Historical Perspectives and Insights on 

Reactor Consequence Analyses

Prepared By 

Hossein P. Nourbakhsh 

November 2008 

Advisory Committee on Reactor Safeguards 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 
 ii

ABSTRACT 

This paper has been prepared for use by the NRC Advisory Committee on Reactor 

Safeguards (ACRS) in its ongoing review of the State-Of-the-Art Reactor Consequence 

Analyses (SOARCA) Project. Major contributions to consequence assessment have 

been summarized to provide insights and historical perspectives on previous state-ofthe-art analyses of the consequences of severe reactor accidents. The feasibility of 

using a simplified approach for updating results from earlier Level-3 probabilistic risk 

assessments (PRAs) such as the NUREG-1150 Study for comparison with aspects of 

SOARCA results has also been discussed. 

The views expressed in this paper are solely those of the author and do not necessarily 

represent the views of the ACRS. 
 iii

TABLE OF CONTENTS 

ABSTRACT.......................................................................................i 

FIGURES.........................................................................................iv 

TABLES...........................................................................................vi 

ABBREVIATIONS..............................................................................vii 

1. INTRODUCTION............................................................................1 

2. MAJOR NRC-SPONSORED ASSESSMENTS OF REACTOR ACCIDENT 

CONSEQUENCES ..........................................................................4

 2.1 Reactor Safety Study (WASH-1400)........................................4 

 2.2 Post TMI-2 Review of Source Term Technical Bases and 

 Sandia Siting Study............................................................9 

 2.3 NUREG-1150 Study..........................................................14 

 2.4 Reassessment of Selected Factors Affecting Siting of Nuclear 

 Power Plants...................................................................23 

3. RECENT ADVANCES IN UNDERSTANDING OF SEVERE ACCIDENT 

 PHENOMENOLOGY AND CONTAINMENT FAILURE MECHANISMS......27 

 3.1 Direct Containment Heating.................................................27 

 3.2 Steam Explosion..............................................................29 

 3.3 Drywell Melt-through in Mark I Containments.........................29 

 . 

4. A SIMPLIFIED APPROACH TO UPDATE THE RESULTS OF EARLIER 

 LEVEL-3 PRAs FOR COMPARISON WITH SOARCA...........................32 

5. SUMMARY AND CONCLUSIONS.....................................................39 

6. REFERENCES..............................................................................40 
 iv

FIGURES

1. Frequency of Fatalities due to Man-Caused Events (WASH-1400)...............7 

2. Frequency of Fatalities due to Natural Events (WASH-1400)......................8 

3. Comparison of frequency distribution (CCDF) of iodine release from 

 Reactor Safety Study release categories for Surry with that assumed 

 for Siting Source Terms.....................................................................11 

4. Surry frequency of core damage from internal events calculated in 

 WASH-1400, NUREG-1150, the IPEs, and the recent result of the 

 SPAR model for Surry (performed in support of SOARCA).......................15 

5. Conditional Probability of Accident Progression Bins at Surry 

(NUREG-1150, p. 3-12).....................................................................16 

6. Conditional Probability of Accident Progression Bins at 

 Peach Bottom (NUREG-1150, p. 4-13).................................................17 

7. Comparison of frequency distribution (CCDF) of iodine release 

 predicted by the NUREG-1150 Study for Surry with that obtained from 

 the Reactor Safety Study release categories along with that 

 assumed in the Sandia Siting Study.....................................................19 

8. Comparison of frequency distributions (CCDFs) of early fatalities 

 predicted by the NUREG-1150 Study with those reported in the 

 Reactor Safety Study (Reproduced from the Report to the Congress 

 from the Presidential Commission on Catastrophic Nuclear accidents, 

 August 1990, Appendix B).................................................................21 

 

9. Comparison of frequency distributions (CCDFs) of latent cancer 

 fatalities predicted by the NUREG-1150 Study with those reported 

 in the Reactor Safety Study (Reproduced from the Report to the 

 Congress from the Presidential Commission on Catastrophic 

 Nuclear accidents, August 1990, Appendix B).......................................21 

10. Frequency of a release that can result in one or more early fatalities 

 due to internal events for NUREG-1150 Plants.....................................22 

11. Comparison of frequency distribution (CCDF) of iodine release 

 predicted by NUREG-1150 Study for Surry with that obtained from 

 the representative source terms..........................................................25
 v

12. Comparison of frequency distribution (CCDF) of population dose 

 (person-rem) to entire region at Surry predicted by the 

 NUREG-1150 Study with those obtained by using representative 

 source terms...................................................................................26 

13. Impact of latest understanding of DCH on probability of accident 

 progression bins for internal events at Zion..........................................29 

14. Impact of latest understanding of drywell melt-through on 

 probability of accident progression bins for internal events at 

 Peach Bottom................................................................................30 

15. Elements of the Proposed Approach to Update the Results of Earlier 

 Level-3 PRAs for Comparison with SOARCA ............................................33 

16. Comparison of frequency distribution (CCDF) of iodine release predicted by 

 NUREG-1150 Study for Surry with that obtained from the representative 

 source terms and their revised frequencies based on results of 

 SPAR model and recent insights on early containment failure mechanisms..........37
 vi

TABLES

1. Summary of Release Categories in Reactor Safety Study..................6 

2. Brief Descriptions of the Characteristics of the Accident Groups..........9 

3. Siting Source Terms (SSTs)........................................................11 

4. Comparison of Conditional Mean Consequences Predicted 

 for SSTs .................................................................................13 

 

5. Characteristics of Surry Release Categories, Internal Events..............24 

6. Radionuclide Release Characteristics into Environment 

 for Surry .................................................................................24 

7. Impact of Latest Understanding of DCH on Probability of 

 Accident Progression Bins for Internal Events at Zion.......................28 

 

8. Impact of Latest Understanding of Drywell Melt-Through on 

 Probability of Accident Progression Bins for Internal Events 

 at Peach Bottom.......................................................................31 

9. Revised mean conditional probability of accident progression bins 

 at Surry ...................................................................................34 

10. Frequencies and Magnitudes of Iodine Releases for 

 Representative Source Terms for Surry (Internal Initiators)................36
 vii

ABBREVIATIONS 

ACRS Advisory Committee on Reactor Safeguards 

AEC Atomic Energy Commission 

ATWS Anticipated Transient without Scram 

BNL Brookhaven National Laboratory 

BWR Boiling Water Reactor 

CCDF Complementary Cumulative Distribution Function 

CCI Core-Concrete Interaction 

CF Containment Failure 

CV Containment Venting 

DBA Design Basis Accident 

DCH Direct containment Heating 

ECF Early Containment Failure 

EDO Executive Director for Operations 

EPRI Electric Power Research Institute 

ESF Engineered Safety Feature 

IPE Individual Plant Examination 

LCF Late Containment Failure 

LLNL Lawrence Livermore National Laboratory 

LOCA Loss-of-Coolant Accident 

LOSP Loss of Offsite Power 

LWR Light Water Reactor 

MACCS MELCOR Accident Consequence Code System 

NCF No Containment Failure 

NEI Nuclear Energy Institute 

NRC Nuclear Regulatory Commission 

NSSS Nuclear Steam Supply System 

PDS Plant Damage State 

PRA Probabilistic Risk assessment 

PWR Pressurized Water Reactor 

QHOs Quantitative Health Objectives 

RCS Reactor Coolant System 

RPV Reactor pressure Vessel 

RST Representative Source Term 

SAMGs Severe Accident Management Guidelines 

SBO Station Blackout 

SGTR Steam Generator Tube Rupture 

SNL Sandia National Laboratories 

SOARCA State-Of-the-Art Reactor Consequence Analyses 

SRM Staff Requirements Memorandum 

SRV Safety Relief Valve 

SST Siting Source Term 

STCP Source Term Code Package 

VB Vessel Breach 

WWF Wet Well Failure 
 1

1 INTRODUCTION 

The probability and offsite consequences 

of severe reactor accidents have been the 

subject of considerable interest and study 

since the earliest days of reactor 

development. 

The first estimates of consequences of 

severe accidents were published in the 

1957 U.S. Atomic Energy Commission 

report (WASH-740) [1], “Theoretical 

Possibilities and Consequences of Major 

Accidents in Large Nuclear Power 

Plants.” This study was an attempt to 

provide upper bounds of the potential 

public hazards resulting from certain 

severe hypothetical accidents. 

Conservative values were used for many 

factors influencing the magnitude of the 

estimated accident consequences. At the 

time, the technology and the state-ofknowledge of severe accidents had not 

progressed to the point where it was 

possible to use quantitative techniques to 

estimate the probabilities of such 

accidents. However, there was a general 

agreement that the probability of 

occurrence of severe accidents in nuclear 

power reactors was exceedingly low. The 

following is quoted from the March 22, 

1957 letter, from Harold S. Vance, Acting 

Chairman, Atomic Energy Commission, to 

Carl T. Durham, Chairman, Joint 

Committee on Atomic Energy, Congress 

of the United States, transmitting an 

advance copy of the WASH-740 report. 

“As to the probabilities of major 

 reactor accidents, some experts held 

 that numerical estimates of a 

 quantity so vague and uncertain as 

 the likelihood of occurrence of 

 major reactor accidents have no 

 meaning. They declined to 

 express their feeling about this 

 probability in numbers. Others, 

 though admitting similar uncertainty, 

 nevertheless ventured to express 

 their opinions in numerical terms. 

 Estimations so expressed of the 

 probability of reactor accidents 

 having major effects on the 

 public ranged from a chance of one 

 in 100,000 to one in a billion per 

 year for each large reactor. 

 However, whether numerically 

 expressed or not, there was no 

 disagreement in the opinion that the 

 probability of major reactor 

 accidents is exceedingly low. 

 ......To reduce the matter of 

 assumed hazards to comparative 

 numbers, let us take the most 

 pessimistic assumptions used and 

 apply them to a case of 100 

 power reactors in operation in the 

 United States. Under these 

 assumptions, the chances of a 

 person being killed in any year by a 

 reactor accident would be less than 

 one in 50 million. By contrast, the 

 present odds of being killed in any 

 year by an automobile 

 accident in the United States stands 

 at about one in 5,000.” 

Since the publication of WASH-740 

report, several systematic studies have 

been made to search out a large 

spectrum of accidents and to use 

quantitative techniques to estimate the 

probabilities, radionuclide release 

characteristics (source terms), and 

potential offsite health consequences. 

