State-of-the-Art Reactor Consequence Analyses: 

Using decades of research and experience to 

model accident progression, mitigation, emergency 

response, and health effects

MODELING POTENTIAL REACTOR 

ACCIDENT CONSEQUENCES

MODELING POTENTIAL REACTOR ACCIDENT CONSEQUENCES | i

PERSONAL STATEMENT FROM BRIAN SHERON 

Director

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

The U.S. Nuclear Regulatory Commission’s 

(NRC’s) Office of Nuclear Regulatory Research 

is a congressionally mandated office that plans, 

recommends, and implements a program of nuclear 

regulatory research, standards development, and 

resolution of generic issues for nuclear power plants 

and other NRC-regulated facilities. We partner 

with other NRC offices, Federal agencies, industry 

research organizations, and international organizations to conduct these activities.

We started the State-of-the-Art Reactor Consequence Analyses (SOARCA) research 

project to estimate the potential health effects from the unlikely event of a commercial 

nuclear power plant accident releasing significant quantities of radioactive material into 

the environment. This project modeled a set of important accident scenarios for two 

plants, Peach Bottom and Surry, which represent two of the most common types of 

plants licensed in the United States. SOARCA considers plant design and operational 

changes not reflected in earlier assessments. The project also takes into account NRC’s 

development of rigorous oversight processes and use of operating experience along 

with improvements in operator training and emergency preparedness. We’ve also 

incorporated decades of national and international research into the tools that NRC 

used to perform this study. 

One of SOARCA’s objectives is explaining severe-accident-related aspects of nuclear 

safety to NRC stakeholders. Stakeholders include members of the public along 

with Federal, State, and local authorities and the companies that hold NRC licenses 

to operate nuclear power plants. SOARCA meets this communication objective 

by documenting its results in reports: NUREG-1935, “State-of-the-Art Reactor 

Consequence Analyses Main Report,” and NUREG/CR-7110, Volume 1, “Peach Bottom 

Integrated Analyses,” and Volume 2, “Surry Integrated Analyses.” Because the NUREG 

reports rely on highly technical explanations, this brochure was developed as a plainlanguage summary of SOARCA’s methods, results, and conclusions. We invite you to 

read this brochure to understand how we used state-of-the-art methods to model these 

unlikely nuclear power plant accidents to understand their potential impact on public 

health and safety.
ii | U.S. Nuclear Regulatory Commission

ACKNOWLEDGEMENTS:

This brochure was the result of the efforts of many NRC staff and the staff of its 

contractors, Sandia National Laboratories, and dycoda, LLC. The contributions of 

Dorothy Andreas (née Collins) and Scott Burnell are especially appreciated. The 

SOARCA team also wishes to thank Eric Knowles and Woody Machalek for their help 

with the graphics and layout.

The SOARCA project team consisted of the following individuals. This group also 

benefited greatly from the contributions of other selected NRC staff members, 

including senior risk analysts from NRC’s headquarters and regional offices, in specific 

areas of the project.

Dorothy Andreas (née Collins)

Jonathan Barr

Terry Brock

Scott Burnell 

Richard Chang

Donald Chung 

Donald Dube 

Tina Ghosh

Chris Hunter

Jocelyn Mitchell

AJ Nosek

José Pires

Robert Prato

Jason Schaperow

Martin Stutzke

Randolph Sullivan

Charles Tinkler

James Vail

Thomas Weaver

Kathy Halvey Gibson

Brian Holian

Patricia Santiago

Brian W. Sheron

Jennifer Uhle

Jimi Yerokun

Management oversight was provided by the following individuals:
MODELING POTENTIAL REACTOR ACCIDENT CONSEQUENCES | iii

KEY RESULTS:

 When operators are successful in using available onsite equipment during the 

accidents analyzed in SOARCA, they can prevent the reactor from melting, or delay or 

reduce releases of radioactive material to the environment.

 SOARCA analyses indicate that all modeled accident scenarios, even if operators 

are unsuccessful in stopping the accident, progress more slowly and release much 

smaller amounts of radioactive material than calculated in earlier studies.

 As a result, public health consequences from severe nuclear power plant accidents 

modeled in SOARCA are smaller than previously calculated.

 The delayed releases calculated provide more time for emergency response actions 

such as evacuating or sheltering for affected populations. For the scenarios analyzed, 

SOARCA shows that emergency response programs, if implemented as planned and 

practiced, reduce the risk of public health consequences.

 Both mitigated (operator actions are successful) and unmitigated (operator actions 

are unsuccessful) cases of all modeled severe accident scenarios in SOARCA cause 

essentially no risk of death during or shortly after the accident.

 SOARCA’s calculated longer term cancer fatality risks for the accident scenarios 

analyzed are millions of times lower than the general U.S. cancer fatality risk.
iv | U.S. Nuclear Regulatory Commission
MODELING POTENTIAL REACTOR ACCIDENT CONSEQUENCES | v

TABLE OF CONTENTS 

1. PROJECT OVERVIEW...............................................................1

WHAT IS THE RESEARCH PROJECT’S PURPOSE?.............................................................2

HOW IS SOARCA STATE-OF-THE-ART? .............................................................................2

HOW DOES SOARCA DIFFER FROM PAST SEVERE ACCIDENT STUDIES?........................3

HOW ARE SEVERE ACCIDENTS AND POTENTIAL HEALTH 

CONSEQUENCES MODELED? ............................................................................................3

WHAT WERE THE STEPS OF THE PROJECT?.....................................................................4

HOW DOES NRC DETERMINE THE VALIDITY OF THIS STUDY?........................................5

2. PROGRESSION OF ACCIDENT SCENARIOS ...........................7

WHICH PLANTS DID SOARCA STUDY?.............................................................................8

WHAT ARE THE DIFFERENCES BETWEEN REACTOR TYPES?..........................................8

HOW WERE SCENARIOS SELECTED?................................................................................9

WHAT ACCIDENT SCENARIOS WERE ANALYZED?.........................................................10

HOW WERE THE ACCIDENTS MODELED?.......................................................................11

HOW LIKELY ARE THESE ACCIDENTS? ...........................................................................12

3. ACTIONS TO MITIGATE ACCIDENTS .....................................17

HOW CAN POTENTIAL ACCIDENTS BE MITIGATED?......................................................18

WHAT ASSURANCE DOES NRC HAVE THAT THESE MITIGATING ACTIONS 

CAN WORK?.....................................................................................................................18

HOW ARE MITIGATING ACTIONS MODELED?................................................................18

WHAT IS THE TIMING OF MITIGATING ACTIONS?.........................................................19

4. RELEASE OF RADIOACTIVE MATERIAL................................23

WHAT RADIOACTIVE MATERIAL DOES SOARCA MODEL? ............................................24

WHAT INFORMATION IS INCLUDED IN MELCOR MODELING?.......................................25

HOW ARE RADIOACTIVE MATERIALS MODELED TO ESCAPE FROM CONTAINMENT?.......25

5. MODELING EMERGENCY PLANS .........................................29

WHAT IS EMERGENCY PLANNING?................................................................................30

WHAT INFORMATION IS INCLUDED IN EMERGENCY PLAN MODELING?.....................31

WHAT DOES MODELING DEMONSTRATE ABOUT EMERGENCY PLANNING?...............31

6. MODELING HEALTH EFFECTS...............................................33

HOW ARE HEALTH CONSEQUENCES REPORTED IN SOARCA?......................................34

HOW ARE LONG-TERM CANCER FATALITY RISKS MODELED?......................................35

WHAT INFORMATION IS INCLUDED IN THE MACCS2 MODELING?...............................37

HOW ARE RADIOACTIVE MATERIALS MODELED TO MOVE DOWNWIND AND 

AFFECT THE POPULATION?..............................................................................................38
vi | U.S. Nuclear Regulatory Commission

7. RESULTS AND CONCLUSIONS..............................................39

WHAT ARE THE RESULTS OF THE MITIGATED SCENARIOS?..........................................40

HOW WOULD OPERATOR ACTIONS MITIGATE ACCIDENTS?.........................................40

WHAT ARE THE RESULTS OF UNMITIGATED SCENARIOS?...........................................40

WHAT DO SOARCA RESULTS INDICATE ABOUT CONSEQUENCES OF 

SEVERE ACCIDENTS?.......................................................................................................41

HOW DO SOARCA RESULTS COMPARE TO PAST STUDIES?..........................................43

HOW DO SOARCA RESULTS COMPARE TO THE NRC SAFETY GOAL 

AND OVERALL U.S. CANCER RISKS?...............................................................................43

WHAT ARE THE LIMITATIONS OF COMPARING SOARCA RESULTS TO THE 

NRC SAFETY GOAL AND OVERALL U.S. CANCER RISKS?..............................................43

GLOSSARY AND REFERENCES.................................................45

LIST OF FIGURES

1.1 Locations of All Operating Reactors in the United States..........................................4

1.2 Flow Chart of the SOARCA Process ............................................................................6

2.1 Peach Bottom Atomic Power Station and Surry Power Station..................................8

2.2 Typical U.S. Boiling-Water Reactor.............................................................................9

2.3 Typical U.S. Pressurized-Water Reactor....................................................................10

3.1 Comparison of SOARCA Accident Progression Timing for Mitigated and

Unmitigated Cases of Peach Bottom Long-Term Station Blackout...........................20

3.2 Comparison of SOARCA Accident Progression Timing for Mitigated and 

Unmitigated Cases of Surry Long-Term Station Blackout.........................................21

4.1 Percentages of Iodine and Cesium Released to the Environment During the

First 48 Hours of the Accident for SOARCA Unmitigated Scenarios, 1982 Siting

Study (SST1), and Historical Accidents.....................................................................26

5.1 Evacuation Timing for Peach Bottom Unmitigated Long-Term Station 

Blackout (hours) .........................................................................................................32

6.1 Transport Pathways of Radioactive Materials..........................................................36

6.2 Information Used to Model Health Effects in SOARCA ............................................37

7.1 Scenario-specific Risk of Dying from Long-Term Cancer for an Individual 

within 10 Miles of the Plant (per reactor-year) .........................................................42

LIST OF TABLES

2.1 Likelihoods of SOARCA Accident Scenarios.............................................................12

4.1 Timing and Quantity of Radioactive Material Released for SOARCA Mitigated

 and Unmitigated Scenarios.......................................................................................27

5.1 Evacuation Groups.....................................................................................................30

7.1 SOARCA Results: Mitigated and Unmitigated Cases...............................................41
MODELING POTENTIAL REACTOR ACCIDENT CONSEQUENCES | vii

ACRONYMS

ac alternating current

BWR boiling-water reactor

CDF core damage frequency

CFR Code of Federal Regulations

CRAC Calculation of Reactor Accident Consequences

dc direct current

EPA U.S. Environmental Protection Agency

EPZ emergency planning zone

ETE evacuation time estimate

HPS Health Physics Society

ISLOCA interfacing systems loss-of-coolant accident

LCF long-term cancer fatality

LNT linear no-threshold

LOCA loss-of-cooling accident

LTSBO long-term station blackout

LWR light-water reactor

MACCS2 MELCOR Accident Consequence Code System, Version 2

NRC U.S. Nuclear Regulatory Commission

PRA probabilistic risk assessment

psi pounds per square inch (pressure)

PWR pressurized-water reactor

SBO station blackout

SGTR steam generator tube rupture

SOARCA State-of-the-Art Reactor Consequence Analyses 

SST siting source term

STSBO short-term station blackout

TISGTR thermally induced steam generator tube rupture

MODELING POTENTIAL REACTOR ACCIDENT CONSEQUENCES | 1

This chapter explains the purpose of the project and 

the overall process for determining the results. 
2 | U.S. Nuclear Regulatory Commission

WHAT IS THE RESEARCH PROJECT’S PURPOSE?

The U.S. Nuclear Regulatory Commission’s (NRC’s) State-of-the-Art Reactor Consequence 

Analyses (SOARCA) research project calculated the realistic outcomes of severe nuclear 

power plant accidents that could release radioactive material into the environment. The 

computer models that produced these calculations incorporated decades of research into 

reactor accidents as well as the current design and operation of nuclear power plants. To 

provide perspective between SOARCA results and the more conservative estimates of 

severe reactor accident outcomes found in earlier NRC publications, SOARCA results are 

compared to the results of one of these previous publications: NUREG/CR-2239, “Technical 

Guidance for Siting Criteria Development,” commonly referred to as the 1982 Siting Study. 

The SOARCA report and this brochure help NRC to communicate severe-accident-related 

aspects of nuclear safety to you, the public; Federal, State, and local authorities; and 

nuclear power plant licensees.

