automatically to remove decay heat if the 

main condenser path should 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.

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

tube rupture.

2.2.5 Electrical System

Figure 12 is a one-line diagram of the Electrical Distribution System (EDS). Safety-related

buses are identified by the use of alphabetic letters (e.g., SWGR-A, MCC-B1, etc.) while the 

non-safety buses use numbers as part of their designations (e.g., SWGR-1 and MCC-2).

The safety-related portions of the EDS include 4,160-volt (V) switchgear buses SWGR-A and

SWGR-B, which are normally powered from the startup transformer SUT-1. In the event that

offsite power is lost, these switchgear buses receive power from emergency diesel generators

EDG-A and EDG-B. The 480-V safety-related load centers (LC-A and LC-B) receive power 

from the switchgear buses via station service transformers SST-A and SST-B. The motor

control centers (MCC-A1 and MCC-B1) are powered directly from the load centers. The MCCs

provide motive power to several safety-related motor-operated valves (MOVs) and to dc buses

DC BUS-A and DC BUS-B via battery chargers BC-A and BC-B. The two 125-V dc 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 4,160-V, safety-related switchgear is provided through distribution panels

PNL-A and PNLB. The 120 V ac vital loads are powered from buses VITAL-A and VITAL-B,

which in turn receive their power from the dc buses through inverters INV-A and INV-B.

The non-safety portions of the EDS reflect a similar hierarchy of power flow. There are

important differences, however. For example, 4,160-V SWGR-1 and SWGR-2 are normally

energized from the unit auxiliary transformer (UAT-1) with backup power available from SUT-1. 

A cross-tie breaker allows one non-safety switchgear bus to provide power to the other. Nonsafety load centers LC-1 and LC-2 are powered at 480 V from the 4,160-V switchgear via SST1 and SST-2. These load centers provide power directly to the non-safety MCCs. The non-vital 

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

and battery charger BC-1 or directly from the 125-V dc battery, BAT-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:

• Component Cooling Water (CCW) – provides cooling to letdown heat exchanger and the 

RHR heat exchanger– assumed to be available at all times.

• It is assumed that the control rods can successfully insert and shutdown the reactor 

under all conditions.
2-6

• It is assumed that the emergency core cooling system (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 

other 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.). 

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

2.3 Plant Layout

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

through A09):

• The main structures of the plant are as follows:

- containment

- auxiliary building

- turbine building

- diesel generator building

- intake structure

- security building

• In Figure 1 (Drawing A-01), the dashed lines represent the fence that separates two 

major parts: the yard and switchyard.

• Switchyard is located outside the yard with a separate security access.

• CST, RWST, UAT, main transformer and SUT are located in the open in the yard.

• All walls shown in Figures 1 through 8 (Drawings A-01 through A-08) should be 

assumed to be fire rated.

• All doors shown in Figures 1 through 8 (Drawings A-01 through A-08) should be 

assumed as fire rated and normally closed.

• Battery rooms A and B are located inside the respective switchgear rooms with 1-hour 

rated walls, ceilings and doors.

• All cable trays are open type. Vertical cable trays are designated as VCBT and 

horizontal cable trays as HCBT. For horizontal cable trays, the number following the 

letters indicates the elevation of the cable tray. For example, HCBT+35A denotes a 

horizontal cable tray at elevation +35 ft (11 meters).

• The stairwell in the auxiliary building provides access to all the floors of the building. The 

doors and walls are fire rated and doors are normally closed.
2-7

2.4 SNPP Drawings

The following 12 pages provide schematic drawings of the generic NPP, SNPP. Drawings A-01

through A-09 are general physical layout drawings providing plan and elevation views of the

plant. These drawings also identify the location of important plant equipment. Drawing A-10

provides a piping and instrumentation diagram (P&ID) for the primary coolant system, and

drawing A-11 provides a P&ID for the secondary systems. Drawing A-12 is a simplified one-line