Such studies include the NRC’s WASH1400 (1975) [2] and NUREG-1150 (1990) 

[3], as well as industry-sponsored 

probabilistic risk assessments (PRAs) 

such as those for Zion (1981) [4], Indian 

Point (1982) [5], Millstone 3 (1983) [6], 

and Seabrook (1983) [7]. 

In 1982, Sandia National Laboratories 

(SNL) performed a study of technical 

aspects of siting for nuclear power 
 2

reactors. The results of this study, also 

known as Sandia Siting Study, were 

published in NUREG/CR-2239 [8], 

"Technical Guidance for Siting Criteria 

Development." This study used five 

generic source terms for analyzing the 

consequences and socio-economic 

impacts of possible plant accidents at 91 

existing or proposed reactor sites. These 

source terms were derived from the 

Reactor Safety Study (WASH-1400) and 

subsequent evaluations. 

Since the publication of the Sandia Siting 

Study, many events have brought a new 

focus to this study and its results. Despite 

accepted arguments that the results of 

this study are overly conservative and do 

not reflect current state-of-the-art in 

evaluating severe accident progression 

and offsite consequences, the results, in 

terms of predicted offsite early fatalities 

and latent cancers, have often been 

quoted by outside organizations to 

illustrate the potential consequences of a 

severe accident at a commercial nuclear 

power plant. 

The NRC staff is currently implementing 

its plan for developing state-of-the-art 

reactor consequence analyses [9]. This 

work will: (1) evaluate and update, as 

appropriate, analytical methods and 

models for realistic evaluation of severe 

accident progression and offsite 

consequences; (2) develop state-of-theart reactor consequence assessments of 

severe accidents; and (3) identify 

mitigative measures that have the 

potential to significantly reduce risk or 

offsite consequences. These analyses 

include external events; consideration of 

all mitigative measures, including the 

newly required extreme damage state 

mitigative guidelines (B.5.b); state-of-theart accident progression modeling, based 

on 25 years of research, to provide a best 

estimate for accident progression, 

containment performance, time of 

release, and fission product behavior; 

more realistic offsite dispersion modeling; 

and site-specific evaluation of public 

evacuation based on updated emergency 

plans. 

In a Staff Requirements Memorandum 

(SRM) dated April 14, 2006 [10], the 

Commission stated that the staff’s 

proposal to examine significant 

radiological release scenarios having 

estimated likelihoods of one in a million or 

greater per year is an appropriate initial 

focus. The Commission also stated that 

“in applying a screening radiological 

release frequency of 10-6 per reactor year, 

the staff should be careful to define 

release groupings such that release 

characteristics are representative of 

scenarios binned into those groups. 

However, where possible, the groups 

should also be sufficiently broad to be 

able to include the potentially risksignificant but lower-frequency scenarios 

(for example, the interfacing systems 

LOCA scenarios that bypass the 

containment).” 

In the April 14, 2006 SRM [10], the 

Commission specifically instructed the 

staff to “work with the ACRS on technical 

issues such as identification of accident 

scenarios to be evaluated, evaluation of 

source terms, credit for operator actions 

or plant mitigation systems, modeling of 

emergency preparedness, modeling of 

offsite consequences, and definition and 

characterization of analysis uncertainty.” 

In an April 2, 2007 SRM [11], the 

Commission directed the staff to “reduce 

the initial scope of this effort [SOARCA] to 

not more than eight plants representing a 

spectrum of plant vendors and 

technologies.” The Commission also 

directed the staff to “conduct the first 

assessments on a subset of the eight 

plants, for example a selected BWR and 

PWR plant, in order to resolve issues 

associated with the integration of methods 

and resolve details associated with 

simulation of plant systems and 

procedures.” 
 3

In its February 25, 2008 report to the 

Commission concerning the SOARCA 

Project [12], the ACRS recommended that 

“as a minimum, a limited set of updated 

Level-3 PRAs for the SOARCA pilot 

plants be performed to benchmark the 

consequence analyses and provide useful 

information to the Commission in deciding 

whether to proceed with a full set of 

consequence analyses.” The ACRS 

further noted that “examination of the 

Level-3 PRA results for the SOARCA pilot 

plants may identify suitable Level-1 event 

scenario screening criteria and simplifying 

assumptions that could be used to 

develop a defensible, simplified 

approach.” 

In a letter dated April 7, 2008 [13], the 

Executive Director for Operations (EDO)

responded to the ACRS report of 

February 25, 2008 on SOARCA, 

indicating that the staff did not agree with 

the ACRS recommendation that a limited 

set of Level-3 PRAs be performed to 

benchmark the SOARCA approach

developed by the staff. In its April 21, 

2008 response to the EDO [14], the 

ACRS noted that the Committee 

continues “to believe that the credibility of 

the SOARCA Project cannot rely on 

confidence in the judgment of the staff 

and on a novel analysis procedure that 

differs substantially from previous stateof-the-art analyses of the consequences 

of severe reactor accidents.” The ACRS 

further noted that “without including 

benchmark analyses similar in scope, it 

will be difficult to demonstrate 

convincingly that reductions in 

consequences that might be indicated by 

the SOARCA results reflect the impact of 

enhancements in plant design and 

operation, and improvements in 

calculation methods for accident 

progression and consequence analysis, 

rather than changes in the scope of the 

calculation.”

In a June 26, 2008 SRM [15], resulting 

from the June 5, 2008 meeting with the 

ACRS, the Commission directed the staff 

to “continue working to address 

Committee concerns, such as with 

SOARCA, ... and, as necessary and 

appropriate, provide timely policy decision 

papers to the Commission to resolve any 

disagreements.” 

This report has been prepared for use by 

the ACRS in its continued dialogue with 

the staff regarding the feasibility of using 

a simplified, yet systematic and 

defensible, approach to update results 

from earlier Level-3 PRAs such as the 

NUREG-1150 Study for comparison with 

aspects of SOARCA results. 

The report begins with an overview of 

major contributions to consequence 

assessment to provide historical 

perspectives and insights on previous 

state-of-the-art analyses of the 

consequences of severe reactor 

accidents. It then discusses how the 

results and insights from NUREG-1150 

study and integrated risk assessment for 

La Salle together with recent advances in 

understanding of severe accident 

phenomenology and containment failure 

mechanisms could be used to update the 

results of such earlier Level-3 PRAs for 

comparison with aspects of SOARCA 

results. 
 4

2. MAJOR NRC-SPONSORED ASSESSMENTS OF 

REACTOR ACCIDENT CONSEQUENCES 

 

2.1 Reactor Safety Study 

(WASH-1400) 

The Reactor Safety Study (WASH-1400) 

[2], was the first systematic attempt to 

provide realistic estimates of risk to the 

public from potential accidents in 

commercial nuclear power plants. This 

1975 study included analytical methods 

for determining both the probabilities and 

consequences of various accident 

scenarios. Event trees and fault trees 

were used to define important accident 

sequences and to quantify the reliability of 

engineered safety systems. Detailed 

investigations were performed to predict 

fission product release from the reactor 

fuel and the subsequent transport and 

behavior within the reactor coolant and 

containment systems. Calculations were 

performed for a number of accident 

sequences and the results of these 

calculations were used to define a series 

of release categories into which all of the 

identified accident sequences were 

placed. 

Two specific reactor designs were 

analyzed in WASH-1400: Peach Bottom 

Atomic Power Station, a boiling water 

reactor (BWR) with a Mark I containment 

and Surry, a 3-loop pressurized water 

reactor (PWR) with a subatmospheric 

containment. Nine PWR release 

categories and five BWR release 

categories were developed in the Reactor 

Safety Study. Each release category was 

represented by its release frequency and 

several other parameters that described 

the radionuclide release characteristics 

(source term). Table 1 presents a 

summary of release categories defined in 

the Reactor Safety Study. These release 

categories represent the full spectrum of 

possible accident progression and 

containment failure modes based on the 

state of knowledge and understanding of 

severe accidents at that time. For 

example, the dominant contributing 

sequence to PWR release category 1 is a 

transient event involving loss of offsite AC 

power with a failure to recover either 

onsite or offsite AC power within about 3 

hours. This sequence considers that an 

in-vessel steam explosion (“alpha mode” 

failure) as the early containment failure 

mode (see Chapter 3 of this report for a 

more recent reassessment of this issue). 

It was also assumed that during steam 

explosion, the volatile fission products 

would be finely dispersed and air 

oxidation would enhance the magnitude 

of the radioactivity releases from the 

failed containment. 

The Reactor Safety Study provided 

insights into expected individual risk from 

characteristic types of releases under 

average population and meteorological 

conditions at the 68 sites at which the first 

100 reactors expected to be operating by 

about 1980. The risk was presented by 

the probability and magnitude of seven 

different consequence measures. These 

consequence measures were early 

fatalities (death within approximately one 

year after a potential accident), early 

illnesses, thyroid nodules, latent cancer 

fatalities, genetic effects, land 

contamination, and property damage 

costs. A major conclusion of the Reactor 

Safety Study was that the low probabilityhigh consequence accidents involving 

core meltdown, containment failure, and 

failure of engineered safety features 

dominated the risk to public. 

Figures 1 and 2 show the likelihood and 

number of fatalities from both nuclear and 

a variety of non-nuclear accidents, 

reported in the Executive Summary of 
 5

WASH-1400 report. These results 

indicated that non-nuclear events were 

about 10,000 times more likely to produce 

large numbers of fatalities than nuclear 

plants. It should be noted that the 

societal risks are stated as 

complementary cumulative distribution 

functions per year in Figures 1 and 2. The 

complementary cumulative distribution 

function (CCDF) shows the frequency that 

a consequence will exceed a given 

magnitude. Also note that the fatalities 

shown in Figures 1 and 2 are those that 

would be predicted to occur within a short 

period of time after the potential reactor 

accident. This was done to provide a 

consistent comparison to the non-nuclear 

events that also cause fatalities in the 

same time frame. 