HOW IS SOARCA STATE-OF-THE-ART?

NRC considers SOARCA a state-of-the-art project because (1) it models accidents 

with the latest plant-specific and site-specific information, (2) it uses an improved 

understanding of how radioactive material behaves during an accident, (3) it examines 

emergency response comprehensively, and (4) it combines modern computer-modeling 

capabilities and very detailed computerized plant models. 

NRC, the nuclear power industry, and international nuclear safety organizations have 

extensively researched plant response to potential accidents that could damage the 

reactor fuel and the containment building, which is designed to keep radioactive 

material from reaching the environment. This research has significantly improved NRC’s 

ability to develop computer models of how nuclear plant systems and operators would 

respond to severe accidents. When NRC developed the SOARCA plant models, the 

staff interviewed plant personnel and examined current plant equipment configurations 

to incorporate each facility’s most current design and operational information. This 

updated information includes:

 Plant owners improved plant safety through enhanced plant designs, emergency 

procedures, inspection programs, and operator training 

 Plant owners have also increased power production 

(referred to as “power uprates”) and lengthened 

operating times between replacing used fuel in the 

reactor – these actions changed the types and amounts 

of radioactive material in used reactor fuel.

 Plant owners improved severe accident mitigation 

strategies, including NRC-required enhancements made 

after the terrorist attacks of September 11, 2001, to 

respond to fires and explosions. These “10 CFR 50.54(hh) 

mitigation” enhancements are named after the relevant 

section of the NRC’s regulations.

 Plant owners and local governments have refined 

and improved emergency preparedness programs and 

equipment to further protect the public in the unlikely 

event of a severe accident.

How to Use this Brochure

This brochure provides tools to help understand SOARCA’s 

processes, terminology, and results. Here are some features 

that you can use:

•	Colored side boxes such as this one explain concepts, 

provide historical information, or explain relevant NRC 

regulations. 

•	Glossary in the appendix defines terms.

•	 References in the appendix provide a list of information 

documents.

If you are viewing this online:

•	Gray, underlined phrases and URLs are linked to the 

NRC Web site.
MODELING POTENTIAL REACTOR ACCIDENT CONSEQUENCES | 3

All of these changes have been considered in SOARCA. 

The SOARCA team applied this accumulated research and 

incorporated plant changes to more realistically evaluate 

the potential health consequences from severe nuclear 

reactor accidents. 

HOW DOES SOARCA DIFFER FROM PAST 

SEVERE ACCIDENT STUDIES?

NRC has previously researched the probabilities and 

potential health consequences of severe accidents and 

documented this research in reports such as WASH-1400, 

”Reactor Safety Study. An Assessment of Accident Risks 

in U.S. Commercial Nuclear Power Plants”, NUREG-1150, 

“Severe Accident Risks: An Assessment for Five U.S. 

Nuclear Power Plants”, and NUREG/CR-2239. The SOARCA 

Report, NUREG-1935, contains details about some of these 

past studies. Since the publication of the earlier studies, 

NRC has participated in many severe accident research 

programs. This work has improved our understanding of 

how heat is transferred and radioactive material moves 

through reactor systems during severe accidents and how 

radioactive material might get out of the containment 

building and move through the surrounding environment. 

NRC incorporated these research results into SOARCA’s 

computer codes. In addition, the SOARCA study used 

a more complete and detailed computer model of the 

reactor, containment, and other buildings onsite. Because 

SOARCA is based on decades of research and uses 

improved modeling tools, the study generates more 

realistic results than past efforts such as the 1982 Siting 

Study. These past studies were based on then-existing 

plant descriptions and knowledge of how severe accidents 

would occur. However, we now know that the predictions 

from these past studies are out of date for realistically 

understanding severe accident consequences. 

HOW ARE SEVERE ACCIDENTS AND 

POTENTIAL HEALTH CONSEQUENCES 

MODELED?

The SOARCA project used sophisticated computer 

programs to calculate the effect a severe accident 

could have on an operating nuclear reactor and the 

possible impact on the public. These programs integrate 

information about reactor systems, components, 

operating history, and the impacts of emergency 

procedures, weather conditions, emergency planning, 

evacuation time estimates, and population. 

What Is a Severe Accident?

A severe accident is a type of accident that may challenge 

safety systems at a level much higher than expected.

A reactor accident occurs when the plant cooling water 

systems are no longer removing heat from the reactor fuel 

(the “core” of the reactor). Extensive core damage could 

melt reactor fuel, which would settle at the bottom of 

the reactor vessel that is designed to hold the fuel. The 

reactor vessel is surrounded by the containment building. 

If cooling water is not restored, however, and the accident 

progresses further, the melted fuel could rupture the bottom 

of the reactor vessel, with the melted fuel flowing onto the 

containment floor. Radioactive material would be released 

from the fuel into the containment atmosphere and could 

potentially escape containment if there were any available 

leakage paths.

Who Is the Project Team?

The project team included engineers and scientists from 

NRC and two contractors, Sandia National Laboratories and 

dycoda, LLC. The team’s expertise included probabilistic 

risk assessment, heat transfer and fluid flow, emergency 

response, atmospheric dispersion, and radiation health 

effects. Team members focused their technical expertise 

on creating and applying detailed computer models to help 

determine realistic consequences of severe nuclear power 

plant accidents.
4 | U.S. Nuclear Regulatory Commission

WHAT WERE THE STEPS OF THE PROJECT?

The SOARCA project took a step-by-step approach to calculate 

the potential consequences of the analyzed severe accidents. 

The project team first decided it could learn more by rigorously 

and realistically analyzing a relatively small number of important 

accident scenarios, rather than carrying out less-detailed modeling 

of many scenarios. Therefore, the team selected a threshold 

to help select scenarios to analyze (Chapter 2 of this brochure 

describes the selection process). SOARCA aimed to assess the 

benefits of 10 CFR 50.54(hh) mitigation measures (put in place 

after 9-11 for responding to fires and explosions) in other accident 

scenarios. We also wanted to provide a basis for comparison to 

past analyses of severe accident scenarios before these mitigation 

measures existed. The project therefore analyzed the selected 

scenarios twice: first assuming that the event proceeds without 

the 10 CFR 50.54(hh) mitigation measures, called “unmitigated” and then assuming 

that the 10 CFR 50.54(hh) mitigation is successful, called “mitigated”. For scenarios 

leading to an offsite release of radioactive material, SOARCA then analyzed the 

material’s atmospheric dispersion, the surrounding area’s emergency response, and 

potential health consequences. Figure 1.2 illustrates this overall approach.

Figure 18. U.S. Operating Commercial Nuclear Power Reactors

PA

NY

ME

IN

WA

OR

CA

NV

MT

ID

UT

AZ NM 

CO

WY

TX

OK

KS

NE

SD

ND

MN

IA

MO

AR

LA

MS AL GA

FL

SC

NC

VA

TN

KY

IL

OH

MI

WI

WV

MD

DE

NJ

CT

RI

MA

NH

VT

AK

HI

Licensed to Operate (104)

REGION I

CONNECTICUT

 Millstone 2 and 3

REGION II

ALABAMA

 Browns Ferry 1, 2, 

and 3

REGION III

ILLINOIS

 Braidwood 1 and 2

Byron 1 and 2

REGION IV

ARKANSAS 

 Arkansas Nuclear 1 

and 2

Peach

Bottom

Surry

Figure 1.1 Locations of All Operating Reactors in the United States.

What Are NRC Regulations?

NRC works diligently to ensure safe operation 

of nuclear power plants, supporting safety by 

developing rules for the proper operation of a 

nuclear power plant. These rules are detailed in 

Title 10 of the Code of Federal Regulations (10 

CFR). Throughout this brochure, we will refer you 

to some of the relevant rules so you can better 

understand how NRC works to protect public 

health and the environment. An online version 

of 10 CFR is available at http://www.nrc.gov/

reading-rm/doc-collections/cfr/. 
MODELING POTENTIAL REACTOR ACCIDENT CONSEQUENCES | 5

HOW DOES NRC DETERMINE THE 

VALIDITY OF THIS STUDY?

Peer Review— A peer review is a review of a research 

project by experts not involved in the project.These 

experts examine the methods and results of the research 

and help improve the work by identifying the project’s 

strengths and areas for improvement. The SOARCA team 

assembled a panel of independent, external experts 

in the fields of risk analysis, severe accident research, 

emergency preparedness, and radiation health effects. 

This group reviewed SOARCA’s methodology, underlying 

assumptions, results, and conclusions to ensure that 

they are technically sound and state-of-the-art. For the 

same reasons, NRC’s Advisory Committee on Reactor 

Safeguards (a standing group of nuclear safety experts) 

also reviewed the project and provided comments. The 

SOARCA team has incorporated the experts’ feedback 

into the reports. 

Sensitivity and Uncertainty Analyses—Scientific 

research strives for valid results based on high-quality 

data and reasonable assumptions. Because data can 

be sparse and uncertain, however, researchers work 

systematically to identify any weaknesses in data and 

assumptions and to consider alternatives. This step is an 

important part of making research results transparent and 

understandable. NRC staff used sensitivity analyses to 

compare how varying individual input assumptions affect 

the outcomes. The results of these sensitivity analyses 

show that the SOARCA results are reasonable considering known uncertainties. In 

addition, NRC is taking a systematic look at potential sources of uncertainty and their 

impact on SOARCA results in a separate uncertainty analysis. The uncertainty analysis 

uses a statistical approach to assess the uncertainties in a more integrated fashion.

What Computer Codes Were Used for 

SOARCA?

SOARCA uses two specialized computer codes to analyze 

severe accidents and offsite consequences. The first, 

MELCOR, calculates accident timing and event progression 

using plant design information and models for the accident 

phenomena. The second, MACCS2 (MELCOR Accident 

Consequence Code System, Version 2), calculates the offsite 

health consequences of an airborne release of radioactive 

material using site-specific information for the area and 

radiological release data from MELCOR.

The MELCOR code was peer reviewed in 1991 by experts 

from national laboratories, universities, and MELCOR code 

users. This peer review provided an independent assessment 

of the technical adequacy of the code. The peer reviewers’ 

recommendations were incorporated into NRC’s research 

and development plan for the code, which has also been 

checked, or “validated”, against numerous experimental 

results over the past several decades.

An expert panel review of the MACCS2 code and SOARCA’s 

MACCS2 modeling choices was conducted in August 2006, 

prior to the start of specific work as part of the Peach Bottom 

and Surry analyses. This expert panel review and the NRC 

staff recommendations influenced much of the development 

that has been undertaken specifically to support SOARCA.
6 | U.S. Nuclear Regulatory Commission

Select accident 

scenarios to

analyze 

Model 

progression of 

accident

Model 

emergency

planning 

Model dispersion 

of radioactive 

material 

Model health

consequences 

No health 

consequences to 

model

If no core 

damage, no 

release 

If core damage, 

and containment

does not prevent 

release 

If core damage, 

but containment 

prevents release 

Model mitigative 

measures 

Figure 1.2 Flow Chart of the SOARCA Process.
This chapter explains the basic information on reactor 

designs and how accident scenarios could lead to 

damage of the reactor core.

MODELING POTENTIAL REACTOR ACCIDENT CONSEQUENCES | 7
8 | U.S. Nuclear Regulatory Commission

WHICH PLANTS DID SOARCA STUDY?

SOARCA analyzed an example of each major type of operating U.S. nuclear reactor: a 

boiling-water reactor (BWR) and a pressurized-water reactor (PWR). The project team 

solicited volunteers from the nuclear industry to participate in the project. The Peach 

Bottom Atomic Power Station in Pennsylvania and the Surry Power Station in Virginia 

volunteered and are the focus of this report. Peach Bottom is a General Electricdesigned BWR with a Mark I containment. Surry is a Westinghouse-designed PWR 

with a large dry containment. These two plants, depicted in Figure 2.1, also were part 

of earlier studies. 

WHAT ARE THE DIFFERENCES BETWEEN REACTOR TYPES?

Figures 2.2 and 2.3 describe some of the major differences between BWRs and PWRs. 

Within these two general types of U.S. commercial nuclear reactors, many variations 

exist in the design of systems, components, and containments at different sites. 

Figure 2.1 Peach Bottom Atomic Power Station 

(top) and Surry Power Station (bottom).
MODELING POTENTIAL REACTOR ACCIDENT CONSEQUENCES | 9

HOW WERE SCENARIOS SELECTED?