diagram of the plant power distribution system.
2-8

Figure 2-1 General Plant Layout
2-9

Figure 2-2 Plant Layout Section AA
2-10

Figure 2-3 Auxiliary Building - Elevation 20 Ft.
2-11

Figure 2-4 Auxiliary Building – Elevation 0 Ft
2-12

Figure 2-5 Auxiliary Building – Elevation +20 Ft.
2-13

Figure 2-6 Auxiliary Building – Elevation +40 Ft.
2-14

Figure 2-7 Auxiliary Building Main Control Room
2-15

Figure 2-8 Turbine Building – Elevation 0 Ft.
2-16

Figure 2-9 Main Control Board
2-17

Figure 2-10 Primary System P&ID
2-18

Figure 2-11 Secondary System P&ID
2-19

Figure 2-12 Electrical One-Line Diagram

3-1

PRINCIPLES OF PRA

The slides that follow were presented on the first day of the NRC-RES/EPRI Fire PRA 

Workshop during the extra day of training dedicated to presenting the fundamentals of the 

various subject areas to be covered during the remainder of the week. 
3-2

Basics of Nuclear Power Plant Probabilistic Risk Assessment
3-3

Overview of PRA
3-4
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Initiating Event Analysis
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Accident Sequence Analysis
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Systems Analysis
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Human Reliability Analysis
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Data Analysis
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Accident Sequence Quantification
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LEVEL 2/LERF Analysis
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4-1

MODULE 1 FIRE PROBABILISTIC RISK ASSESMENT

The following is a short description of the Fire PRA technical tasks covered in Module 1. For 

further details, refer to the individual task descriptions in Volume 2 of EPRI 1011989, 

NUREG/CR-6850.

• Fire PRA Component Selection (Task 2). The selection of components that are to be 

credited for plant shutdown following a fire is a critical step in any Fire PRA. Components 

selected would generally include many components credited in the 10 CFR 50 Appendix R 

post-fire SSD analysis. Additional components will likely be selected, potentially including 

any and all components credited in the plant’s internal events PRA. Also, the proposed 

methodology would likely introduce components beyond either the 10 CFR 50 Appendix R 

list or the internal events PRA model. Such components are often of interest due to 

considerations of multiple spurious actuations that may threaten the credited functions and 

components.

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

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

may be screened out if they contain no components or cables identified in Tasks 2 and 3, 

and if they cannot lead to a plant trip due to either plant procedures, an automatic trip signal, 

or technical specification requirements.

• Plant Fire-Induced Risk Model (Task 5). This task discusses steps for the development

of a logic model that reflects plant response following a fire. Specific instructions have been 

provided for treatment of fire-specific procedures or preplans. These procedures may impact 

availability of functions and components, or include fire-specific operator actions

(e.g., self-induced-station-blackout).

• 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.

• Post-Fire Human Reliability Analysis (Task 12). This task considers operator actions for 

manipulation of plant components. Task 12 is covered in limited detail in the 

PRA/Systems module. In particular, those aspects of Task 12 that deal with identifying and 

incorporating human failure events (HFEs) into the plant response model are discussed. 

Methods for quantifying human error probabilities (HEPs) are deferred to Module 4.

• 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 
4-2

possible (e.g., in fire frequency and non-suppression probability) to identification of 

sources without quantitative estimation. The treatment may also include one-at-a-time 

variation of individual parameter values or modeling approaches to determine the effect 

on the overall fire risk (sensitivity analysis).
4-3

Introduction and Overview: The Scope and Structure of PRA/Systems Analysis 

Module
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Sample Plant Description
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Task 2 –Fire PRA Component Selection
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Task 5 – Fire-Induced Risk Model Development
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Task 4 - Qualitative Screening
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Task 7 - Quantitative Screening
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Task 14 – Fire Risk Quantification
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Task 15 – Uncertainty and Sensitivity Analysis
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NUREG/CP-0303, Vol. 1 Methods for Applying Risk Analysis to Fire Scenarios (MARIAFIRES) – 2012 

Module 1: PRA