The WASH-1400 report stimulated a great 

deal of debate after its release. In June 

1976, the Committee on Interior and 

Insular Affairs of the U.S. House of 

Representatives, chaired by 

Representative Morris Udall, held 

hearings on the findings of the Reactor 

Safety Study. These hearings found that 

the Reactor Safety Study seemed to be 

misleading in the certainty and 

comprehensiveness of its conclusions 

[16]. Rep. Udall suggested that an outside 

review panel be formed to take a closer 

look at how the study arrived at its 

conclusions [16]. The NRC then asked Dr 

Harold Lewis of the University of 

California-Santa Barbara to chair an 

independent review group, which 

produced what is now known as the 

“Lewis report” [17]. 

The Lewis report concluded that the 

WASH-1400 study was overall a 

“conscientious and honest effort", an 

“important advance” over earlier 

quantitative analysis of reactor safety, and 

with a “sound methodology” that should 

be used more widely by the NRC. Among 

the shortcomings that the Lewis 

Committee identified in the Reactor 

Safety Study was the lack of scrutability of 

the calculation/analysis process. The 

Lewis report was particularly critical of the 

Executive Summary of the WASH-1400 

report for being “a poor description of the 

contents of the report” and for not 

adequately indicating the full extent of the 

consequences of, and the uncertainties in 

the probabilities of, reactor accidents. For 

this reason, the NRC withdrew its 

endorsement of the Executive Summary 

although it did not repudiate the study 

itself. 

 
 6

Table 1 Summary of Release Categories in Reactor Safety Study 

(WASH-1400, NUREG-75/014, p. v-4) [2] 
 7

Figure 1: Frequency of Fatalities due to Man-Caused Events (WASH-1400) [2]
 8

Figure 2: Frequency of Fatalities due to Natural Events (WASH-1400) [2]
 9

2. 2 Post TMI-2 Review of 

Source Term Technical 

Bases and Sandia Siting 

Study 

Following the publication of WASH-1400 

and the accident at Three Mile Island, 

Unit 2 (TMI-2), work was initiated to 

evaluate the predictive methods for 

calculating fission product release and 

transport. The results of this evaluation 

are contained in NUREG-0772 [18], 

“Technical Bases for Estimating Fission 

Product Behavior during LWR Accidents.” 

The development of this report was 

prompted, in part, by the December 21, 

1980 letter, from the Nuclear Safety 

Oversight Committee to President Carter 

[19], noting the questions raised at the 

time regarding iodine release and 

recommending that they should be 

answered by analyses and 

experimentation on an expedited basis. 

The NUREG-0772 evaluation resulted in 

several conclusions that represented 

significant departures from the Reactor 

Safety Study assumptions including the 

conclusion that cesium iodide (CsI) would 

be the expected predominant iodine 

chemical form under most postulated light 

water reactor (LWR) accident conditions. 

The potential impact of the NUREG-0772 

findings on reactor regulation was also 

examined and the results were 

documented in NUREG-0771 [20]. These 

studies formed the basis for the 

designation of five accident groups as 

being representative of the spectrum of 

potential accident conditions. Brief 

descriptions of characteristics of the five 

accident groups are presented in Table 2. 

Table 2 Brief Descriptions of the Characteristics of the Accident Groups 

(NUREG-0771, p. 8) [20] 
 10

Figure 3: Comparison of frequency distribution (CCDF) of iodine release from Reactor 

Safety Study release categories for Surry with that assumed for Siting Source Terms 

For the purpose of decisionmaking in 

such areas as siting and emergency 

response, NRC developed a generic set 

of radiological releases characterized as 

Siting Source Terms to represent the five 

accident groups. Table 3 presents the 

NRC-defined Siting Source Terms 

(denoted as SST1-5). The idea was that 

by adjusting the probabilities associated 

with each of the five source terms, the set 

could be made to approximately represent 

any current LWR design. Detailed 

probabilistic risk assessments (PRAs) 

were not performed for all reactors. 

However, based on the available PRAs at 

the time, NRC suggested that 

representative frequencies for the SST1, 

SST2, and SST3 were 1x10-5, 2x10-5, and 

1x10-4 respectively. Figure 3 provides a 

comparison of the frequency distributions 

in the form of CCDF of iodine release 

from the Reactor Safety Study release 

categories for Surry with iodine release 

assumed in SST1, SST2, and SST3. 

Note that the frequencies shown in Figure 

3 are exceedance frequencies (CCDFs) 

rather that frequency assigned to each 

source term. It should also be noted that 

iodine is not the sole radionuclide of 

importance to consequence analyses. 

Other fission product species notably 

cesium, tellurium, and ruthenium have 

also significant impact on the 

consequences of severe accidents. 

However, iodine is the dominant 

contributor to early fatalities and its 

frequency of release is presented 

throughout this report as an illustrative 

example. 

 

1.00E-08

1.00E-07

1.00E-06

1.00E-05

1.00E-04

1.00E-03

1.00E-05

1.00E-04

1.00E-03

1.00E-02

1.00E-01

1.00E+00

 Release Fraction

Exceedance Frequency

SST1 

SST2 

SST3 

♦ PWR Release Categories (WASH-1400) 

□ Siting Source Terms (NUREG/CR-2239) 
 11

Table 3 Siting Source Terms (SSTs) 

(NUREG/CR-2239, p. 2-13) [8] 
 12

Sandia National Laboratories (SNL) 

performed a study of the technical 

aspects of siting of nuclear power 

reactors. The results of this study were 

published in NUREG/CR-2239 [8], also 

known as the Sandia Siting Study. This 

study used Siting Source Terms at 91 

existing or proposed reactor sites. 

Consequence analyses were performed 

for each of the five source terms. Table 4 

compares the relative magnitude 

(normalized to 100 for source term SST1) 

of the mean values (using approximately 

100 sampled weather sequences) of 

selected consequences, given the 

occurrence of each of the five SSTs. 

These calculations assumed an 1120 

MWe PWR, population distribution (based 

on the 1970 census) and wind rose for 

Indian Point, New York City meteorology, 

and Summary Evacuation of persons 

within 10 miles. These results indicate 

that the mean consequences calculated 

for the SST1 release exceed those for the 

SST2 by 1 to 4 orders of magnitude and 

for SST3, SST4, and SST5 by 4 to 7 

orders of magnitude. 

As stated above, the Sandia Siting Study 

assumed an 1120 MWe PWR reactor at 

each of the 91 sites. Thus, its results were 

not directly representative of the actual 

potential consequences of an accident at 

the site. In a subsequent study, the 

results of NUREG/CR-2239 were scaled 

linearly by power level to derive an 

approximation of potential consequences 

for the actual reactor at the site. The 

results of this study were published in 

NUREG/CR-2723 [21]. This report also 

examined the financial consequences of 

potential accidents at nuclear power 

plants. Such information on the range of 

consequences was thought to be useful in 

a reevaluation of the liability limits of the 

Price-Anderson Act. 
 13

Table 4, Comparison of Conditional Mean Consequences Predicted for SSTs 

(NUREG/CR-2239, p. 2-14) [8] 
 14

2.3 NUREG-1150 Study 

In the 1980s, a substantial research 

program on severe accident 

phenomenology was initiated. Updated 

computational models for severe accident 

analysis were developed and published in 

BMI-2104 [22]. A technical reassessment 

of severe accident source term 

technology for U.S. Light Water Reactors 

(LWRs) was published in NUREG-0956 

[23]. This reassessment involved 

reviewing experimental and analytical 

results from severe accident research 

programs sponsored by the NRC and the 

nuclear industry. As a result of these 

activities, the Source Term Code Package 

(STCP) [24] was developed as an 

integrated tool for source term evaluation. 

Subsequently, the severe accident 

analysis code MELCOR [25] was 

developed based, in part, on the STCP. 

The NUREG-1150 study [3] was a major 

effort to put the insights gained from the 

research on system behavior and 

phenomenological aspects of severe 

accidents into a risk perspective. An 

important characteristic of this study was 

the inclusion of the uncertainties in the 

calculations of core damage frequency 

and risk due to incomplete understanding 

of reactor systems and severe accident 

phenomena. The elicitation of expert 

judgment was used to develop probability 

distributions for many accident 

progression, containment loading, 

structural response, and source term 

issues. Five specific commercial nuclear 

power plants were analyzed in NUREG1150: Surry, a 3-loop Westinghouse PWR 

with a subatmospheric containment; Zion, 

a 4-Loop Westinghouse PWR with a large 

dry containment; Sequoyah, a 4-loop 

Westinghouse PWR with an icecondenser containment; Peach Bottom, a 

BWR-4 reactor with a Mark I containment; 

and Grand Gulf, a BWR-6 reactor with a 

Mark III containment. For Surry and 

Peach Bottom, the study included the 

analyses of both internal and external 

events. 

Figure 4 shows the internal core damage 

frequency calculated in NUREG-1150 

study for Surry as compared to that from 

the Reactor Safety Study and those 

reported in Individual Plant Examinations 

(IPEs) for PWRs. The IPE results 

presented are the range for the point 

estimate frequencies reported in NUREG1560 [26] and do not include estimates of 

uncertainty. Also presented in Figure 4 

are the recent results of the SPAR model 

for Surry obtained as part of the SOARCA 

Project. As shown in Figure 4, the recent 

results of the SPAR model for Surry 

indicate a core damage frequency that is 

one order of magnitude lower than the 

mean values obtained in the NUREG1150 study. The staff has yet to report on 

how much of this difference is due to 

hardware modifications and procedural 

improvements that had been implemented 

by the plant since the NUREG-1150 study 

and how much of this difference is due to 

advances in the PRA state of the art. 

Containment performance plays an 

important role in the assessment of the 

risk associated with severe accidents. 

The primary concerns for containment 

performance are how well the 

containment can withstand the pressure 

and temperature loads associated with 

severe core damage accidents and 

whether the containment is bypassed. 

For scenarios in which containment 

integrity is maintained, fission product 

release is small. For those scenarios 

leading to containment failure, fission 

product release depends on the timing as 

well as the size of the break in
 15

Figure 4: Surry frequency of core damage from internal events calculated in WASH1400, NUREG-1150, the IPEs, and the recent result of the SPAR model for Surry 

(performed in support of SOARCA) 

containment. Early containment failure 

can be important because it tends to 

result in shorter warning times for 

initiating public protective measures, and 

because radionuclide releases would 

generally be more severe than those from 

late containment failure. The mode of 

containment failure (i.e., gross failure 

versus leakages through failure of 

penetrations) influences the amount of 

radioactive materials inside the 

containment that would be released to the 

environment. For the NUREG-1150 

study, accident progression and 

containment performance were analyzed 

using a single accident progression event 

tree developed for each plant, which was 

evaluated with the EVNTRE code [27]. 