The project team sought to focus its attention and resources on the important severe 

accident scenarios for Peach Bottom and Surry found in past risk studies, such as 

NUREG-1150, “Severe Accident Risks: An Assessment for Five U.S. Nuclear Power 

Plants”. The project narrowed its approach by using an accident sequence’s possibility of 

damaging reactor fuel (also called the reactor “core”), or core damage frequency (CDF), 

as an indicator of risk. 

The SOARCA scenarios were selected from the results of existing probabilistic risk 

assessments (PRAs). The scenario-selection process used updated and benchmarked 

standardized plant analysis risk (SPAR) models and available plant-specific external 

events information from 2005. Core damage scenarios from previous staff and 

licensee PRAs were identified and combined into common core damage groups 

that have similar timing and response for important severe accident phenomena and 

similar containment or safety systems. The groups were screened according to their 

approximate CDFs to identify the most risk-significant groups. SOARCA analyzed 

scenarios with a CDF equal to or greater than 1 in a million reactor-years. SOARCA also 

Figure 2.2 Typical U.S. Boiling-Water Reactor.

A BWR cools the reactor core and 

generates steam to turn a turbine 

using a single loop of water, as 

distinct from a PWR (see figure 

2.3) that has separate loops for 

cooling the reactor and generating 

steam. Heat from nuclear fission 

in the reactor core converts the 

water to steam. The steam travels 

through the steam line to the 

turbine generator where it turns 

the generator to make electricity. 

The steam then enters the 

condenser where it is cooled back 

into liquid water and is pumped 

back into the reactor to repeat 

the process. The BWR’s water is 

pressurized to about 1,100 pounds 

per square inch (psi) pressure so 

it boils at about 550 °F. A typical 

BWR core contains between 400 

and 800 fuel assemblies, and each 

fuel assembly holds 75 to 100 

fuel rods. The BWR in this figure 

is shown with a Mark I style of 

containment. More information is 

available at http://www.nrc.gov/

reactors/bwrs.html

Reactor

 Vessel

Reactor Building

Suppression

Pool

Suppression

Pool

Saf

Injec

Syst

Primary 

Containment

Inerted Drywell

Steel Shell

Wa

Sto

Ta

Control Rods

Reactor

Core

Steam to Turbine

Water from Condenser

NOT TO SCALE 
10 | U.S. Nuclear Regulatory Commission

Figure 2.3 Typical U.S. Pressurized-Water Reactor.

A PWR has separate coolant loops to cool the 

reactor and generate the steam. The PWR’s 

coolant loop (known as the primary loop) is 

under very high pressure (about 2,300 psi) 

to prevent water from boiling. The water is 

pumped through the reactor core where it is 

heated to about 600°F before being routed 

to the steam generators. The water travels 

through thousands of small tubes inside the 

steam generators where it heats secondary 

loop water at a lower pressure (about 900 psi) 

to produce saturated steam at about 530°F. 

This steam enters the main steam line that 

routes it to the turbine generator. From the 

turbine generator, the steam then enters the 

condenser that cools it back to water so it can 

be pumped back to the steam generator to 

repeat the cycle. A typical PWR core has 150 

to 250 fuel assemblies, and each assembly 

contains 200 to 300 fuel rods in a 14x14 to 

17x17 matrix. Each PWR reactor has 2, 3, 

or 4 steam generators connected to it. The 

PWR in this figure is shown with a large dry 

containment. More information is available at 

http://www.nrc.gov/reactors/pwrs.html.

Reactor

Cavity

Accumulator

Reactor

 Vessel

Reactor Building

Suppression

Pool

Reactor

 Vessel

Steam

Generator

Safety

Injection

System

Reinforced 

Concrete 

Containment

Building

nt

well

l

Water 

Storage

Tank

Control

Rods

Control Rods

Reactor

Core

Reactor

Core

Containment

Spray

Steam to Turbine

Steam to Turbine

Water from Condenser

Water from Condenser

NOT TO SCALE NOT TO SCALE

sought to analyze scenarios leading to an early failure or 

bypass of the containment where the CDF is equal to 

or greater than 1 in 10 million reactor-years, since these 

scenarios have a potential for higher consequences and 

risk. This approach allowed a more detailed analysis of 

accident consequences for the more likely, although still 

remote, accident scenarios.

WHAT ACCIDENT SCENARIOS WERE 

ANALYZED?

For both Peach Bottom and Surry, the team modeled 

loss of all alternating current (ac) electrical power or 

“station blackout (SBO)” scenarios caused by earthquakes 

more severe than anticipated in the plant’s design. SBO 

frequencies from flood or fire scenarios were combined 

with the earthquake frequency for scenario selection; 

however, SOARCA modeled the earthquake scenario 

What is a Probabilistic Risk Assessment?

Probabilistic risk assessment (PRA) is an engineering 

approach to systematically identify potential nuclear power 

plant accident scenarios and estimate their likelihoods of 

occurrence and consequences. Each accident scenario begins 

with an initiating event (such as a loss of offsite power or 

earthquake) followed by a combination of equipment failures 

and operator actions that can lead to core damage and the 

release of radioactive materials from the containment. The 

information developed by a PRA is useful in identifying plant 

vulnerabilities. Pioneered by NRC in the 1970s, PRA has 

been adopted by nuclear power plant operators and regulators 

worldwide as a tool that complements other approaches to 

assess nuclear power plant safety. 
MODELING POTENTIAL REACTOR ACCIDENT CONSEQUENCES | 11

because this presented the most severe challenge to the 

plant operators as well as offsite emergency responders, 

and had the highest probability of occurring. 

Long-Term Station Blackout (LTSBO)—In this scenario, 

the plant loses all ac power sources, but battery backups 

operate safety systems for about 4-8 hours until the 

batteries are exhausted. 

Short-Term Station Blackout (STSBO)—In this 

scenario, the site loses all power (even the batteries), all 

of the safety systems immediately become inoperable, 

and core damage occurs in the “short term.”1 The STSBO 

scenario starts with a more extreme earthquake than the 

one that starts the LTSBO. 

In addition, the team analyzed two scenarios for Surry 

in which radioactive material could potentially reach the 

environment by bypassing containment features. These 

are discussed in more detail in Chapter 4. 

Interfacing Systems Loss-of-Coolant Accident (ISLOCA)—In this scenario, a random 

failure of valves ruptures low-pressure system piping outside containment that connects 

with the high-pressure reactor system inside containment. 

Thermally Induced Steam Generator Tube Rupture (TISGTR)—This scenario is a 

lower probability variation of the STSBO. While the core is overheating and boiling off 

the available water, extremely hot steam and hydrogen circulating through the steam 

generator rupture a steam generator tube resulting in a pathway for radioactive material 

to escape to the non-radioactive portion of the plant and potentially to the environment.

Peach Bottom and Surry both have two reactor units on the site. Multiunit accidents 

(events leading to reactor core damage at multiple units on the same site) could be 

caused by certain initiators such as an earthquake. Most PRAs and health consequence 

studies developed to date do not explicitly consider multiunit accidents because NRC 

policy is to apply the Commission’s “Safety Goals for the Operation of Nuclear Power 

Plants” (51 FR 28044) and subsidiary risk acceptance guidelines on a “per reactor” 

basis. Therefore only single-reactor accidents were evaluated in SOARCA.

HOW WERE THE ACCIDENTS MODELED?

The SOARCA team modeled the accident scenarios and their potential to damage the 

core as realistically as possible by gathering detailed information about each of the two 

plants studied. The team asked plant staff for specific information about the design 

and operation of each plant system. The models’ realism is enhanced by incorporating 

recent U.S. and international research about severe accidents and accounting for 

additional structures within containment (such as internal walls, piping, pumps, and 

heat exchangers) and buildings adjacent to the containment.

The state-of-the-art MELCOR computer code modeled how each scenario would unfold 

at each plant. The MELCOR results describe the following:

1 This terminology for long-term SBO and short-term SBO is consistent with that used in past NRC studies 

including NUREG-1150, “Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants”.

What is a Station Blackout?

Reactor cooling systems at nuclear power plants are 

powered by alternating current (ac) power. This ac power 

is normally supplied by offsite power sources via the 

electrical grid but can be supplied by onsite sources such as 

emergency diesel generators if needed. A station blackout 

(SBO) involves the total loss of ac power when both offsite 

and onsite ac power sources fail. During an SBO, reactor 

cooling is temporarily provided by systems that do not 

rely on ac power, such as turbine-driven pumps that are 

driven by steam from the reactor. Batteries also are used to 

provide direct current (dc) power to control the turbinedriven pumps and to power instrumentation. Historically, 

risk models have indicated that the station blackout is an 

important contributor to overall nuclear power plant risk. 
12 | U.S. Nuclear Regulatory Commission

Table 2.1 Likelihoods of SOARCA Accident Scenarios

Reactor Site Accident Scenario Probability/Frequency

Surry Long-Term Station Blackout 1 event in ~ 50,000 years

Peach Bottom Long-Term Station Blackout 1 event in ~ 300,000 years

Surry Short-Term Station Blackout 1 event in ~ 500,000 years

Surry Short-Term Station Blackout with Thermally 

Induced Steam Generator Tube Rupture 1 event in ~ 3 million years

Peach Bottom Short-Term Station Blackout 1 event in ~ 3 million years

Surry Interfacing Systems Loss-of-Coolant Accident 1 event in ~ 30 million years

 How the plant and its emergency systems perform in response to an accident. 

 How the reactor core behaves as it heats up beyond normal temperature limits.

 How the fuel itself, the reactor piping, and the containment building behave under 

extremely high temperatures.

 Whether radioactive material reaches the environment and, if so, how it occurs and 

how much material is released.

This information is based on the plant’s design and physical safety systems. In addition, 

nuclear plants have a series of redundant and diverse safety measures to back up the 

designed safety systems. Chapter 3 of this brochure discusses how the SOARCA project 

models the actions that can potentially prevent or mitigate the release of radioactive 

material and ultimately protect the public. If a scenario caused a release of radioactive 

material, the team used another computer code (MACCS2) to calculate the offsite health 

consequences of the release; Chapters 4, 5, and 6 provide more details about this step.

HOW LIKELY ARE THESE ACCIDENTS? 

Overall, the SOARCA scenarios have core damage likelihoods that range from about 1 

accident in 50,000 years to 1 accident in 30 million years. Table 2.1 shows the likelihoods 

for each scenario in order of more likely scenarios to less likely scenarios. Although the 

chances of these scenarios ever occurring are very small, probabilistic risk assessments 

have shown that these scenarios are very important core damage sequences. 

SOARCA examines the effectiveness of actions to mitigate each accident (should 

one occur) and to prevent radioactive material from reaching the public and the 

environment. The likelihoods of the scenarios selected for SOARCA were based on: a 

review of NUREG-1150; the Individual Plant Examination of External Events (IPEEEs) 

conducted by licensees in the 1990s; NRC-developed SPAR models of external events; 

licensee-sponsored PRAs; and other NRC-sponsored studies. There was no attempt to 

match the stated likelihoods to any one particular study. Rather, they reflect the expert 

opinion of the NRC staff, based on all these sources of information available in 2005 

when the scenarios were selected. Updated information could affect these estimates. 

For example, NRC staff expects to gain further insight into seismic and flooding event 

scenarios when U.S. nuclear power plants implement recommendations from the 

Fukushima Near-Term Task Force report (July 2011). 
MODELING POTENTIAL REACTOR ACCIDENT CONSEQUENCES | 13

Historical Perspective: Three Mile Island and Chernobyl

Many people are familiar with the Three Mile Island (pictured left) and Chernobyl (pictured right) accidents. Although 

SOARCA did not examine these historical accidents, this brochure provides information about them so readers can compare 

the results of this research study to real events. 

On March 28, 1979, the Three Mile Island accident occurred in Pennsylvania as a result of equipment 

malfunctions, design-related problems, and worker errors. The accident melted almost half the reactor core of 

Unit 2 and released contaminated water and radioactive material into the containment building. A very small 

amount of radioactive material reached the environment. It remains the most serious accident in U.S. commercial 

nuclear power plant operating history although no plant workers or members of the nearby community were 

injured or killed. A long-term follow-up study by the University of Pittsburgh that evaluated local, county, and State 

population data from 1979 through 1998 concluded that there is not an increase in overall cancer deaths among the 

people living within a 5-mile radius of Three Mile Island at the time of the accident. This accident brought about 

sweeping changes for nuclear power plants and heightened oversight by NRC. 