The accident progression event trees 

made extensive use of the available 

severe accident computer code 

calculations and experimental results. As 

noted in NUREG-1150, “computer 

analyses cannot, in general, be used 

directly and alone to calculate branching 

probabilities in the accident progression 

event tree. Since the greatest source of 

uncertainty is typically associated with the 

modeling of severe accident phenomena, 

the results of a single computer run 

(which uses a specific model) do not 

characterize the branching uncertainty.” It 

was therefore necessary to use sensitivity 

studies, uncertainty studies, and expert 

judgment to characterize the likelihood of 

alternative events that affect the course of 

an accident. The elicitation of expert 

judgment was used to develop probability 

distributions for many accident 

progression, containment loading, and 

structural response issues. 

The NUREG-1150 results of the 

containment analyses for Surry and 

Peach Bottom are summarized in Figures 

5 and 6, respectively. Figure 5 displays 

the Surry NUREG-1150 results for the 

conditional probabilities of seven 

containment-related accident progression 

bins (e.g., vessel breach (VB), alpha, 

early containment failure) for each of 

seven plant damage states (e.g., LOSP)

WASH-1400 NUREG-1150 PWRs (IPEs)

1E-06

1E-05

1E-04

1E-03

median 

median

mean 

SPAR Model (SOARCA) 

5th 

95th 
 16

Figure 5: Conditional Probability of Accident Progression Bins at Surry 

(NUREG-1150, p. 3-12) [3] 
 17

Figure 6: Conditional Probability of Accident Progression Bins at Peach Bottom 

(NUREG-1150, p. 4-13) [3]
 18

As it is indicated in Figure 5, on a plant 

damage state frequency weighted 

average, the conditional mean probability 

of early containment failure from internally 

initiated accidents for Surry was found to 

be very small (about 1 percent) in the 

NUREG-1150 Study. The primary 

mechanism leading to early containment 

failure were loads from Direct 

Containment Heating (DCH) for accident 

sequences with high reactor coolant 

system (RCS) pressures at vessel breach 

and in-vessel steam explosion for 

sequences with low RCS pressure at 

vessel breach. The NUREG-1150 

estimate of the likelihood of the early 

containment failure is substantially lower 

than the WASH-1400 result even though 

the phenomena of DCH had not been 

identified at the time of the Reactor Safety 

Study. In addition to the lower assumed 

containment capacity, the prediction of 

containment loading in the Reactor Safety 

Study was unrealistically high. It should 

be noted that the characterization of 

containment performance in the Reactor 

Safety Study was simplistic in comparison 

to the NUREG-1150 Study. 

Figure 6 displays the Peach Bottom 

NUREG-1150 results for the conditional 

probabilities of 10 containment-related 

accident progression bins (e.g., vessel 

breach (VB), late wet well failure (WWF)) 

for each of six plant damage states such 

as station blackout (SBO). As it is 

indicated in Figure 6, on a plant damage 

state frequency weighted average, the 

conditional mean probability of early 

drywell failure from internally initiated 

accidents for Peach Bottom was found to 

be about 52 percent. For Peach Bottom, 

although the early containment failure 

was found to be quite likely in the 

NUREG-1150 Study, the mechanism 

resulting in failure as well as the modes 

and locations of failure were quite 

different from those considered in the 

Reactor Safety Study. In the Reactor 

Safety Study, the most likely failure 

location was assessed to be the upper 

portion of the suppression pool. In the 

NUREG-1150 Study, other mechanisms 

of containment failure, such as direct 

attack of drywell wall by molten core 

debris, were found to be also important. 

The dominant location of overpressure 

failure in the NUREG-1150 Study was 

assessed to be the lifting of drywell head 

by stretching the head bolts. 

One of the major activities of the NUREG1150 study was the development of 

fission product source terms for a 

spectrum of accident conditions. The 

source terms were calculated using a 

simplified parametric algorithm. The 

parametric equations describe the source 

terms as the product of release fractions 

and transmission factors at successive 

stages in the accident progression for a 

variety of release pathways, a variety of 

accident progressions, and nine classes 

of radionuclides. Probability distributions 

for important parameters were developed 

using the elicitation of expert judgment to 

augment the analytical results to reflect 

model uncertainties. The source term 

analysis resulted in characterizing 

thousands of source terms (20,000 for 

Surry) associated with tens of plant 

damage states, hundreds of accident 

progression bins, and the variation in 

source term phenomenological issues 

which were included in the propagation of 

uncertainties. In Figure 7, frequency 

distribution (CCDF) of iodine release 

predicted by the NUREG-1150 Study for 

Surry is compared with that obtained from 

the Reactor Safety Study release 

categories along with that assumed in 

Sandia Siting Study. As shown in Figure 

7, the mean frequency of iodine release is 

generally lower than that obtained in the 

Reactor Safety Study. In particular, the 

mean frequency of high (more than 10 

percent) release of iodine predicted in the 

NUREG-1150 Study is more than one 

order of magnitude less than that 

obtained in the Reactor Safety Study. 
 19

Figure 7: Comparison of frequency distribution (CCDF) of iodine release predicted by 

 NUREG-1150 study for Surry with that obtained from the Reactor Safety 

 Study release categories along with that assumed in the Sandia Siting 

 Study 

It should also be noted that the NUREG1150 Study calculated the mean 

frequency of iodine releases similar to 

one assumed in SST1 to be less than 10-6

(the screening frequency used in 

SOARCA). 

In NUREG-1150, the source terms were 

grouped according to the warning time 

and their potential to cause early or latent 

cancer fatalities. Through this 

“partitioning” process, the large numbers 

of calculated radioactive releases were 

collected into a small set of source term 

groups (30 to 60 in number for each 

plant). This set of groups was then used 

in the offsite consequence calculations. 

The consequence measures, early 

fatalities, population dose (person-rem), 

and latent cancer fatalities, were 

calculated for each source term group by 

the MAACS code [28]. The output of 

MACCS for each source term group is a 

distribution of the consequences, 

conditional on occurrence of the source 

term, which incorporates the uncertainty 

(variability) due to weather as well as the 

uncertainty in the underlying health (doseresponse) models. 

Comparisons of frequency distributions 

(CCDFs) of early and latent cancer 

fatalities predicted by the NUREG-1150 

Study with those reported in the Reactor 

1.00E-08

1.00E-07

1.00E-06

1.00E-05

1.00E-04

1.00E-03

1.00E-05

1.00E-04

1.00E-03

1.00E-02

1.00E-01

1.00E+00

 Release Fraction

Exceedance Frequency

SST1 

SST2 

SST3 

♦ PWR Release Categories (WASH-1400) 

□ Siting Source Terms (NUREG/CR-2239) 

NUREG-1150 Study (Surry), Mean 
 20

Safety Study are presented in Figures 8 

and 9, respectively. These figures are 

reproduced from an August 1990 Report 

to the Congress from the Presidential 

Commission on Catastrophic Nuclear 

accidents [29]. As shown in Figures 8 

and 9, the consequences calculated in the 

NUREG-1150 study are significantly lower 

than those obtained in the Reactor Safety 

Study. 

The NUREG-1150 risk analyses were 

performed to provide the frequency of a 

large release, defined at the time as the 

one that could cause one or more early 

fatalities. The results of such evaluations 

for NUREG-1150 plants are shown in 

Figure 10, reproduced from NUREG-1150 

(Vol. 1, pp. 13-10). As shown in Figure 

10, the mean frequency of a release, due 

to internal events, to cause one or more 

early fatalities is less than 10-6. 

In spite of the plant specific nature of the 

NUREG -1150 quantitative results (e.g., 

core damage, frequency and offsite 

consequences), this study provides 

valuable insights into severe accident 

phenomenological issues and associated 

state-of-knowledge uncertainties which 

are very useful to the study of plants with 

similar NSSS and containment designs. 

The insights from the NUREG-1150 Study 

have been used in several areas of 

reactor regulation including the 

development of alternative radiological 

source terms for evaluating design basis 

accidents at nuclear reactors. In 1995, the 

NRC published NUREG-1465 [30], which 

defined an alternative accident source 

term for regulatory applications. The 

release fractions for the alternative 

accident source terms were derived from 

the insights and simplifications of the 

NUREG-1150 source term analyses 

documented in NUREG/CR-5747 [31]. 

The NRC is also using source terms 

derived from the NUREG-1150 Study in 

an effort to revise criteria for protective 

action recommendations for severe 

accidents [32]. 
 21

Figure 8: Comparison of frequency distributions (CCDFs) of early fatalities predicted by 

 the NUREG-1150 study with those reported in the Reactor Safety Study 

(Reproduced from the Report to the Congress from the Presidential Commission on 

Catastrophic Nuclear accidents, August 1990, Appendix B) [29] 

Figure 9: Comparison of frequency distributions (CCDFs) of latent cancer fatalities 

 predicted by the NUREG-1150 study with those reported in the Reactor 

 Safety Study 

(Reproduced from the Report to the Congress from the Presidential Commission on 

Catastrophic Nuclear accidents, August 1990, Appendix B) [29] 
 22

Figure 10: Frequency of a release that can result in one or more early fatalities due to 

internal events for NUREG-1150 Plants 

(Reproduced from NUREG-1150, Vol. 1, pp. 13-10) [3] 
 23

2.4 Reassessment of 

Selected Factors 

Affecting Siting 0f 

Nuclear Power Plants 

In 1997, Brookhaven National Laboratory 

performed a series of probabilistic 

consequence assessments in support of 

an effort to reassess siting criteria [33]. 

This study took into account the insights 

gained from the NUREG-1150 Study [3] 

and the integrated risk assessment for 

LaSalle1

 [34] and examined 

consequences in a risk-based format 

consistent with the quantitative objectives 

of the NRC’s Safety Goal Policy. 

The approach taken in this study was to 

use sets of representative source terms 

(for internal events) and their associated 

frequencies for LaSalle and each of the 

NUREG-1150 plants. These source 

terms were applied to a set of 

hypothetical sites with various population 

densities and locations of urban centers 

which would encompass most of the 

existing reactor sites. The site 

meteorology was selected such that, from 

a meteorological standpoint, the 

calculation of consequences would be 

likely to bound 80 percent of existing 

reactor sites. Sensitivity calculations 

were also performed to evaluate the 

effects of emergency protective action 

assumptions on the risk of prompt fatality 

and latent cancer fatality, as well as 

 

1

 It should be noted that the integrated risk 

assessment for LaSalle, a BWR with Mark II 

containment, was begun before the 

NUREG-1150 analysis and the LaSalle 

integrated risk assessment program 

supplied the NUREG-1150 Study with many 

analysis methods including methods for 

handling and propagating statistical 

uncertainties in an integrated way through 

the entire analysis, and BWR thermalhydraulic models which were adapted for the 

Peach Bottom and Grand Gulf analyses. 

population relocation. The study 

concluded that the prompt and latent 

fatality risks at all generic sites met the 

Quantitative Health Objectives (QHOs) of 

the NRC’s Safety Goal Policy by margins 

ranging from one to more than three 

orders of magnitude. Results from the 

study also indicated that both the quantity 

of radioactivity released during a severe 

accident, as well as the likelihood of 

release, were considerably lower than 

those predicted by earlier studies (e.g., 

the Sandia Siting Study of 1982). For this 

reason, the severe accident risks 

estimated in earlier studies were 

concluded to be unduly pessimistic, 

based on a revised understanding of 

severe accidents. 

Developing representative source terms 

involved the determination of a set of 

reactor accident progression groups, the 

dominant plant damage states and the 

associated release characteristics for 

each reactor design which represented 

the full spectrum of severe accidents. A 

small set of source terms (4 to 7 for each 

plant) was developed [35] by considering 

release categories which accounted for a 

spectrum of possible timing and modes of 

containment failure. For each containment 

failure mode, the source terms were 

selected based on the dominant accident 

progression characteristics leading to the 

containment failure. The magnitudes of 

releases for each release category were 

obtained by using the mean values of the 

probability distributions of source term 

parameters used in NUREG-1150 and the 

LaSalle integrated risk assessment 

studies. For example, the representative 

source terms for Surry consist of four 

release categories as shown in Table 5. 

These release categories correspond to 

early containment failure (ECF), late 

containment failure (LCF), no containment 

failure, and containment bypass due to 

interfacing-system LOCA (Event V). 
 24

Table 5: Characteristics of Surry Release Categories, Internal events 

(NUREG/CR-6295, pp 3-19) [33] 

Table 6: Radionuclide Release Characteristics into Environment for Surry, Internal Events 

 (NUREG/CR-6295, pp3-19) [33] 
 25

The results of the NUREG-1150 Study 

indicate that the station blackout 

sequences are the largest contributors to 

internal event mean core damage 

frequency for Surry and are the dominant 

plant damage states leading to the first 

three release categories (see Figure 5). 

The radionuclide release characteristics 

into environment for different Surry 

release categories are presented in Table 

6. The magnitudes of releases for each 

release category were obtained by 

utilizing the basic parametric equation 

used in NUREG-1150. The mean values 

of the probability distributions of source 

term parameters associated with the 

corresponding accident progression 

characteristics were used in these 

calculations. The energy, timing, and 

duration of releases are based on the 

results of STCP calculations performed in 

support of NUREG-1150. These 

representative source terms have been 

used in a number of other studies 

including evaluation of risk importance of 

containment and related ESF system 

performance requirements [36] and the 

Regulatory Analysis Technical Evaluation 

Handbook (NUREG/BR-0184) [37]. 

Figure 11 presents a comparison of 

frequency distribution (CCDF) of iodine 

release predicted by the NUREG-1150 

Study for Surry with that obtained from 

the representative source terms used in 

NUREG/CR-6295. As shown in Figure 

11, these source terms represent well the 

spectrum of mean iodine releases 

predicted by the NUREG-1150 Study. 

Consequence calculations using 

representative source term are reported in 

References 33 and 36. A comparison of 

the mean frequency distribution of 

population dose (person-rem) to the entire 

region at Surry predicted by the NUREG1150 Study with those obtained by using 

representative source terms is presented 

in Figure 12. 

1.00E-08

1.00E-07

1.00E-06

1.00E-05

1.00E-04

1.00E-03

1.00E-05

1.00E-04

1.00E-03

1.00E-02

1.00E-01

1.00E+00

 Release Fraction

Exceedance Frequency (per reactor-year)

RSTs

NUREG-1150 (Surry), Mean

ECF

Bypass (V)

LCF

NCF

°

Figure 11: Comparison of frequency distribution (CCDF) of iodine release predicted by 

 NUREG-1150 Study, internal events, for Surry with that obtained from the 

 representative source terms (RSTs)
 26

1.00E-08

1.00E-07

1.00E-06

1.00E-05

1.00E-04

1.00E-03

1.00E+00

1.00E+01

1.00E+02

1.00E+03

1.00E+04

1.00E+05

1.00E+06

1.00E+07

1.00E+081.00E+09

 Poulation Dose (person-rem)

Exceedance Frequency (per reactor-year)

RSTs 

NUREG-1150 (Surry), Mean ECF Bypass (V) LCF

NCF

°

Figure 12: Comparison of frequency distribution (CCDF) of population dose (person- 

 rem) to entire region at Surry predicted by the NUREG-1150 Study, internal 

 events, with those obtained by using representative source terms (RSTs) 

It should be noted that a generic set of 

accident sequences and associated 

frequencies and source term parameters 

has also been recently developed by 

Electrical Power Research Institute 

(EPRI) and Nuclear Energy Institute (NEI) 

as a part of risk-informed evaluation of 

protective action strategies for nuclear 

plant off-site emergency planning [38]. 

However a review of these source terms 

is beyond the scope of the present study. 
 27

3. RECENT ADVANCES IN UNDERSTANDING OF 

SEVERE ACCIDENT PHENOMENOLOGY AND 

CONTAINMENT FAILURE MECHANISMS 

Since completion of the NUREG-1150 

Study, more analytical and experimental 

studies have been performed to address 

many severe accident issues, including 

direct containment heating, Mark I liner 

attack, and in-vessel steam explosion. 

Reference 39 provides an overview of the 

state of knowledge and uncertainties 

associated with severe accident 

phenomena and source tem issues. 

Some recent advances in understanding 

of the severe accident issues and 

containment failure mechanisms 

important to consequence assessment 

are provided below for use in 

demonstrating the feasibility of developing 

a simplified approach for updating the 

results of NUREG-1150 Study. 

3.1 Direct Containment 

Heating 

The concern for DCH arises only if vessel 

breach occurs while the reactor is at 

elevated pressure. For such cases, the 

expulsion of molten core debris could lead 

to very rapid and efficient heat transfer to 

the containment atmosphere, possibly 

accompanied by oxidation reactions and 

hydrogen combustion that further 

enhance the energy transfer. The 

pressurization accompanying this process 

is referred to as direct containment 

heating. 

The potential for DCH to cause 

containment failure depends on several 

factors, such as the primary system 

pressure, the size of opening in the 

vessel, the temperature and composition 

of the core debris exiting the vessel, the 

containment pressure and composition 

before the vessel breach, the amount of 

water in the cavity, and the dispersive 

characteristics of the reactor cavity. 

 

Since the completion of NUREG-1150, 

advances have been made in 

understanding of the DCH phenomena. 

The NRC staff identified DCH as a major 

issue for resolution in its revised severe 

accident research plan [40] and 

sponsored analytical and experimental 

programs for understanding the key 

physical processes in DCH. A number of 

experiments were performed in support of 

DCH issue resolution for PWRs. These 

experiments included both separate 

effects tests and integral effects tests, 

simulating the DCH processes in scaled 

models of the Zion, Surry, and Calvert 

Cliffs containments. The results of an 

assessment of the probability of 

containment failure due to DCH for the 

Zion nuclear power plant were published 

in NUREG/CR-6075 and its supplement 

[41]. By using the Risk Oriented Accident 

Analysis Methodology (ROOAM) [42], it 

was concluded that the containment 

failure probability due to DCH at Zion is 

so low as to be considered physically 

unreasonable. The basic understanding 

upon which the approach to quantification 

of DCH loads is based is that intermediate 

compartments trap most of the debris 

dispersed from the reactor cavity and that 

the thermal-chemical interactions during 

this dispersal process are limited by the 

incoherence in the steam blowdown and 

melt entrainment processes. With this 

understanding, it was possible to reduce 

most of the complexity of DCH 

phenomena to a single parameter: the 

ratio of the melt-entrainment time 

constant to the blowdown time constant, 

which is referred to as the coherence 

ratio. Reference 43 provides further
 28

discussions on application of ROAAM to 

address the DCH issue for 34 

Westinghouse plants with large dry or 

subatmospheric containment. As a part of 

a study to assess the risk importance of 

containment and related ESF system 

performance requirements, the accident 

progression event trees (APET) for Zion 

that had been used for NUREG-1150 was 

modified to reflect the latest 

understanding of DCH phenomena in 

Zion [36]. This also included incorporation 

of the containment fragility curve that was 

used in the NUREG-6075 study [41] and 

the Zion IPE [44]. The results of updated 

evaluation of the conditional probability of 

accident progression bins for internal 

events, as compared with the original 

results of NUREG-1150, are summarized 

in Table 7 (see also Figure 13). The 

conditional probability of early 

containment failure was found to be very 

low for the Zion plant. This is due to the 

fact that the Zion containment capacity 

was high and the expected containment 

loads from the core melt accidents were 

not high enough to threaten the integrity 

of the containment during the early stage 

of an accident. In addition, a large 

fraction of the plant damage states 

resulted in low RCS pressure at the time 

of vessel breach, which lowered the 

potential for loads associated with DCH. 

In view of more recent understanding of 

in-vessel steam explosion (alpha mode 

failure), the only major physical 

phenomena contributing to early 

containment failure in PWRs with large 

dry containments is containment isolation 

failure. 

 

Table 7: Impact of latest understanding of DCH on probability of accident progression 

bins for internal events at Zion (NUREG/CR-6418, pp. 2-9) [36] 
 29

Figure 13: Impact of latest understanding of DCH on probability of accident progression 

bins for internal events at Zion [39] 

 

3.2 Steam Explosion 

During progression of severe accidents, 

molten debris from the damaged core 

would at some point begin to fall into the 

lower plenum of the reactor vessel. If an 

appropriate amount of water remained in 

the lower plenum, molten core material 

falling into the lower plenum could 

potentially cause a steam spike and, if 

severe enough, an explosion. Rapid 

steam pressure rise and missile, resulting 

from in-vessel steam explosion (alpha 

mode failure) has been identified as a 

potential challenge to the containment in 

past studies [2,3]. However, a more 

recent assessment of this issue in 1996 

by an NRC sponsored steam explosion 

review group [45] concluded that alpha 

mode failure is of very low probability that 

is of little or no significance to the overall 

risk. 