On April 26, 1986, an accident destroyed Unit 4 of the nuclear power station at Chernobyl, Ukraine, in the 

former USSR. The series of events that led to this accident could not occur at U.S. commercial power reactors 

because U.S. reactors have different plant designs, robust containment structures, and operational controls 

to protect them against the combination of lapses that led to the accident at Chernobyl. Its operators ran an 

experiment that led to a sudden surge of power, destroying the reactor core and releasing massive amounts of 

radioactive material into the environment. About 30 emergency responders died in the first 4 months after the 

accident. The health of the evacuated population and populations in contaminated areas of Belarus, the Russian 

Federation, and Ukraine has been monitored since 1986. Monitoring efforts to date indicate that a lack of prompt 

countermeasures resulted in increased risk of thyroid cancer to members of the public, most notably among 

people who were children or young adults at the time of the accident. No other health effects are attributed to 

the radiological exposure in the general population. Chernobyl’s design, which differed significantly from reactors 

operating in the United States, made it vulnerable to such a severe accident.

NRC Fact Sheets about Three Mile Island and Chernobyl Accidents are available at:

•	http://www.nrc.gov/reading-rm/doc-collections/fact-sheets/3mile-isle.html

•	http://www.nrc.gov/reading-rm/doc-collections/fact-sheets/chernobyl-bg.html.
14 | U.S. Nuclear Regulatory Commission

Historical Perspective: Fukushima Dai-ichi and NRC Response

On March 11, 2011, a 9.0-magnitude earthquake struck Japan about 231 miles northeast of Tokyo off the east coast of Honshu 

Island. The earthquake led to the automatic shutdown of 11 reactors at 4 sites (Onagawa, Fukushima Dai-ichi, Fukushima 

Dai-ni, and Tokai). At Fukushima Dai-ichi, which includes General Electric BWR Mark I reactors similar to the Peach Bottom 

plants, diesel generators provided electricity to plant systems until about 40 minutes later. At that point, a tsunami, estimated 

to have exceeded 45 feet (14 meters) in height, appeared to have caused the loss of all alternating current (ac) power and 

most emergency diesel generators to the six Fukushima Dai-ichi reactors. Three Fukushima Dai-ichi reactors (Units 1-3) were 

in operation at the time of the earthquake, and three (Units 4-6) were shut down for routine refueling and maintenance. Due 

to lack of ac power to pump water into Units 1 through 3 to cool the nuclear fuel, some of the fuel melted. The melted fuel 

cladding reacted with steam and generated hydrogen gas. The hydrogen reached critical levels and caused explosions. The 

reactor damage, along with hydrogen gas explosions inside the units, released radioactive material into the environment. The 

earthquake and tsunami devastation in the area significantly delayed offsite assistance. Additional systems were finally able to 

use seawater to cool the reactors, and Japan continues work on stabilizing these plants.

Since the events at Fukushima began to unfold, NRC has been working to understand the events in Japan and relay important 

information to U.S. nuclear power plants. Not long after the emergency began, NRC established a task force of senior NRC 

experts to determine lessons learned from the accident and to initiate a review of NRC regulations to determine if additional 

measures should be taken immediately to ensure the safety of U.S. nuclear power plants. The task force issued its report on 

July 12, 2011, concluding that continued U.S. plant operation and NRC licensing activities presented no imminent risk. The 

Task Force also concluded that enhancements to safety and emergency preparedness are warranted and made a dozen general 

recommendations for Commission consideration. The NRC is currently implementing many of those recommendations to 

enhance U.S. nuclear plant safety.

An appendix to the main SOARCA report briefly compares and contrasts what we currently know about Fukushima with 

insights from the Peach Bottom SOARCA analyses. The NRC Web site has additional information on the Fukushima accident 

and NRC’s response:

•	 http://www.nrc.gov/reading-rm/doc-collections/fact-sheets/fs-japan-events.html

•	 http://www.nrc.gov/japan/japan-info.html

Fukushima Dai-ichi site prior to the March 11, 2011, earthquake and tsunami.
MODELING POTENTIAL REACTOR ACCIDENT CONSEQUENCES | 15

Comparison of Fukushima Accident to SOARCA Analyses

The SOARCA study had nearly completed its peer review when the Fukushima Dai-ichi accident occurred on March 

11, 2011. Following the accident, the U.S. Department of Energy (DOE) and the NRC began a cooperative effort to 

use the MELCOR code for a forensic analysis of event progression to develop a more detailed understanding of the 

accident. This cooperative effort is ongoing.

Based on limited information currently available, the Fukushima accident is in some ways similar to a few of the 

Peach Bottom scenarios analyzed in SOARCA. The SOARCA team compared and contrasted the Fukushima accident 

and the SOARCA study for the following topics: (1) operation of the reactor core isolation cooling (RCIC) system, 

(2) hydrogen release and combustion, (3) 48-hour truncation of releases in SOARCA, (4) multiunit risk, and (5) 

spent fuel pool risk. It must be emphasized that we need much more information to be certain about what actually 

occurred in the Fukushima reactors. Our current uncertainty prevents us from drawing firm conclusions regarding 

comparisons with SOARCA results. 

As the NRC learned more about the damage to plant safety functions gathered over the weeks and months 

following these events, many similarities became apparent between SOARCA’s calculated damage progression in the 

Peach Bottom SBO accident scenarios and the progression of events at Fukushima. These similarities include the 

following:

•	the sequence and timing of events that followed the loss of core cooling, including the start of core damage and 

radioactive material release from the fuel, 

•	 challenges to containment integrity from the loss of fuel heat removal and the accumulation of hydrogen 

generated during fuel damage within the reactor vessel, and

•	the destructive effects of hydrogen combustion in the reactor building.

Some notable differences were also obvious between the events that unfolded at Fukushima and the Peach Bottom 

LTSBO scenario studied in the SOARCA project. These differences, for example the use and timing of certain safety 

systems, led the NRC staff to take a closer look at the models used and assumptions made in the LTSBO analyses. 

SOARCA analysis results were qualitatively compared to the preliminary events and information available in the 

evaluation of the Fukushima Dai-ichi accident. SOARCA’s conclusions remain valid in light of information currently 

available from the events that unfolded at Fukushima. 

MODELING POTENTIAL REACTOR ACCIDENT CONSEQUENCES | Modeling Potential Reactor Accident Consequences | 17

 This chapter explains the basic information 

on operator actions that mitigate the effect 

of accidents by preventing core damage or 

preventing, delaying, or reducing release of 

radioactive materials.
18 | U.S. Nuclear Regulatory Commission

Defense-in-Depth Philosophy 

“Defense in depth” is NRC’s approach to 

designing and operating nuclear facilities to 

prevent and mitigate accidents that could release 

radioactive materials. The key is creating multiple 

independent and redundant layers of defense to 

compensate for potential human and mechanical 

failures so that no single layer, no matter how 

robust, is exclusively relied upon. Defense-indepth includes the use of redundant and diverse 

key safety functions and emergency response 

measures. For further information, see Speech No. 

S-04-009, “The Very Best-Laid Plans (the NRC’s 

Defense-in Depth Philosophy).”

HOW CAN POTENTIAL ACCIDENTS BE 

MITIGATED?

In addition to the redundant and diverse physical systems 

designed to prevent accidents, NRC and plant owners understand 

the importance of having preplanned emergency measures in the 

unlikely event an accident occurs. NRC expects these emergency 

measures will mitigate accident consequences by preventing 

core damage or preventing, delaying, or reducing the release of 

radioactive material. NRC requires plant operators to maintain 

detailed emergency procedure plans for the entire range of 

possible accidents. These plans include the following:

Emergency operating procedures—These procedures 

list operator actions to mitigate possible nuclear power plant 

emergencies.

Severe accident management guidelines—These are operator 

guidelines to mitigate accidents that are more severe than what 

the facility was designed to handle.

Security-related (10 CFR 50.54(hh)) mitigation measures—These measures 

include plans and resources that nuclear plants put in place to meet additional NRC 

requirements following the terrorist attacks on September 11, 2001 to mitigate 

scenarios involving the loss of large areas of the plant due to fire or explosions.

WHAT ASSURANCE DOES NRC HAVE THAT THESE MITIGATING 

ACTIONS CAN WORK?

NRC requires its licensees to train and practice emergency operating procedures 

in simulators that replicate the plant control rooms at each site. NRC also requires 

that plant owners have developed severe accident management guidelines and 

implemented the security-related mitigation measures to ensure that they have proper 

equipment, procedures, and training. NRC inspectors observe these activities to 

ensure NRC regulations are met at each plant. 

HOW ARE MITIGATING ACTIONS MODELED?

SOARCA is the first detailed analysis that quantifies the value of the security-related 

10 CFR 50.54(hh) mitigating actions in responding to potential accident conditions. This 

equipment and procedures were intended to be used to maintain or restore safety 

functions under circumstances associated with loss of large areas of the plant due to 

explosions or fire. The NRC anticipates, however, that plant operators could use this 

equipment for other types of accidents. 

Therefore, for each plant, two cases of each scenario are modeled. 

Mitigated Case—In the first case, the SOARCA team modeled what would happen 

if the operators are fully successful in carrying out the mitigating actions. The project 

team accomplished this by holding tabletop exercises with senior reactor operators 

and emergency response personnel at Peach Bottom and Surry to determine what 

actions would be taken to mitigate each scenario analyzed including the time required to 

implement each action. Many of these actions are designed to help in the case of large 

fires and explosions but could potentially be used for the scenarios analyzed in SOARCA.
MODELING POTENTIAL REACTOR ACCIDENT CONSEQUENCES | 19

Unmitigated Case—To understand the value of 10 CFR 50.54(hh) mitigating actions 

and to provide a basis for comparing SOARCA results to past studies, the team 

also analyzed an “unmitigated case” for each scenario. These unmitigated cases 

assumed that the plant failed to implement 10 CFR 50.54(hh) measures. Although the 

earthquakes considered in the SOARCA scenarios exceed the plants’ designs, the more 

rugged engineered safety features are assumed to survive in both the unmitigated and 

mitigated cases. These actions respond to design-basis events too, for which operators 

have more specific procedures and frequent training. The unmitigated cases modeled 

the sequence of events that lead to fuel damage, release of radioactive materials, and 

offsite health consequences.

WHAT IS THE TIMING OF 

MITIGATING ACTIONS?

Detailed MELCOR modeling 

demonstrated that plant operators 

can have time during accident 

scenarios to perform the necessary 

emergency actions. Figures 3.1 

and 3.2 compare the mitigated and 

unmitigated timelines for the Peach 

Bottom and Surry long-term station 

blackout scenarios until the release 

starts (for the unmitigated case).

Historical Perspective: How Have Mitigation 

Capabilities Improved Since 9/11?

In response to the terrorist attacks of September 11, 2001, NRC and operating reactor 

licensees worked together to develop improved mitigation methods for events that 

could disable large areas of a nuclear power plant. As a result, operating reactor 

licensees purchased equipment and developed procedures for each site to better 

mitigate such events. NRC codified the requirements for this additional mitigation in 

Title 10 CFR 50.54 (hh). These mitigation measures include the following for the two 

plants analyzed in SOARCA:

•	 Portable diesel-fuel powered pumps (pictured).

•	 Portable generators to provide electricity to power critical instrumentation and to 

open or close valves.

•	 Prestaged air bottles to open or close air-operated valves.

•	 Procedures for operating steam-turbine-driven pumps without power.

•	Designated make-up water sources.

PRAs commonly include a human reliability analysis to represent the likelihood of 

operator actions. SOARCA evaluated human actions through tabletop exercises, 

walkdowns, simulator runs, and other inputs from licensee staff.
20 | U.S. Nuclear Regulatory Commission

0

4

5

13

20

Station blackout

Operators position, connect and 

start alternate electricity

Operators manually control the 

cooling water flow (by the 4th hour)

Operators align and start 

portable pumps 

(from the 4th to the 10th hour)

Accident Mitigated - No Release

Station blackout

Backup batteries deplete

Reactor coolant flow stops

Lower head of reactor dries out

Lower head of reactor and 

containment fail

Release of radioactive material starts

Mitigated Case Hours Unmitigated Case

Figure 3.1 Comparison of SOARCA Accident Progression Timing for Mitigated 

and Unmitigated Cases of Peach Bottom Long-Term Station Blackout.
MODELING POTENTIAL REACTOR ACCIDENT CONSEQUENCES | 21

0

4

5

8

21

Station blackout

Operators start to align emergency generators and pumps

Emergency diesel injection starts into the reactor coolant system

Refill emergency condensate storage tank

Accident Mitigated - No Release

Station blackout

Steam generator cooling flow stops

Core damage starts

Reactor vessel lower head fails

Containment fails and release of

radioactive material to the

environment starts

Mitigated Scenario Hours Unmitigated Scenario

Backup batteries deplete

16

45

Figure 3.2 Comparison of SOARCA Accident Progression Timing for Mitigated 

and Unmitigated Cases of Surry Long-Term Station Blackout.

Modeling Potential Reactor Accident Consequences | 23

Modeling Hypothetical Accidents at Nuclear Power Plants | 23

This chapter explains how the project modeled the 

release of radioactive material and what information 

is used in the calculations.

MODELING POTENTIAL REACTOR ACCIDENT CONSEQUENCES | 23
24 | U.S. Nuclear Regulatory Commission

The SOARCA models showed that mitigating actions can prevent core damage or 

reduce or delay a release of radioactive material. For the scenarios examined, the 

SOARCA team also modeled unmitigated cases that lead ultimately to a release to 

the environment. The MELCOR computer code models the behavior of radioactive 

materials to the point that they escape from containment. 

WHAT RADIOACTIVE MATERIAL DOES SOARCA MODEL? 

SOARCA took a detailed approach to considering radioactive substances, or 

radionuclides. In SOARCA, MELCOR calculations of reactor accident response are based 

on realistic estimates of decay heat generated by the radionuclides in the reactor core. 

MELCOR organizes the radionuclides by chemical similarity to track them as they are 

released from the reactor core and move through piping, the containment building, and 

other buildings on their way to the environment. The offsite consequences computer 

code (MACCS2) tracks radionuclides based on how long they remain radioactive, their 

biological importance, and how much is expected to be released from the core. 

How Does Containment Work? 

As part of the defense-in-depth philosophy, NRC requires all currently operating reactors to have three physical barriers that 

protect the public and environment from potential releases of radioactive material: 

Containment Building—enclosure around a nuclear reactor to confine radioactive material that otherwise might be 

released to the atmosphere in the event of an accident. 

Reactor Vessel— metal enclosure that holds the reactor core and the cooling water. 

Fuel Rods—long, slender tubes that hold uranium fuel for nuclear reactor 

use. Fuel rods are assembled into bundles that are loaded individually into 

the reactor core (see image below).

Note: Typical large dry containment shown. Surry has a large dry, 

containment where the interior pressure is kept lower than atmospheric 

pressure. Surry’s containment is expected to perform similarly during a 

severe reactor accident. 

Diagram of components of a reactor fuel assembly
MODELING POTENTIAL REACTOR ACCIDENT CONSEQUENCES | 25

Cesium and Iodine—These two radionuclide groups affect offsite consequence 

analysis because they are released as part of an accident, and the human body can get 

significant radiation doses from them. 

Other radionuclides—MELCOR and MACCS2 also consider other 

radiological inventory in the analysis, and consequence results in NUREG-1935 include 

health effects from the radionuclides released in the accident.

WHAT INFORMATION IS INCLUDED IN MELCOR MODELING?

How physical and chemical processes influence the behavior of radioactive material 

while the core heats up.

How the accident’s extremely high temperatures influence particles’ behavior at the 

molecular level and their physical states (e.g., turning them into gas or small particles 

that can settle or move through the air).

How the radioactive material moves within the containment and reactor coolant system 

(before exiting containment).

How engineered safety systems (such as water sprays and air fan coolers) impact the 

behavior of radioactive material to prevent their release.

If and at what rate the accident releases radioactive material into the environment.

HOW ARE RADIOACTIVE MATERIALS MODELED TO ESCAPE 

FROM CONTAINMENT?

The following sections describe the timing of radioactive material movement while 

onsite and its release to the environment. Figure 4.1 shows how much of the reactor 

core’s available radioactive iodine (I-131) and cesium (Cs-137) is released to the 

environment during the first 48 hours of the accident. 

Peach Bottom Scenarios (Unmitigated Cases)

Long-Term Station Blackout— 20 hours after the scenario begins, molten core 

material penetrates the bottom head of the reactor vessel, pours onto the containment 

floor, spreads across the floor, and contacts the steel containment shell, melting a hole 

through it. 

Short-Term Station Blackout— About 8 hours after the scenario begins, molten core 

material penetrates the bottom head of the reactor vessel, pours onto the containment 

floor, spreads across the floor, and contacts the steel containment shell, melting a hole 

through it.

For the two analyzed Peach Bottom station blackout events, while the core is in 

the reactor vessel, radioactive material moves from the core into the bottom of the 

suppression pool as relief valves send steam into the suppression pool. Some material 

deposits on reactor vessel and pipe surfaces on its way to the suppression pool; the 

rest is retained in the suppression pool as the steam is condensed in the pool.

Surry Scenarios (Unmitigated Cases)

Long-Term Station Blackout— About 45 hours after the scenario begins, the 

pressure in the containment building exceeds the building’s limits, tearing the 

containment liner and cracking the reinforced concrete.
26 | U.S. Nuclear Regulatory Commission

Short-Term Station Blackout— About 25 hours after the scenario begins, the 

pressure in the containment building exceeds the building’s limits, tearing the 

containment liner and cracking the reinforced concrete.

For the two analyzed Surry station blackout events, while the fuel is overheating, 

radioactive material enters the containment building through ruptured reactor coolant 

system piping. Some material deposits on the inside surfaces of the reactor coolant 

system as it moves to the containment building. The remaining contained material 

deposits in the containment building.

Figure 4.1 Percentages of Iodine and Cesium Released to the Environment 

During the First 48 Hours of the Accident for SOARCA Unmitigated 

Scenarios, 1982 Siting Study (SST1), and Historical Accidents.

This figure compares how much iodine-131 and cesium-137 that are normally in the reactor core gets released in 

each accident scenario. The SOARCA unmitigated releases are much smaller than estimated in the earlier 1982 Siting 

Study Siting Source Term 1 (SST1) case. Also note that these releases can begin as early as 3.5 hours (for Surry 

STSBO with TISGTR) to as late as 45 hours (for Surry LTSBO), and some of these releases develop over a period of 

time. For comparison, releases from the Chernobyl and Three Mile Island accidents are included. 

* Chernobyl release data is estimated at 20-40 percent for cesium-137 and 50-60 percent for iodine-131. Three Mile Island released 

an extremely small quantity of iodine-131 (~ 15 curies) and zero cesium-137.

100%

90%

80%

70%

60%

50%

40%

30%

20%

10%

Cesium

Iodine

Peach Bottom

STSBO 

(Unmitigated) 

Chernobyl* Three 

Mile 

Island*

Peach Bottom 

LTSBO 

(Unmitigated)

Surry

LTSBO 

(Unmitigated)

Surry

STSBO 

(Unmitigated)

Surry TISGTR 

(Unmitigated)

Surry

ISLOCA 

(Unmitigated)

1982 Siting 

Study (SST1)
MODELING POTENTIAL REACTOR ACCIDENT CONSEQUENCES | 27

Interfacing Systems Loss-of-Coolant Accident— The scenario begins with the 

hypothesized random failure of 2 valves in series, rupturing a pipe outside of the 

containment building. This provides a path from the reactor core to the environment 

which bypasses containment. About 13 hours after the scenario begins, the accident 

progresses to the point where the fuel overheats and gaseous radioactive particles are 

released through this path. When the overheating fuel is in the reactor vessel, some 

of the radioactive material moves from the fuel through the ruptured pipe and into the 

safeguards building. Most of this radioactive material deposits on reactor vessel and 

Table 4.1 Timing and Quantity of Radioactive Material Released for SOARCA 

Mitigated and Unmitigated Scenarios

From the initiating event, about 

how long until radioactive material 

is released to the environment?

About how much of the available 

radioactive material (Iodine-131 

and Cesium-137) is released 

during the first 48 hours?

Mitigated Case Unmitigateda Mitigated Case Unmitigateda

Peach Bottom Long- Term Station Blackout no release 20 hours no release Iodine: 2%

Cesium: <1%

Peach Bottom Short- Term Station 

Blackout

no release 8 hours no release Iodine: 12%

Cesium: 2%

Surry Long-Term Station Blackout no release 45 hours no release Iodine: <1%

Cesium: <1%

Surry Short-Term Station Blackout no release 

modeled in 

MACCS2

25 hours no release 

within 48 

hours*

Iodine: 1%

Cesium: <1%

Surry Thermally Induced Steam Generator 

Tube Rupture

3.5 hours 3.5 hours Iodine: <1%

Cesium: <1%

Iodine: 1%

Cesium: <1%

Surry Interfacing Systems Loss–ofCoolant Accidentno release 13 hours no release Iodine: 16%

Cesium: 2%

* For the mitigated Surry STSBO, the reactor vessel would fail; however, the containment would not fail until about 66 hours after the blackout. 

A review of available resources and emergency plans shows that adequate mitigation measures could be brought onsite within 24 hours and 

connected and functioning within another 24 hours. Therefore, 66 hours would allow time for mitigation through measures transported from 

offsite, and this mitigation would avert containment failure such that radioactive material would not be released to the environment.

As supported by the SOARCA analyses, it is shown that the accidents evaluated could be mitigated through the actions of the onsite and offsite 

response agencies. The evaluation of the mitigation of source term and truncation of the accident at 48 hours further expands upon the response 

resources through identification of corporate, local, State, and Federal offsite resources. The responsibilities and resources of each of these 

organizations are described in onsite and offsite emergency response plans. These response organizations would mobilize upon request and as 

needed to respond to a severe nuclear power plant accident. These resources are in addition to the mitigative actions by the licensee though the 

use of safety and security enhancements, including SAMGs and 10 CFR 50.54(hh) mitigation measures. At Fukushima Dai-ichi, power supplies 

arrived onsite in less than 12 hours (INPO 11-005).
28 | U.S. Nuclear Regulatory Commission

pipe surfaces and safeguards building (next to containment) filters, with a fraction of it 

entering the environment.

Short-Term Station Blackout with Thermally Induced Steam Generator Tube 

Rupture— About 3.5 hours after the scenario begins, high-pressure, hightemperature gas circulating through the reactor coolant system ruptures a steam 

generator tube, a steam generator safety relief valve is opened, allowing gaseous 

radioactive particles to flow out of the broken tube bypassing the containment 

building. This rupture creates about a 1-inch diameter hole. Minutes later, a reactor 

coolant system pipe also ruptures—creating about a 2-foot diameter hole. In the period 

of time between the two ruptures, much of the radioactive material deposits in the 

failed steam generator, and this settling helps prevent much of it from flowing out into 

the environment. After the pipe rupture, the radioactive material primarily flows into and 

deposits in the containment.
Modeling Hypothetical Accidents at Nuclear Power Plants | 29

This chapter explains emergency planning and 

how emergency response was modeled.

MODELING POTENTIAL REACTOR ACCIDENT CONSEQUENCES | 29
30 | U.S. Nuclear Regulatory Commission

Table 5.1 Evacuation Groups

Schools School populations within 10 miles of the site 

General Public  People within 10 miles of the site who evacuate in response to the evacuation order

Special 

Facilities 

Special-needs population, including residents of hospitals, nursing homes, assisted living 

communities, and prisons within 10 miles of the site 

Nonevacuating 

Public

A portion of the public within 10 miles of the site who refuse to evacuate (assumed to be 0.5 

percent of the population)

Shadow Shadow evacuation occurs when members of the public evacuate from areas that are not under 

official evacuation orders, typically beginning when a large- scale evacuation is ordered

Tail The last 10 percent of the public to evacuate from the 10-mile EPZ

For scenarios leading to core damage and subsequent release of radioactive materials 

to the environment, the local public may be evacuated and/or sheltered. SOARCA 

models tracked the dispersion of radioactive material and analyzed the effect of carrying 

out emergency response for these scenarios. This chapter provides more information 

about how the SOARCA project modeled emergency plans during a severe accident. 

In all scenarios except one, the releases’ delayed timing (even without mitigative 

actions) allowed time to evacuate the local population. In that one scenario, evacuation 

began at the time of radiation release. However, the calculated individual long-term 

cancer fatality risk within 10 miles of the plant is similar to the other scenarios analyzed 

because of the lower probability of core damage due to this scenario. 

WHAT IS EMERGENCY PLANNING?

NRC requires nuclear power plants to have onsite and offsite emergency plans as a 

defense-in-depth measure. NRC evaluates the plants’ emergency planning to ensure 

they can execute their plans and coordinate State and Federal responses. Emergency 

plans focus on protecting public health and safety with the following objectives:

Onsite Objective—Stop the accident. NRC requires the utilities to have onsite 

response that includes technical, maintenance, and management staff that can respond 

within an hour of the accident’s start. Each year, the licensees train and drill this 

capability, and NRC inspects it.