3.3 Drywell Melt-through in 

Mark I Containments 

Drywell liner melt-through (caused by 

direct contact with core debris) has been 

found to be the most important contributor 

to early containment failure for Mark I 

containments. This failure mode is only 

possible for Mark I containments because 

the pedestal and drywell floor are at the 

same level, and core debris can easily 

reach the containment liner. The steel 

liner is the containment pressure 

boundary, and such a breach (i.e. drywell 

melt-through) would constitute an early 

containment failure. 

The issue of Mark I drywell shell (liner) 

melt-through at Peach Bottom was 

assessed by the NUREG-1150 molten 

core-containment interaction panel. The 

0.001

0.01

0.1

1

Isolation Failure

Early CF

Late CF

Bypass

No CF

Accident Progression Bin (APB)

Conditional Probability of APB A

Updated Evaluations NUREG-1150 Results
 30

Figure 14: Impact of latest understanding of drywell melt-through on probability of 

accident progression bins for internal events at Peach Bottom [39]

 

results of expert panel elicitation are 

reported in Reference 46. There were 

two schools of thought on this issue. 

Some experts felt that melt spreading is a 

hydrodynamically limited phenomenon 

and that there is a high probability of 

drywell failure, even with the presence of 

water. Other experts felt the movement of 

the debris is thermodynamically limited 

and will be impeded by crust formation 

and the presence of water. The experts 

provided subjective probability of drywell 

failure for several different scenarios, 

characterized by five parameters. These 

parameters were the RPV pressure, 

debris flow rate from the vessel, debris 

superheat, un-oxidized metal content of 

the debris, and the presence of water on 

the drywell floor. 

Since the completion of NUREG-1150, 

the NRC has sponsored analytical and 

experimental programs to address and 

resolve this so-called “Mark I Liner Attack” 

issue. The results of an assessment of 

the probability of Mark I containment 

failure by melt attack of the liner were 

published in NUREG/CR-5423 [47] and 

NUREG/CR-6025 [48]. It was concluded 

that, in the presence of water, the 

probability of early containment failure by 

melt-attack of the liner is so low as to be 

considered physically unreasonable. 

As part of a study to assess the risk 

significance of containment and related 

ESF system performance requirements 

[36], the Peach Bottom APET that had 

been used for the NUREG-1150 study 

was modified to reflect the more recent 

understanding of the drywell melt-through 

mechanism. The results of the updated 

evaluation of the conditional probability of 

accident progression bins for internal 

events, as compared with the original 

results of NUREG-1150, are summarized 

in Figure 14 (see also Table 8). The 

results of updated evaluations indicate a 

decrease in the probability of containment 

failure but not as much as one might 

expect. This is because of the possibility 

of multiple failure modes. Also, in the 

ATWS and SBO accident groups (the risk 

dominated plant damage states), there is 

a significant probability that the vessel 

failure will occur when there is no water in 

the pedestal. 

0.0001

0.001

0.01

0.1

1

Early WWF

Early DWF

Late WWF

Late DWF

Venting

No CF

Accident Progression Bin (APB)

Conditional Probability of APB 

Updated Evaluations NUREG-1150 Results
 31

Table 8: Impact of latest understanding of drywell melt-through on probability of accident 

progression bins for internal events at Peach Bottom (NUREG/CR-6418, pp. 2-15) [36] 
 32

4. A SIMPLIFIED APPROACH TO UPDATE THE 

RESULTS OF EARLIER LEVEL-3 PRAs FOR 

COMPARISON WITH SOARCA 

Although performing Level-3 PRAs for the 

pilot plants is the best way to benchmark 

the SOARCA methodology, results and 

insights from the NUREG-1150 Study and 

Integrated Risk Assessment for LaSalle, 

together with more recent advances in 

understanding of the severe accident 

issues and containment failure 

mechanisms, could be used for 

developing a simplified, yet systematic 

and defensible, approach to update the 

results of such earlier Level-3 PRAs for 

comparison with aspects of SOARCA 

results. The major elements of such an 

approach are depicted in Figure 15. 

Developing a small set of accident 

progression groups and their associated 

frequencies and release characteristics 

(source terms) to represent the full 

spectrum of potential severe accidents is 

an essential element of the proposed 

approach. The representative source 

terms developed previously [35, 33] and 

used in a number of other NRC studies 

[33, 36, 37] could be used as a starting 

point. As it was noted in Section 2.4, 

these representative source terms were 

developed based on the insights from the 

results of the NUREG-1150 Study and the 

Integrated Risk Assessment for LaSalle. 

Extending these representative source 

terms to include fire and seismic initiators 

(for Surry, Peach Bottom, and LaSalle 

Plants) as well as their further refinements 

to represent more accident progression 

groups and/or plant damage states may 

be necessary. Presenting these 

representative source terms and their 

associated frequencies in the form of 

complementary cumulative distribution 

functions (CCDFs) provides a logical 

framework for the subsequent 

adjustments and comparison with the 

SOARCA results. 

The representative accident groups and 

their associated frequencies can be 

revised based on the more recent insights 

on core damage frequencies (e.g., results 

of SPAR models) and containment failure 

modes and mechanisms. For example, 

the results of NUREG-1150 Study for 

conditional probability of accident 

progression bins for each “summary plant 

damage state” at Surry (see Figure 5) can 

be revised to reflect the current 

knowledge of containment failure modes 

and mechanisms The results of SPAR 

model for core damage frequencies 

obtained as a part of SOARCA, together 

with the results of conditional probabilities 

of accident progression bins can then be 

used to revise the representative accident 

groups and their associated frequencies 

for Surry. 

Table 9 presents an illustrative example 

of the impact of current knowledge and 

understanding of early containment failure 

on NUREG-1150 results for the 

conditional probability of accident 

progression bins at Surry. The results of 

NUREG-1150 Study indicate that, for the 

internal and fire initiators, the mean 

conditional probability of early 

containment failure is on the order of 0.01 

and the early containment failure is only 

due to in vessel steam explosion and 

DCH. It should be noted that the Surry 

containment is maintained at a 

subatmospheric pressure (10psia) during 

operation with a continued monitoring of 

the containment leakage, and thus the 

NUREG-1150 Study concluded that the 

likelihood of pre-existing leaks of 

significant size is negligible. Therefore, 

with the more recent understanding of 

DCH and in vessel steam explosion 

phenomena (as discussed in Chapter 3), 
 33

Figure 15: Elements of the Proposed Approach to Update the Results of Earlier Level-3 

PRAs for Comparison with SOARCA 

NUREG-1150 Study and the 

Integrated Risk Assessment 

for LaSalle 

Develop Spectrum of Accident 

Groups and their Associated 

Frequencies and Radionuclide 

Release Characteristics 

Revise Accident Groups and their 

Release Frequencies 

Recent Insights into Core 

Damage Frequencies (e.g., 

SPAR models) and 

Containment Failure Modes 

and Mechanism 

Revise Release Characteristics 

(magnitude and timing) 

Recent Insights on Severe 

Accident Progression and 

Source Term Issues (e.g., 

new MELCOR Code 

calculations) 

Assessment of Mitigating 

Measures (e.g., SAMGs, 

EDMGs ) 

Further Revise Release 

Frequencies and/or Release 

Characteristics

Compare Results with 

Those obtained by 

SOARCA 

MACCS2 Analysis 

Compare Results of 

Consequence Measures with 

Those Obtained by SOARCA 
 34

Table 9: Revised mean conditional probability of accident progression bins at Surry 

Summary PDS Group 

(Mean Core Damage Frequency) 

Internal initiators 

(4.1E-05)

Summary 

Accident 

progression 

Bin Group LOSP 

(2.8E-05) 

ATWS 

(1.4E-06) 

Transients

(1.8E-06) 

LOCAs 

(6.1E-06) 

ISLOCA 

(1.6E-06)

SGTR 

(1.8E-06)

Fire 

(1.1E-05) 

Sei

LL

(1.9

Early CF -- (0.008)(a) -- (0.003) -- (0.001) -- (0.006) -- (0.018) 0.0

(0.0

Late CF 0.084 (0.079) 0.046 (0.046) 0.014 (0.013) 0.056 (0.055) 0.305 (0.292) 0.2

(0.2

Bypass (0.003) (0.078) (0.007) (1.0) (1.0) (0.0

No CF 0.913 (0.909) 0.876 (0.873) 0.979 (0.979) 0.944 (0.939) 0.695 (0.690) 0.6

(0.6

 (a) Numbers in parentheses are the results of the NUREG-1150 Study. 
 35

it can be concluded that, for the internal 

and fire initiators, the early containment 

failure for Surry is of very low conditional 

probability that it is of little or no 

significance to the overall risk. However, 

this conclusion does not hold for the 

seismic initiators. The results of the 

NUREG-1150 Study indicate that, for the 

seismic initiators, the mean conditional 

probability of early CF is on the order 0.1 

and most of these early failures of the 

containment are initial failures due to 

steam generator (SG) and reactor coolant 

pump (RCP) support failures. The staff 

has yet to report whether it has addressed 

this separate failure mechanisms in 

SOARCA. Revisions to frequencies of 

plant damage states together with the 

conditional probability of accident 

progression bins for each “summary plant 

damage state” provide a framework for 

systematic evaluation of selection of the 

accident groups and their associated 

containment failure modes and release 

frequencies in SOARCA. 

 As an illustrative example, for the 

purpose of comparing the results of 

radionuclide releases in SOARCA for 

internal initiators at Surry with those of the 

updated NUREG-1150 Study, the 

representative source terms for Surry, 

discussed in Section 2.4 (see Table 6), 

can be refined by defining an additional 

source term representing steam generator 

tube rupture (SGTR) plant damage state 

(PDS GROUP 7 of NUREG-1150 Study 

for Surry, Internal Initiators). For this 

additional bypass release category, based 

on the results of NUREG-1150 for the 

most probable bin, it is assumed the 

safety relief valves (SRVs) on the 

secondary system are stuck open, the 

containment spray system never operates 

during the accident, the core-concrete 

interaction (CCI) takes place promptly 

following vessel breach (VB), and there is 

no overlaying water pool to scrub the 

releases due to CCI. The magnitude of 

radionuclide releases into the 

environment for this release category can 

be obtained by the basic parametric 

equation used in the NUREG-1150 Study. 

The mean values of the probability 

distributions of the source term 

parameters associated with the 

corresponding accident progression 

characteristics can be used in these 

calculations. Table 10 provides a 

summary of the frequencies and 

magnitudes of Iodine releases into the 

environment for representative source 

terms for Surry. The mean fractional 

iodine release of 0.20 for the SGTR 

representative source term is based on 

the mean value reported for the most 

probable bin within that plant damage 

state group (NUREG/CR-4551, Vol. 1, 

Table 3.3-7) [49]. A comparison of 

frequency distribution (CCDF) of iodine 

release predicted by the NUREG-1150 

Study for Surry with that obtained from 

these representative source terms is 

shown in Figure 16. 

Based on discussions provided earlier, for 

internal initiators, the early containment 

failure for Surry is of very low probability 

that it is of little or no significance to the 

overall risk. Therefore, a representative 

source term associated with the early 

containment failure can be discarded. It 

should be noted that each representative 

source term represents many accident 

progression bins with different 

frequencies and release magnitudes 

within the accident progression group it 

represents. There is an overlap of release 

magnitudes between the accident 

progression groups. Therefore, by 

eliminating the representative source term 

associated with early containment failure, 

care must be taken to represent other low 

probability and high release bins. The 

results of the NUREG-1150 Study for 

Surry (NUREG/CR-4551, Vol. 1 Table 

2.5.4) [49] indicate that for the bypass 

accident progression group due to 

interfacing-system LOCA (Event V), there 

is a conditional probability of ~0.15 that 

there is no water in the path of the 

radionuclide releases before entering into 
 36

Table 10: Frequencies and Magnitudes of Iodine Releases for Representative Source Terms for Surry (In

Frequency 

Release 

Category 

Summary PDS 

Group 

Containment 

Failure Time 

Containment 

Failure 

Mode 

Based on 

NUREG-1150 

Study 

Revised 

Based on 

Results of 

SPAR Model

and no Early 

Failure of 

Cont. 

RSUR1 LOSP CF at VB 

(ECF) 

Rupture 2.9E-07 ---- 

RSUR2 LOSP Late CF 

(LCF) 

Leak 2.4E-06 1.5E-07 

RSUR3 LOSP No CF 

(NCF) 

No CF 3.3E-05 1.95E-06 

RSUR4 Bypass (V) NCF Bypass 1.6E-06 

Wet (~85%) 

Dry (~15%) 

3.5E-07 

Wet (~3.0E-07) 

Dry (~5.0E-08) 

RSUR5 Bypass (SGTRs) NCF Bypass 1.8E-06 5.5E-07 
 37

 

ECF

LCF

Bypass (SGTR)

RSTs based on NUREG-1150 Study

V (dry)

NCF

Bypass (V)

NUREG-1150

■ RSTs based on results of SPAR Model and no ECF

◊

V (wet)

SOARCA (V)

1.00E-08

1.00E-07

1.00E-06

1.00E-05

1.00E-04

1.00E-03

1.00E-051.00E-04

1.00E-03

1.00E-02

1.00E-01

1.00E+00

 Release Fraction

Exceedance Frequency

ECF

LCF

Bypass (SGTR)

RSTs based on NUREG-1150 Study

V (dry)

NCF

Bypass (V)

NUREG-1150

■ RSTs based on results of SPAR Model and no ECF

◊

V (wet)

the environment (dry releases). In the 

absence of the decontamination effect of 

water in the release path, the release 

magnitude would be higher. An additional 

source term representing this bypass subgroup can be defined. The magnitude of 

radionuclide releases into the 

environment for this release sub-category 

can be obtained by the basic parametric 

equation used in the NUREG-1150 Study. 

The results of the revised frequencies of 

the representative source terms for Surry 

is also presented in Table 10. The mean 

fractional iodine release of 0.37 shown in 

Table 10 for the V bypass accident 

progression sub-group (Dry releases) is 

based on the mean value reported for the 

most probable bin within that sub-group 

(NUREG/CR-4551, Vol. 1, Table 3.3-4). 

The frequency distribution (CCDF) of 

iodine release for Surry obtained from 

revision to representative source terms, 

based on the results of SPAR model and 

recent insights on early containment 

failure mechanisms, is also shown in 

Figure 16. As shown in Figure 16 (see 

also Table 10), the revised frequency of 

the V bypass accident progression subgroup (dry releases) is about 5.0 x 10-8. 

The staff has yet to report whether it has 

included the releases associated with 

such a bypass sequence for SOARCA. It 

is not clear how such an accident 

progression group can be excluded from 

SOARCA based on the screening 

frequency (10-7 for bypass sequences) 

argument, without the benefit of such 

information obtained from the Level-2 

PRAs. 

 

The release characteristics (magnitude 

and timing) of the representative source 

terms can also be revised based on the 

more recent insights on severe accident 

progression and source term issues (e.g., 

new MELCOR code calculations). The 

Figure 16: Comparison of frequency distribution (CCDF) of iodine 

release predicted by NUREG-1150 Study for Surry with that obtained 

from the representative source terms (RSTs) and their revised 

frequencies based on results of SPAR model and recent insights on 

early containment failure mechanisms 
 38

staff has not yet provided any document 

on accident progression and source term 

analysis using severe accident code 

MELCOR. It is not clear how the state-ofknowledge uncertainties associated with 

the accident progression phenomena and 

source term analyses are addressed. 

The revised representative source terms 

and their associated frequencies in the 

form of CCDFs provide a possible 

framework for evaluating the selection of 

the accident sequences and the 

magnitude of the radionuclide releases 

into the environment obtained as a part of 

the SOARCA process. These 

radionuclide release characteristics are 

functions of the design and operating 

procedures of the plant and, unlike the 

results of consequence measures, they 

are independent of the population 

distribution, meteorology, and the 

emergency protective assumptions at the 

reactor site. 

The frequencies and/or release 

characteristics of the representative 

source terms may be further revised 

based on the results of the assessment of 

the effectiveness of mitigating measures. 

Such mitigating measures are based on 

Severe Accident Management Guidelines 

(SAMGs) and other new procedures, such 

as mitigating measures resulting from 

B.5.b and other like programs that were 

not in place when the earlier 

consequence studies, including the 

NUREG-1150 Study, were performed. 

Development of a simplified accident 

progression event tree which only 

considers significant high level issues, 

based on the insights from the NUREG1150 Study and more recent advances in 

understanding of the severe accident 

issues, containment failure mechanisms, 

and effectiveness of potential mitigating 

measures would facilitate the revision to 

release frequencies and providing a 

systematic approach to justify the 

selection of sequences for SOARCA. 

Such simplified accident progression 

event tree would be particularly useful for 

assessment of the effectiveness of 

mitigating measures resulting from 

SAMGs and other new programs (e.g. 

B.5.b). 

Using MACCS2 code, Consequence 

analyses could be performed to determine 

consequence measures (e.g., population 

dose, early fatalities, and latent cancer 

fatalities) for each representative source 

terms. The frequency distributions of 

consequence measures in the form of 

CCDFs provide a useful framework for 

comparison with the results of SOARCA, 

as well as for their comparison with 

historical results. 
 39

5. SUMMARY AND CONCLUSIONS 

The probability and offsite consequences 

of severe reactor accidents have been the 

subject of considerable interest and study 

since the earliest days of reactor 

development. Several systematic studies 

have been made in the past to search out 

a large spectrum of accidents and to use 

quantitative techniques to estimate the 

probabilities, radionuclide release 

characteristics (source terms), and 

potential offsite health consequences. An 

overview of major contributions to 

consequence assessment was presented 

to provide a historical perspective and 

insights on previous state-of-the-art 

analyses of the consequences of severe 

reactor accidents. 

The staff is currently implementing its plan 

for developing state-of-the-art reactor 

consequence analyses, using a “novel 

analysis procedure that differs 

substantially from previous state-of-the-art 

analyses of the consequences of severe 

reactor accidents” [13]. Although 

performing Level-3 PRAs for the pilot 

plants is the best way to benchmark 

SOARCA methodology, it is feasible to 

use the results and insights from the 

NUREG-1150 Study and Integrated Risk 

Assessment for LaSalle, together with 

more recent advances in understanding of 

the severe accident issues and 

containment failure mechanisms, and 

develop a simplified, yet systematic and 

defensible, approach to update the results 

of such earlier Level-3 PRAs for 

comparison with aspects of SOARCA 

results. 
 40

6. REFERENCES 

 

1. U.S. Atomic Energy Commission, 

 “Theoretical Possibilities and 

 Consequences of Major Accidents in 

 Large Nuclear Power Plants,” 

 Brookhaven National Laboratory, 

 WASH-740, March 1957. 

2. U.S. NRC, “Reactor safety Study - 

 An Assessment of Accident 

 Risks in U.S. Commercial Nuclear 

 Power Plants,” WASH-1400 

 (NUREG- 75/014), October 1975. 

3. U.S. NRC, “Severe Accident 

 Risks: An Assessment for Five U.S. 

 Nuclear Power Plants,” Final 

 Summary Report, NUREG-

 1150, December 1990. 

4. Commonwealth Edison Company, 

 "Zion Probabilistic Safety Study," 

 1981. 

5. Power Authority of the State of New 

 York and Consolidated Edison 

 Company of New York, Inc., "Indian 

 Point Probabilistic Safety Study," 

 1982. 

6. Northeast Utilities, “Millstone Unit 3 

 Probabilistic Safety Study,” 1983. 

7. Pickard, Low and Garrick, Inc., 

 “Seabrook Station Probabilistic 

 Safety Assessment,” 1983. 

8. Aldrich, D.C., J. L. Sprung, D. J. 

 Alpert, K. Diegert, R.M. Ostmeyer, 

 L. T. Ritchie, D.R. Strip, J. D. 

 Johnson, K. Hanson, and J. 

 Robinson, “Technical Guidance for 

 Siting Criteria Development,” 

 NUREG/CR-2239, Sandia National 

 Laboratories, December 1982. 