Offsite Objective—Protect the local population through implementation of protective 

actions that include evacuating and sheltering. NRC requires utilities to have offsite 

response support from local and State agencies. The Federal Emergency Management 

Agency inspects this capability every 2 years. Emergency planning zones (EPZs) help 

define where detailed protective strategies would be used during an emergency. 

Every plant must have NRC-approved emergency action levels that dictate declaring an 

emergency well before a severe accident could cause a core melt or radiation release. 

This timing is designed to ensure that emergency plans are implemented before the 

plant is in a serious state and that members of the public are well on their way to 

evacuation before any release begins.
MODELING POTENTIAL REACTOR ACCIDENT CONSEQUENCES | 31

What Are Emergency Planning Zones (EPZs)?

Two EPZs around each nuclear power plant help define what protective 

action strategies will be used during an emergency. Predetermined 

protective action plans are in place for the EPZs to avoid or reduce dose 

from potential exposure of radioactive materials. Utilities base the size 

and shape of their EPZs on site-specific conditions, unique geographical 

features of the area, and demographic information. The detailed 

planning for the EPZs enables emergency responders to extend actions 

beyond the EPZ if conditions warrant. 

Plume Exposure Pathway EPZ—The plume exposure pathway EPZ 

has a radius of about 10 miles from the reactor site. The actions for this 

EPZ can include sheltering, evacuating, and taking potassium iodide 

pills to protect people who inhale or ingest airborne radioactive iodine. 

Ingestion Exposure Pathway EPZ—The ingestion exposure 

pathway EPZ has a radius of about 50 miles from the reactor site. 

The actions for this EPZ can include a ban of contaminated food and 

water to protect people from radioactive material in the food chain. 

Ingestion of contaminated food and water is not treated in the SOARCA 

analyses because 

adequate 

supplies of food 

and water are 

available in the 

United States 

and can be 

distributed to 

areas affected 

by a reactor 

accident.

What Are NRC Regulations?

Emergency Plans

The planning standards of 10 CFR 50.47, “Emergency Plans,” require 

nuclear plant licensees to develop comprehensive emergency response 

plans that include the support of State and local response organizations. 

Licensees must establish procedures to immediately notify offsite 

authorities of an emergency and establish warning systems to provide 

early notification and clear instruction to the public. Licensees 

must demonstrate to NRC that protective measures can and will be 

implemented in the event of a radiological emergency. For details, see 

http://www.nrc.gov/reading-rm/doc-collections/cfr/part050/part050-

0047.html.

18

19

20

21

18 9 10

11

12

13

14

23

18

24

16

17

18

6

4

3

2

1

8

7

WHAT INFORMATION IS 

INCLUDED IN EMERGENCY 

PLAN MODELING?

The SOARCA team modeled the specific 

emergency plans for each site using 

detailed information that included the 

following:

 Population based on data from the 2000 

U.S. Census and projected to 20052. 

 Evacuation time estimates from 

emergency plans.

 Plans to relocate populations from 

contaminated areas.

Using each site’s emergency plan 

information, the SOARCA team organized 

the population into several groups and 

modeled each group’s evacuation timing 

along with the timing of the accident. Table 

5.1 provides a description of some of the 

groups. Other population groups modeled 

include people who leave on their own 

initiative prior to the evacuation order as 

well as people who do not evacuate. 

WHAT DOES MODELING 

DEMONSTRATE ABOUT 

EMERGENCY PLANNING?

The MACCS2 computer code calculates 

the radiation dose to the public based on 

evacuating, sheltering, and returning to the 

area after the event. Figure 5.1 illustrates 

the modeled timing of the unmitigated 

Peach Bottom LTSBO scenario and the 

timing of emergency response. Because 

the analyzed accident scenarios take 

several hours to start releasing radioactive 

material to the environment, this provides 

time for the population to evacuate 

before potential radiation exposure. The 

analysis considered seismic impacts on 

emergency response (e.g., loss of bridges, 

traffic signals, and delayed notification). 

2 2010 U.S. Census data was not used because 

most calculations were already completed by the 

time it was released. Changes in population over the 

last decade are not expected to have a significant 

impact on any of the reported individual cancer 

fatality risks.
32 | U.S. Nuclear Regulatory Commission

NRC Staff during an emergency preparedness drill

However, the MACCS2 modeling showed that seismic 

impacts for these two sites did not impact risk calculations 

because seismic impacts only affect the immediate phase 

of the accident when people are sheltering or evacuating. 

SOARCA’s risk calculations are dominated instead by longterm exposure of the population after they return home 

when told it is safe to do so. 

Figure 5.1 shows that groups are sheltered and evacuated 

before radioactive release begins. The timeline notes key 

accident progression and emergency response events. 

In each analyzed scenario, the plants follow their stated 

emergency response plans and promptly notify offsite 

authorities who activate their emergency notification 

systems (sirens) and direct the public to evacuate. 

Station Blackout 

(0:00)

Release of radioactive 

materials to the environment 

starts (assuming accident 

proceeds unmitigated) 

(20:00)

Plant declares 

site area 

emergency (0:15)

Sirens sounded for 

general emergency (1:30)

Plant declares general 

emergency (0:45)

Sirens sounded for site 

area emergency (1:00)

Evacuation complete for EPZ 

population (including schools, 

general public, special facilities, and 

tail and excluding those who choose 

not to evacuate) (7:00)

0 2 4 6 8 10 12 14 16 18 20

Figure 5.1 Evacuation Timing for Peach Bottom Unmitigated Long-Term 

Station Blackout (hours).
Modeling Hypothetical Accidents at Nuclear Power Plants | 33

This chapter describes the models to calculate 

health consequences for SOARCA scenarios that 

release radioactive materials to the environment.

MODELING POTENTIAL REACTOR ACCIDENT CONSEQUENCES | 33
34 | U.S. Nuclear Regulatory Commission

The team modeled the unmitigated scenarios’ calculated 

releases and subsequent health consequences. Even 

in the unmitigated scenarios, modeling indicated that 

essentially no one would die from acute radiation 

exposure (due to the length of time for the accident 

to progress and the relatively small releases) and that 

there would be a very small possibility of long-term 

cancer fatalities. This chapter provides an explanation and 

background information about how SOARCA modeled the 

health consequences.

HOW ARE HEALTH CONSEQUENCES 

REPORTED IN SOARCA?

Exposure to radiation can have a variety of different 

health effects depending on the specific type and 

intensity of exposure. In addition, radiation affects 

different people in different ways. Large, high-intensity 

exposures can cause acute health effects that range from 

nausea and skin reddening to death. In addition to acute 

health effects, radiation exposures are related to the 

occurrence of cancer later in life. The two types of health 

consequences reported in SOARCA are early fatalities 

from very large and intense exposures and fatalities that 

result from radiation-induced cancers. 

Early Fatality Risk—Individual deaths that occur shortly 

(usually within a few weeks or months) after exposure to 

large doses of radiation. The report provides this number 

as the average individual risk of an early fatality. For 

scenarios analyzed, the early fatality risk is essentially 

zero. 

Long-Term Cancer Fatality Risk—Cancer fatalities 

that occur years after exposure to radiation. This number 

represents the average individual risk of dying from cancer due to radiation exposure 

following the specific hypothesized severe accident scenario. For the scenarios 

analyzed, long-term cancer fatality risk is very small.

How Is Radiation Measured?

Units that measure how much radioactive material decays 

over a period of time:

•	Curie (Ci)

•	 Becquerel (Bq): 1 Bq = 2.7 x 10-11 Ci

Units that measure the effects of ionizing radiation on 

humans:

•	rem 

•	 Sievert (Sv): 1 Sv = 100 rem

More information about radiation and its health effects is 

at http://www.nrc.gov/about-nrc/radiation/rad-healtheffects.html. 

A Geiger counter is a tool that measures radiation in the 

environment. 
MODELING POTENTIAL REACTOR ACCIDENT CONSEQUENCES | 35

HOW ARE LONG-TERM CANCER 

FATALITY RISKS MODELED?

Modeling long-term cancer fatality risk is 

controversial because medical researchers 

disagree on the evidence that describes the 

adverse effects of low radiation doses. The 

SOARCA project used two long-term cancer 

fatality risk models to provide additional 

information on the effects of different modeling 

approaches on the potential range of health 

consequences:

Linear-no-threshold dose response model—

This model is based on the conclusion that 

any amount of radiation dose (no matter how 

small) can incrementally increase cancer risk. It 

is a basic assumption used in many regulatory 

limits, including NRC’s regulations and past 

assessments.*

Truncation dose response model—To provide 

additional information on the potential range 

of health consequences, the SOARCA project 

calculated long-term cancer fatality risk assuming 

the linear-no-threshold model and a range of 

truncation or cutoff doses below which the 

cancer risk is not quantified. When comparing 

offsite consequence results for the linear-no-threshold model and truncation model, these 

truncation values make the already small long-term cancer fatality risk values even smaller 

(by orders of magnitude in some cases).

SOARCA uses two dose truncation values:

620 mrem per year—This represents the U.S. average individual background dose 

including medical exposures. 

5 rem (5,000 mrem) per year with a 10 rem lifetime cap—This value was chosen based 

on the Health Physics Society position statement in “Radiation Risk in Perspective” (July 

2010).

* Use of the linear no-threshold model for low radiation exposures (below 0.1 sievert or 10 rem) to project 

future long-term cancer fatality risk to individuals receiving such exposures is currently being debated 

within the scientific community. Many radiation protection organizations, such as the U.S. National Council 

on Radiation Protection and Measurements, the International Commission on Radiological Protection, 

the United Nations Scientific Committee on Exposure to Atomic Radiation, and the U.S. Health Physics 

Society, caution that there is considerable uncertainty when computing cancer deaths resulting from 

small additional exposures to large populations over many years and should only be done under explicit 

conditions such as in the SOARCA project or not at all.

As a resident of the United States, how am I 

exposed to radiation?

SOARCA studies health effects in situations where a severe accident 

releases radiation to the public. To provide some perspective, people 

generally receive an average total dose of ionizing radiation of about 

620 millirem per year. Of this total, the chart shows that natural 

sources of radiation account for about 50 percent and manmade 

sources account for the other 50 percent.

Sources of Radiation Exposure in the United States

Natural Sources - 50%

~310 millirem (0.31 rem)

Source: NCRP Report No.160(2009) 

Full report is available on the NCRP Web site at www.NCRPpublications.org.

Industrial and

Occupational - .1%

Consumer Products - 2%

Nuclear Medicine - 12%

Internal - 5%

Terrestrial (Soil) - 3%

Cosmic (Space) - 5%

Manmade Sources - 50%

~310 millirem (0.31 rem)

Radon and

Thoron - 37%

Medical

Procedures -

36%
36 | U.S. Nuclear Regulatory Commission

The MACCS2 code looks at atmospheric transport of radioactive material using a cloud, or 

plume, that travels in a straight line following the wind direction. This model of short-term 

and long-term dose accumulation includes several pathways: radiation from the plume 

(cloudshine), radiation from material that reaches the ground (groundshine); inhalation, 

deposition onto the skin, and food and water ingestion. The ingestion pathway was not 

used in the analyses reported here because uncontaminated food and water supplies are 

abundant within the United States, and it is unlikely that the public would eat radioactively 

contaminated food. The following dose pathways are included in the reported risk metrics:

 Cloudshine during plume passage

 Groundshine during the emergency and long-term phases from deposited aerosols.

 Inhalation during plume passage and following plume passage from radioactive dust 

kicked up by weather or human and vehicle traffic. This dust factor covers both the 

emergency and long-term phases.

Figure 6.1 Transport Pathways of Radioactive Materials.
MODELING POTENTIAL REACTOR ACCIDENT CONSEQUENCES | 37

WHAT INFORMATION IS INCLUDED IN THE MACCS2 MODELING?

 When and at what rate the accident releases radioactive material into the 

environment (from MELCOR analysis described in Chapter 4).

 Protective measures (such as evacuation) taken by the offsite population (from the 

modeling of emergency plans described in Chapter 5).

 Site-specific weather data.

 Downwind transport of the radioactive material released into the environment.

 How each type of radionuclide will impact the body.

 Radiation exposure of the offsite population and the health effects caused by this exposure.

Figure 6.2 Information Used to Model Health Effects 

in SOARCA.