9. U.S. NRC, “Plan for Developing 

 State-of-the-Art Reactor 

 Consequence Analyses,” SECY-05-

 0233, December 22, 2005. (Official 

 Use Only - Sensitive Internal 

 Information, Limited to NRC Unless 

 the Commission Determines 

 Otherwise) 

10. Memorandum to Luis A. Reyes, 

Executive Director for Operations, 

NRC, from Kenneth R. Hart, Acting 

Secretary, NRC; Subject: Staff 

Requirements – SECY-05-0233-Plan 

for Developing State-of-the-Art 

Reactor Consequence Analyses, 

dated April, 14, 2006. (Official Use 

Only - Sensitive Internal Information, 

Limited to NRC Unless the 

Commission Determines Otherwise) 

11. Memorandum to Luis A. Reyes, 

Executive Director for Operations, 

NRC, and Eliot B. Brenner, Director, 

Office of Public Affairs, NRC, from 

Kenneth R. Hart, Acting Secretary, 

NRC; Subject: Staff RequirementsCOMSECY-06-0064 – State-of-theArt Reactor Consequence Analyses 

Communications Plan, dated April 2, 

2007. (Official Use Only - Sensitive 

Internal Information, Limited to NRC 

Unless the Commission Determines 

Otherwise) 

12. Report dated February 25, 2008, from 

William J. Shack, Chairman, ACRS, to 

Dale E. Klein, Chairman, NRC, 

Subject: State-of-the-Art Reactor 

Consequence Analyses Project. 

13. Letter dated April 7, 2008, from Luis 

A. Reyes, Executive Director for 

Operations, NRC, to William J. Shack, 

Chairman, ACRS, Subject: State-ofthe-Art Reactor Consequence 

Analyses Project. 
 41

14. Letter dated April 21, 2008, from 

William J. Shack, Chairman, ACRS, to 

Luis A. Reyes, Executive Director for 

Operations, NRC, Subject: Response 

to Your April 7, 2008 Letter Regarding 

State-of-the-Art Reactor 

Consequence Analyses (SOARCA) 

Project. 

15. Memorandum to Frank P. Gillespie, 

Executive Director, ACRS, and R. W. 

Borchardt, Executive Director for 

Operations, NRC, from Annette ViettiCook, Secretary, NRC; Subject: Staff 

Requirements - Meeting with Advisory 

Committee on Reactor Safeguards 

(ACRS), Thursday, June 5, 2008, 

Commissioners' Conference Room, 

One White Flint North, Rockville, 

Maryland, dated June 26, 2008. 

16. Keller W. and M. Modarres, “A 

Historical Overview of Probabilistic 

Risk Assessment Development And 

Its Use In the Nuclear Power Industry: 

A Tribute to the Late Professor 

Norman Carl Rasmussen,” Reliability 

Engineering & System Safety, Vol. 89, 

Issue 3, pp. 271-285, September 

2005, 

17. Lewis H, et al., “Risk Assessment 

Review Group Report to the US 

Nuclear Regulatory Commission,” 

NUREG/CR-0400, 1978. 

18. U.S. NRC, “Technical Bases for 

Estimating Fission Product Behavior 

During LWR Accidents,” NUREG0772, 1981. 

19. Letter from Nuclear Safety Oversight 

Committee to President Jimmy Carter, 

dated December 21, 1980. 

20. U.S. NRC, “Regulatory Impact of 

Nuclear Reactor Accident Source 

Term Assumption,” NUREG-0771, 

Draft for Comment, June1981. 

21. Strip D. R., “Estimates of the Financial 

Consequences of Nuclear Power 

Reactor Accidents,” NUREG/CR2723, Sandia National Laboratories, 

September 1982. 

22. Gieseke J. A., et al., “Radionuclide 

Release under Specific LWR Accident 

Conditions,” Battelle Memorial 

Institute Report, BMI-2104, February 

1985. 

23. U.S. NRC, “Reassessment of the 

Technical Bases for Estimating 

Source Terms,” NUREG-0956, July 

1986. 

24. U.S. NRC, “Source Term Code 

Package: A Users’ Guide,” 

NUREG/CR-4587, BMI-2138, July 

1986. 

25. U.S. NRC, “MELCOR 1.8.0: A 

Computer Code for Nuclear Reactor 

Severe Accident Source Term and 

Risk Assessment Analyses,” 

NUREG/CR-5531, SAND90-0364, 

January 1991. 

26. U.S. NRC, “Individual Plant 

Examination Program: Perspectives 

on Reactor Safety and Plant 

Performance,” NUREG-1560, 

November 1996. 

27. Griesmeyer, J. M. and L. N. Smith, “A 

Reference Manual for the Event 

Progression Analysis Code 

(EVNTRE),” NUREG/CR-5174, 

SAND88-1607, September 1989. 

28. Chanin, D. I., J Rollstin, J. Foster, and 

L. Miller,” MACCS Version 1.5.11.1: A 

Maintenance Release of the Code,” 

NUREG/CR-6059, Sandia National 

Laboratories, October 1993. 
 42

29. Presidential Commission on 

 Catastrophic Nuclear Accidents, 

 “Report to the Congress from 

 Presidential Commission on 

 Catastrophic Nuclear Accidents,” 

 Washington, DC, August 1990, 

 (http://www.state.nv.us/nucwaste/ne

 ws/rpccna/pcrcna01.htm) 

30. Soffer, L., et al., “Accident Source 

 Terms for Light-Water Nuclear 

 Power Plants,” NUREG-1465, 

 February 1995. 

31. Nourbakhsh, H. P., “Estimate of 

 Radionulide Release Characteristics 

 into Containment under Severe 

 Accident Conditions.” NUREG/CR-

 5747, BNL-NUREG-52289, 

 Brookhaven National Laboratory, 

 November 1993. 

32. U.S. NRC, “Review of NUREG-

 0654, Supplement 3, ‘Criteria for 

 Protective Action Recommendations 

 for Severe Accidents,’” Draft 

 NUREG/CR-XXXX, Volume 1, 

 SAND2007-XXXX, 2007. 

33. Davis, R.E., A.L. Hanson, V. 

 Mubayi, and H.P. Nourbakhsh, 

 “Reassessment of Selected 

 Factors Affecting Siting of Nuclear 

 Power Plants,” Brookhaven National 

 Laboratory, NUREG/CR-6295, 

 February 1997. 

34. Brown, T.D., et al., “Integrated Risk 

 Assessment for the LaSalle Unit 2 

 Nuclear Power Plant,” Sandia 

 National Laboratories, NUREG/CR-

 5305, Vol. 1, August 1992. 

35. Nourbakhsh, H. P., “Representative 

 Source Terms for Severe LWR 

 Accidents,” Brookhaven 

 National Laboratory, Technical 

 report, 1992. 

36. Nourbakhsh, H. P., A. L. Hanson, 

 and W. T. Pratt, “Risk Importance of 

 Containment and Related ESF 

 System Performance 

 Requirements,” Brookhaven 

 National Laboratory, NUREG/CR-

 6418, BNL-NUREG-52489, 

 November 1998. 

37. U.S. NRC, “Regulatory Analysis 

 Technical Evaluation Handbook,” 

 NUREG/BR-0184, January 1997. 

38. Leaver, D., ”Risk-Informed 

 Evaluation of Protective Action 

 Strategies for Nuclear Plant Off-

 Site Emergency Planning,” Polestar 

 Applied Technology, Inc., EPRI, NEI, 

 September 2007. 

39. Nourbakhsh, H.P., and T.S. Kress, 

 “Assessment of Phenomenological 

 Uncertainties in Level 2 PRAs,” 

Proceeding of workshop on 

 Evaluation of Uncertainties in 

 Relation to Severe Accidents and 

 Level-2 Probabilistic Safety Analysis, 

Aix-en-Provence, 7-9 November 

 2005, NEA/CSNI/R(2007), 2007. 

40. U.S. Nuclear Regulatory 

 Commission, “Severe Accident 

 Research Program Plan Update,” 

 NUREG-1365, Rev. 1, December 

 1992. 

41. Pilch, M. M., H. Yan, and T.G. 

 Theofanous, “The Probability of 

 Containment Failure by Direct 

 Containment Heating in Zion,” 

 NUREG/CR-6075, SAND93-1535, 

 December 1994. 

42. Theofanous, T. G., “Dealing with 

 Phenomenological uncertainty in 

 Risk Analysis,” Proceeding of 

 Workshop in Advanced Topics in 

 Risk and Reliability Analysis – 

 Model Uncertainty: Its 

 Characterization and Quantification, 
 43

 Annapolis, Maryland October 20-22, 

 1993, NUREG/CP-0138, 1994. 

43. Pilch, M. M., et al., “Resolution of 

 Direct Containment Heating Issue 

 for all Westinghouse Plants with 

 Large Dry Containment or 

 Subatmospheric Containment,” 

 NUREG/CR-6338, February 

 1996. 

44. Zion IPE, “Commonwealth-Edison 

 Zion Station: Individual Plant 

 Evaluation”, NRC Doc. 9204290315, 

 April 1992. 

45. Basu, S. and T. Ginsberg, “A 

 Reassessment of the Potential for 

 an Alpha-Mode Containment 

 Failure and a Review of Broader Fuel-

 Coolant Interaction Issues,” 

 NUREG-1524, August 1996. 

46. Harper, F. T., et al., “Evaluation of 

 Severe Accident Risks: 

 Quantification of Major Input 

 Parameters, Experts’ Determination 

 of Containment Loads and Molten 

 Core Containment Interaction 

 ssues,” Sandia National 

 Laboratories, NUREG-CR-4551, 

 SAND86-1309, Vol. 2, Rev. 1, Part 

 2, April 1991. 

47. Theofanous, T.G., et al., “The 

 Probability of Liner Failure in a Mark 

 I Containment,” NUREG/CR-5423, 

 1989. 

48. Theofanous, T.G., et al., “The 

 Probability of Mark I Failure by Melt 

 –Attack of the Liner,” NUREG/CR-

 6025, November 1993. 

49. Breeding, R. J., J. C. Helton, W. B. 

 Muffin, and L. N. Smith, “Evaluation 

 of Severe Accident Risks: Surry Unit 

 1, Main Report,” NUREG/CR-4551 

 SAND86-1309 Vol. 3, Rev. 1, 

 Part 1, October 1990.