Model Airborne and Deposited Concentrations

Report Health Effects

• Long-term cancer fatalities

Health Effects From 

Chronic Exposure

Model Habitability Criteria

Information that includes:

• Cost-based decision to 

decontaminate, interdict, condemn

Model Late Doses

Information that includes:

• Radionuclide type

• Exposure pathway

• Target organs

Report Health Effects

• Early fatalities

• Long-term cancer fatalities

Health Effects From 

Early Exposure

Model People Behavior

Information that includes:

• Population

• Emergency Response

Model Early Doses

Information that includes:

• Type of radionuclide

• Exposure pathway

• Target organs

Release of Radioactive 

Material into Environment

Atmospheric Transport 

and Dispersion

Information that includes:

• Type of radioactive material released

• Amount of radioactive material released

• Timing of the release

Information that includes:

• Site-specific weather data
38 | U.S. Nuclear Regulatory Commission

HOW ARE RADIOACTIVE MATERIALS MODELED TO MOVE 

DOWNWIND AND AFFECT THE POPULATION?

Radioactive materials are released from plant buildings as aerosol particles in a plume of 

steam and other gases. MACCS2 uses site-specific weather data to calculate the downwind 

concentration of radioactive material in the plume and the resulting population exposures 

and health effects. MACCS2 then applies a statistical model to calculate the average 

individual fatality risk as a result of the variability in the weather. 

SOARCA modeled individual radiation exposure from inhaling the aerosol particles and by 

direct radiation from aerosol particles in the air and on the ground. A small portion of this 

exposure occurs during the early phase of the accident when the aerosol particles are being 

released from the plant buildings and while people are evacuating. Most of this is long-term 

exposure after land is decontaminated and people are allowed to return home. SOARCA 

modeled evacuees returning home based on guidance that outlines when it would be safe 

to do so. For the Surry model, SOARCA uses the U.S. Environmental Protection Agency’s 

“Manual of Protective Action Guides for Nuclear Incidents” to determine when the 

population can return to an area. For the Peach Bottom model, SOARCA uses Pennsylvaniaspecific criteria. This calculation also includes doses to the population in lightly contaminated 

areas where they were neither evacuated nor relocated. SOARCA did not model people 

who were exposed by eating food contaminated by aerosol particles because emergency 

plans will prevent distribution of contaminated food and because of the expected availability 

of uncontaminated food from other areas.
Modeling Hypothetical Accidents at Nuclear Power Plants | 39

This chapter summarizes the results and 

conclusions from the SOARCA research project.

MODELING POTENTIAL REACTOR ACCIDENT CONSEQUENCES | 39
40 | U.S. Nuclear Regulatory Commission

The SOARCA results demonstrate the potential benefits of the mitigation measures 

analyzed in this project. SOARCA shows that successful mitigation either prevents core 

damage or prevents, delays, or reduces offsite health consequences. In addition, the 

SOARCA team ran scenarios that demonstrate the consequences if certain mitigation 

measures are not successful. The unmitigated scenario results presented in this chapter 

demonstrate that, even in these cases, the public health consequences are very low.

WHAT ARE THE RESULTS OF THE MITIGATED SCENARIOS?

All mitigated cases of SOARCA scenarios, except for one, result in prevention of core damage 

and/or no offsite release of radioactive material. The only mitigated case still leading to an 

offsite release was the Surry thermally induced steam generator tube rupture. In this case, 

mitigation is still beneficial in that it keeps most radioactive material inside containment and 

delays the onset of containment failure by about 2 days.

Early Fatality Risk—As a result, the mitigated scenarios show zero risk of early fatalities 

from radiation exposure.

Long-Term Cancer Fatality Risk—As a result, the mitigated scenarios result in either zero 

risk or very small risk of long-term cancer fatalities, depending on the specific scenario. 

HOW WOULD OPERATOR ACTIONS MITIGATE ACCIDENTS?

The operators mitigate the station blackout and thermally induced steam generator tube 

rupture (TISGTR) accidents by manually controlling pumps to inject water to keep the 

reactor core covered and cooled. The reactor’s steam-turbine-driven safety injection pumps 

would be used in conjunction with portable diesel generators (stored onsite) to provide 

instrument readings. The operators mitigate the interfacing systems loss-of-coolant accident 

(ISLOCA) scenario by using normal plant equipment and procedures to ensure sufficient 

flow of water through the reactor coolant system to keep the core cooled.

WHAT ARE THE RESULTS OF UNMITIGATED SCENARIOS?

Early Fatality Risk—The unmitigated scenarios result in essentially zero risk of early 

fatalities. Although these unmitigated scenarios result in core damage and release of 

radioactive material to the environment, the release is delayed, which allows the population 

to take protective actions (including evacuation and sheltering). Therefore, the public would 

not be exposed to dangerous amounts of radioactive material.

Long-Term Cancer Fatality Risk—For the unmitigated scenarios, the individual risk of 

a long-term cancer fatality is calculated to be very small—regardless of which distance 

interval (e.g., 0-10 miles, 0-20 miles, ... 0-50 miles) or low-dose calculation model is used. 

Table 7.1 summarizes the results based on the linear-no-threshold dose response model for 

estimating the risk for individuals located within 10 miles of each plant.
MODELING POTENTIAL REACTOR ACCIDENT CONSEQUENCES | 41

WHAT DO SOARCA RESULTS INDICATE ABOUT 

CONSEQUENCES OF SEVERE ACCIDENTS?

The SOARCA results for the two plants analyzed are as follows. These results, while 

specific to Peach Bottom and Surry, may be generally applicable to plants with similar 

designs. Additional work would be needed to confirm this, however, since differences 

exist in plant-specific designs, procedures, and emergency response characteristics.

 When operators are successful in using available on-site equipment during the 

accidents analyzed in SOARCA, they can prevent reactor fuel from melting, or delay 

or reduce releases of radioactive material to the environment.

 SOARCA analyses indicate that all modeled accident scenarios, even if operators 

are unsuccessful in stopping the accident, progress more slowly and release much 

smaller amounts of radioactive material than calculated in earlier studies.

 As a result, public health consequences from severe nuclear power plant accidents 

modeled in SOARCA are smaller than previously calculated.

 The delayed releases calculated provide more time for emergency response actions 

such as evacuating and sheltering for affected populations. For the scenarios 

analyzed, SOARCA shows that emergency response programs, if implemented as 

planned and practiced, reduce the risk of public health consequences.

Table 7.1 SOARCA Results: Mitigated and Unmitigated Cases

About how likely is the 

accident to occur?

About what is the annual average risk* of a 

long-term cancer fatality for this scenario for an 

individual located within 10 miles of the plant?

Mitigated Case Unmitigated Case

Peach Bottom LTSBO 1 event in 300,000 reactor 

years

zero 1 in 3 billion

Peach Bottom STSBO 1 event in 3 million reactor 

years

zero 1 in 20 billion

Surry LTSBO 1 event in 50,000 reactor 

years

zero 1 in 1 billion

Surry STSBO 1 event in 500,000 reactor 

years

zero** 1 in 10 billion

Surry TISGTR 1 event in 3 million reactor 

years

1 in 10 billion 1 in 10 billion

Surry ISLOCA 1 event in 30 million 

reactor years

zero 1 in 100 billion

* Estimated risks below 1 in 10 million reactor years should be viewed with caution because of the potential impact of events not studied in 

the analyses and the inherent uncertainty in very small calculated numbers.

** For the mitigated Surry STSBO, the reactor vessel would fail; however, the containment would not fail until about 66 hours after the 

blackout. A review of available resources and emergency plans shows that adequate mitigation measures could be brought onsite within 24 hours 

and connected and functioning within 48 hours. Therefore, 66 hours would allow time for mitigation via equipment brought to the site from offsite, 

and this mitigation would avert containment failure such that radioactive material would not be released to the environment.
42 | U.S. Nuclear Regulatory Commission

Figure 7.1 Scenario-specific Risk of Dying From Long-Term Cancer for an 

Individual within 10 Miles of the Plant (per reactor-year).

 Both mitigated (operator actions are successful) and unmitigated (operator actions 

are unsuccessful) cases of all modeled severe accident scenarios in SOARCA cause 

essentially no risk of death during or shortly after the accident.

 SOARCA’s calculated longer term cancer fatality risks for the accident scenarios 

analyzed are millions of times lower than the general U.S. cancer fatality risk.

Figure 7.1 compares SOARCA’s scenario-specific latent cancer fatality risks for an 

individual within 10 miles of the plant to the NRC Safety Goal and to an extrapolation of 

the 1982 Siting Study SST1 results.

* The 1982 Siting Study did not calculate risk of long-term cancer deaths. Therefore, to compare the 1982 Siting Study 

(SST1) results to SOARCA’s results for risk of long-term cancer death, the SST1 release was put into the MACCS2 code 

files for Peach Bottom and Surry unmitigated STSBO calculations.

Figure 7.1 Scenario-specific Risk of Dying From Long-Term Cancer for an

Individual within 10 Miles of the Plant (Per Reactor-Year).

PEACH BOTTOM SURRY

Peach Bottom

STSBO 

Peach Bottom 

LTSBO

Surry

LTSBO

Surry

STSBO

Surry 

TISGTR

Surry

ISLOCA 

1982 Siting 

Study (SST1)

1982 Siting 

Study (SST1)

1 in 

100 thousand

1 in 

1 million

1 in 

10 million

1 in 

100 million

1 in 

1 billion

1 in 

10 billion

1 in 

100 billion

1 in 

1 trillion

NRC SAFETY GOAL FOR LONG-TERM CANCER FATALITIES = 2 in 1 million

SOARCA unmitigated scenario

SOARCA mitigated scenario

1982 Siting Study (SST1)*
MODELING POTENTIAL REACTOR ACCIDENT CONSEQUENCES | 43

HOW DO SOARCA RESULTS COMPARE TO PAST STUDIES?

The SOARCA offsite consequence calculations are generally smaller than reported 

in earlier studies. To provide perspective between SOARCA results and the more 

conservative estimates of severe reactor accident outcomes found in earlier NRC 

publications, SOARCA results are compared to the results of one of these previous 

publications: NUREG/CR-2239, “Technical Guidance for Siting Criteria Development,” 

commonly referred to as the 1982 Siting Study. For example, the 1982 Siting Study 

calculated 92 early fatalities for Peach Bottom and 45 early fatalities for Surry for 

the siting source term 1(SST1) release of radioactive material. In contrast, SOARCA 

calculated essentially zero early fatalities for both sites. The exact basis for long-term 

cancer fatality results in the 1982 Siting Study could not be recovered. The 1982 

Siting Study’s computer code (CRAC2) is no longer available and some of the models 

and modeling choices used in that study could not be reconstructed. Therefore, the 

SOARCA team compared SOARCA results with the 1982 Siting Study results by 

replacing the SOARCA source term with the larger source term (SST1) assumed in 

the 1982 Siting Study. Figure 7.1 shows this comparison for individuals within 10 miles 

of the plant. The long-term cancer fatality calculations based on the 1982 Siting Study 

SST1 source term are higher than the long-term cancer fatality calculations for SOARCA 

scenarios, however the difference diminishes when considering larger areas out to 

a distance of 50 miles from the plant because in both studies, large populations are 

assumed to be exposed to small annual doses from returning home after the accident.

HOW DO SOARCA RESULTS COMPARE TO THE NRC SAFETY 

GOAL AND OVERALL U.S. CANCER RISKS?

To gain perspective, it can be helpful to compare the SOARCA results with the NRC 

Safety Goal and the average annual risk of a cancer fatality in the United States. The 

NRC’s Safety Goal states that cancer fatality risk to the population near an operating 

nuclear power plant should not exceed one-tenth of one percent (0.1 percent) of the 

sum of cancer fatality risks resulting from all causes. According to the American Cancer 

Society, 1 in 553 people died of cancer in the United States in 2006. One-tenth of one 

percent of this number equals about 1 death per 553,000 people. The NRC Safety Goal 

for long-term cancer fatality risk from nuclear power plant operation (i.e., 2 in 1 million) 

is set at 1,000 times lower than the sum of cancer fatality risks resulting from all other 

causes (i.e., ~ 2 in 1,000) per year. As shown in Table 7.1, SOARCA’s calculated cancer 

fatality risks for the postulated events in this study range from about 1 in 1 billion per 

year to 1 in 100 billion per year.

The calculated cancer fatality risks from the scenarios analyzed in SOARCA are 

thousands of times lower than the NRC Safety Goal and millions of times lower than 

the general U.S. cancer fatality risk.

WHAT ARE THE LIMITATIONS OF COMPARING SOARCA RESULTS 

TO THE NRC SAFETY GOAL AND OVERALL U.S. CANCER RISKS?

Comparisons of SOARCA’s calculated long-term cancer fatality risks to the NRC Safety 

Goal and the average annual U.S. cancer fatality risk from all causes are provided to 

give context that may help the reader to understand the contribution to cancer risks 

from these nuclear power plant accident scenarios. However, such comparisons 

have limitations for which the readers should be aware. Relative to the safety goal 
44 | U.S. Nuclear Regulatory Commission

comparison, the safety goal is intended to encompass all accident scenarios. SOARCA 

does not examine all scenarios typically considered in PRA, even though it includes the 

important scenarios. In fact, any analytical technique, including PRA, will have inherent 

limitations of scope and method. As a result, comparison of SOARCA’s scenariospecific calculated long-term cancer fatality risks to the NRC Safety Goal is necessarily 

incomplete. However, it is intended to show that adding multiple scenarios’ low risk 

results in the 1 in 1 billion to 1 in 100 billion range to approximate a summary risk from 

all scenarios, would yield a summary result that is also below the NRC Safety Goal of 2 

in 1 million.

Relative to the U.S. average annual individual risk of a cancer fatality comparison, the 

sources of an individual’s cancer risk include a complex combination of age, genetics, 

lifestyle choices, and other environmental factors whereas the consequences from 

a severe accident at a nuclear plant are involuntary and unlikely to be experienced by 

most individuals.
Modeling Hypothetical Accidents at Nuclear Power Plants | 45

MODELING POTENTIAL REACTOR ACCIDENT CONSEQUENCES | 45
46 | U.S. Nuclear Regulatory Commission

GLOSSARY

Acute health effects—Health effects which occur within two months of exposure

Advisory Committee on Reactor Safeguards (ACRS)—The ACRS is an independent 

review committee that advises the Commission, independent of the NRC technical 

staff, regarding the licensing and operation of reactor facilities and related safety 

issues, the adequacy of proposed reactor safety standards, technical and policy issues 

related to the licensing of new reactor designs, and other matters referred to it by the 

Commission.

Boiling-Water Reactor—In a commercial boiling-water reactor, the reactor core creates 

heat and a single loop both delivers steam to the turbine generator and returns water to 

the reactor core to cool it. The cooling water is force-circulated by electrically powered 

pumps. Emergency cooling water is supplied by other pumps that can be powered by 

onsite diesel generators. Other safety systems, such as the containment building air 

coolers, also need electric power.

Containment Structure—An enclosure around a nuclear reactor to confine radioactive 

material that otherwise might be released to the atmosphere in the event of an 

accident. Pressurized-water reactor containments are usually cylindrical with a domeshaped top and made of steel-reinforced concrete and a steel liner.

Coolant—A substance circulated through a nuclear reactor to remove or transfer heat. 

All commercial nuclear reactors in the United States use water. 

Core Damage—Events that heat up the reactor core to the point at which fuel damage 

is anticipated or the drying out and heatup of the reactor core to the point at which 

prolonged oxidation and severe fuel damage lead to release of radioactive material from 

the fuel. 

Core Damage Frequency—An expression of the likelihood that, given the way a reactor 

is designed and operated, an accident could cause the fuel in the reactor to heat up to 

the point at which it would be damaged and potentially melt.

Early Fatalities—Human deaths that occur shortly after exposure to radiation, usually 

within a few weeks or months.

Emergency Operating Procedures (EOPs)—Plant-specific procedures containing 

instructions for operating staff to implement preventive measures for managing 

accidents.

Emergency Planning Zones (EPZ)—The EPZs around each nuclear power plant 

help define what protective action strategies will be used during an emergency. 

Predetermined protective action plans are in place for each site and are designed to 

avoid or reduce dose from potential exposure of radioactive materials. Utilities base the 

size and shape of their EPZs on site-specific conditions, unique geographical features 

of the area, and demographic information. The plume exposure EPZ extends about 10 

miles from the plant, and the ingestion EPZ extends about 50 miles from the plant.

Evacuation Time Estimate (ETE)—The estimated time to mobilize and evacuate the 

public from a defined area. The ETE considers residents of the EPZ, transients, people 

visiting but not living within the EPZ, and special facilities including schools.
MODELING POTENTIAL REACTOR ACCIDENT CONSEQUENCES | 47

Feedwater—Water supplied to the reactor pressure vessel (in a boiling-water reactor) or 

the steam generator (in a pressurized-water reactor) that removes heat from the reactor 

fuel rods by boiling and becoming steam. The steam becomes the driving force for the 

plant turbine generator.

Ingestion Pathway—The potential routes for radionuclides from various sources to enter 

water, the food chain, or get into a person’s mouth in day-to-day activities.

Long-Term Cancer Fatalities—Cancer fatalities that occur years after exposure to radiation. 

MACCS2—A general-purpose computer code for estimating offsite impacts following 

release of radioactive material. MACCS2 is applicable to diverse reactor and nonreactor 

situations. It considers atmospheric transport and dispersion under time-variable 

weather conditions, short- and long-term mitigation actions, and exposure pathways to 

determine health effects and economic costs. 

MELCOR—An integrated, engineering-level computer code used to model the 

progression of postulated accidents in light-water reactors as well as nonreactor 

systems (e.g., spent fuel pool and dry cask). MELCOR is a modular code consisting 

of three general types of packages: (1) basic physical phenomena, (2) reactor-specific 

phenomena, and (3) support functions. These packages model the major systems of a 

nuclear power plant and their associated interactions. 

Mitigating Actions—Actions performed by plant operators to prevent core damage and/

or the release of radioactive material.

Pressurized-Water Reactor—In a commercial pressurized light-water reactor, (1) the 

reactor core creates heat, (2) pressurized water in the primary coolant loop carries the 

heat to the steam generator, and (3) the steam generator converts the water into steam 

in a secondary loop to drive the turbine generator to produce electricity.

Probabilistic Risk Assessment—A method to calculate risk by assessing both the 

probability of an event and its consequences. This procedure involves asking a series of 

three questions called the “risk triplet:” (1) What can go wrong? (2) How likely is it? (3) 

What would be the consequences?

Radiation—Energy that travels in the form of waves or high-speed particles. Alpha 

particles, beta particles, gamma rays, x-rays, neutrons, high-speed electrons, high-speed 

protons, and other particles capable of producing ions. Radiation, as used in 10 CFR Part 

20 “Standards for Protection Against Radiation,” does not include nonionizing radiation 

such as radio waves or microwaves, or visible, infrared, or ultraviolet light (see also 10 

CFR 20.1003, “Definitions”).

Reactor Core—The central portion of a nuclear reactor which contains the fuel 

assemblies, moderator, neutron poisons, control rods, and support structures. The 

reactor core is where fission takes place.

Reactor Fuel—Boiling-water reactors and pressurized-water reactors use ceramic pellets 

containing enriched uranium dioxide (UO2). These pellets are stacked and sealed inside 

long, slender, zirconium metal-based alloy (Zircaloy) tubes to form fuel rods. Fuel rods 

are assembled into bundles called fuel assemblies that are loaded into the reactor core. 
48 | U.S. Nuclear Regulatory Commission

Reactor-Year—The operation of one nuclear reactor for 1 year. 

Severe Accident— A severe accident may challenge safety systems beyond a nuclear 

power plant’s design limits, potentially damaging or degrading the reactor core and its 

containment buildings.

Severe Accident Management Guidelines (SAMGs)—Guidelines that plants voluntarily 

put in place in the late 1990s to contain or reduce the impact of accidents that damage 

a reactor core. 

More term definitions are available online at the NRC Glossary at

www.nrc.gov/reading-rm/basic-ref/glossary.html. 

SELECTED REFERENCES

NUREG-1935: State-of-the-Art Reactor Consequence Analyses - Main Report. 

NUREG/CR-7110, Volumes 1 and 2: State-of-the-Art Reactor Consequence Analyses – 

Peach Bottom Integrated Analyses and Surry Integrated Analyses.

NRC REFERENCES:

NRC Fact Sheets and Brochures 

http://www.nrc.gov/reading-rm/doc-collections/fact-sheets/ 

Backgrounder on Chernobyl Nuclear Power Plant Accident 

http://www.nrc.gov/reading-rm/doc-collections/fact-sheets/chernobyl-bg.html

Backgrounder on Three Mile Island Accident 

http://www.nrc.gov/reading-rm/doc-collections/fact-sheets/3mile-isle.html

NUREG/BR-0164, Rev. 5: NRC—Regulator of Nuclear Safety 

http://www.nrc.gov/reading-rm/doc-collections/nuregs/brochures/br0164/r5/

NUREG/BR-0175, Rev. 1: A Short History of Nuclear Regulation 1946-1999 

http://www.nrc.gov/reading-rm/doc-collections/nuregs/brochures/br0175/

NUREG-1350, Vol. 23: Information Digest, 2011–2012. 

http://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr1350/

REFERENCES FOR HISTORICAL ACCIDENTS:

Nuclear Emergency Response Headquarters – Government of Japan, “Report of the 

Japanese Government to the IAEA Ministerial Conference on Nuclear Safety – The 

Accident at TEPCO’s Fukushima Nuclear Power Stations,” IAEA Ministerial Conference 

on Nuclear Safety, Vienna, Austria, June 7, 2011.

Institute of Nuclear Power Operations, “Special Report on the Nuclear Accident at the 

Fukushima Dai-ichi Nuclear Power Station, “ INPO-11-005, Institute of Nuclear Power 

Operations, Atlanta, GA, November 2011.

Report of The President’s Commission on The Accident at Three Mile Island. October, 1979.

Rogovin, M. & Frampton, G. T. NUREG/CR-1250, Vols. III: Three Mile Island: A Report to 

the Commissioners and to the Public. 1980.

Walker, J. S. Three Mile Island: A Nuclear Crisis in Historical Perspective. University of 
MODELING POTENTIAL REACTOR ACCIDENT CONSEQUENCES | 49

California Press: Berkeley, CA. 2004.

The Chernobyl Forum. Chernobyl’s Legacy: Health, Environmental and Socio-Economic 

Impacts and Recommendations to the Governments of Belarus, the Russian Federation 

and Ukraine. September 2005. http://www.iaea.org/http://www.iaea.org/

Organization for Economic Cooperation and Development (OECD) Nuclear Energy 

Agency, “Chernobyl: Assessment of Radiological and Health Impacts, 2002 Update of 

Chernobyl: 10 Years On.” http://www.oecd-nea.org/rp/chernobyl/ 

REFERENCES FOR HISTORICAL RISK AND CONSEQUENCE 

STUDIES:

WASH-1400 (NUREG-75/014): Reactor Safety Study. An Assessment of Accident Risks 

in U.S. Commercial Nuclear Power Plants. 1975.

NUREG/CR-2239, Technical Guidance for Siting Criteria Development. November 1982.

NUREG-1150, Severe Accident Risks: An Assessment for Five U.S. Nuclear Power 

Plants. 1991.

REFERENCES FOR ACCIDENT PROGRESSION:

NUREG/CR-6119, Vol. 1, Rev. 3, MELCOR Computer Code Manuals, Vol. 1: Primer and 

User’s Guide, Version 1.8.6. 2005.

NUREG/CR-6042, Rev. 2: Perspectives on Reactor Safety. 2002. 

NUREG/CR-6613, SAND97-0594: Code Manual for MACCS2 User’s Guide. 1997.

REFERENCES FOR EMERGENCY PREPAREDNESS:

NUREG/CR-6864: Identification and Analysis of Factors Affecting Emergency 

Evacuations. 2005.

NUREG/CR-6953, Vol. I-II: Review of NUREG-0654, Supplement 3, Criteria for 

Protective Action Recommendations for Severe Accidents. 2007. 

NUREG/CR-6863: Development of Evacuation Time Estimate Studies for Nuclear Power 

Plants. 2008. 

NUREG/CR-6981: Assessment of Emergency Response Planning and Implementation 

for Large Scale Evacuations. 2008. 

REFERENCES FOR HEALTH EFFECTS:

Fact Sheet on Biological Effects of Radiation http://www.nrc.gov/reading-rm/doccollections/fact-sheets/bio-effects-radiation.html. 

NRC Radiation Protection Web site http://www.nrc.gov/about-nrc/radiation.html. 

Health Physics Society, “Radiation Risk in Perspective” (July 2010).


U.S. Nuclear Regulatory Commission

NUREG/BR-0359

Revision 1

December 2012