State-of-the-Art Reactor 

Consequence Analyses 

(SOARCA) Report

Office of Nuclear Regulatory Research

NUREG-1935
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owned rights.
State-of-the-Art Reactor 

Consequence Analyses 

(SOARCA) Report

Manuscript Completed: June 2012

Date Published: November 2012

Prepared by:

Richard Chang, Jason Schaperow, Tina Ghosh, 

Jonathan Barr, Charles Tinkler, Martin Stutzke

Office of Nuclear Regulatory Research

NUREG-1935

iii

ABSTRACT

Accident phenomena and offsite consequences of severe reactor accidents have been the subjects

of considerable research over the last several decades by the U.S. Nuclear Regulatory 

Commission (NRC). As a consequence of this research focus, analyses of severe accidents at 

nuclear power reactors are more detailed, integrated, and realistic than at any time in the past. A 

desire to leverage this capability to address conservative aspects of previous reactor accident 

analyses was a major motivating factor in the genesis of the State-of-the-Art Reactor 

Consequence Analyses (SOARCA) project. By applying modern analysis tools and techniques, 

the SOARCA project developed a body of knowledge regarding the realistic outcomes of select 

severe nuclear reactor accidents. To accomplish this objective, the SOARCA project’s 

integrated modeling of accident progression and offsite consequences used both state-of-the-art 

computational analysis tools and best modeling practices drawn from the collective wisdom of 

the severe accident analysis community. This study has focused on providing a realistic 

evaluation of accident progression, source term, and offsite consequences for select scenarios for 

the Peach Bottom Atomic Power Station and Surry Power Station. By using the most current 

emergency preparedness practices and plant capabilities, as well as the best available modeling, 

these analyses are more realistic than past analyses. These analyses also consider mitigative 

measures (e.g., emergency operating procedures, severe accident management guidelines, and 

Title 10 to the Code of Federal Regulations (10 CFR) 50. 54(hh) measures), contributing to a 

more realistic evaluation.

Paperwork Reduction Act Statement

The NUREG does not contain information collection requirements and, therefore, is not subject

to the requirements of the Paperwork Reduction Act of 1995 (44 USC 3501, et seq.).

Public Protection Notification

The NRC may not conduct or sponsor, and a person is not required to respond to, a request for 

information or an information collection requirement unless the requesting document displays a 

current valid OMB control number.

v

CONTENTS

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

FIGURES....................................................................................................................................viii

TABLES........................................................................................................................................ ix

EXECUTIVE SUMMARY ......................................................................................................... xi

ACKNOWLEDGEMENTS ...................................................................................................... xxi

ACRONYMS and ABBREVIATIONS..................................................................................xxiii

1.0 INTRODUCTION................................................................................................................. 1

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

1.2 Objective ................................................................................................................................. 2

1.3 Approach ................................................................................................................................. 3

1.4 Historical Perspectives............................................................................................................ 4

1.4.1 WASH-1400 (NUREG-75/014), “Reactor Safety Study: An Assessment 

of Accident Risks in U.S. Commercial Nuclear Power Plants,” 1975 ......................... 4

1.4.2 NUREG/CR-2239, “Technical Guidance for Siting Criteria Development,” 1982 ..... 6

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

Power Plants,” 1990...................................................................................................... 6

1.5 Scope ....................................................................................................................................... 7

1.6 Basis of Accident Selection..................................................................................................... 9

1.7 Mitigated and Unmitigated Cases ......................................................................................... 10

1.8 Uncertainty Analysis............................................................................................................. 12

1.9 Structure of NUREG-1935 and Supporting Documents....................................................... 12

2.0 ACCIDENT SCENARIO SELECTION............................................................................ 15

2.1 Approach ............................................................................................................................... 15

2.2 Scenarios Initiated by Internal Events................................................................................... 18

2.3 Scenarios Initiated by External Events.................................................................................. 19

2.4 Accident Scenarios Selected for Surry.................................................................................. 20

2.4.1 Surry Internal Event Scenarios ................................................................................... 20

2.4.2 Surry External Event Scenarios.................................................................................. 22

2.5 Accident Scenarios Selected for Peach Bottom .................................................................... 23

2.5.1 Peach Bottom Internal Event Scenarios...................................................................... 23

2.5.2 Peach Bottom External Event Scenarios..................................................................... 23
vi

2.6 Generic Factors...................................................................................................................... 24

3.0 MITIGATIVE MEASURES ASSESSMENT ................................................................... 27

3.1 Site-Specific Mitigation Strategies........................................................................................ 27

3.1.1 Scenarios Initiated by External Events....................................................................... 28

3.1.2 Scenarios Initiated by Internal Events ........................................................................ 31

3.2 Truncation of Releases.......................................................................................................... 32

4.0 SOURCE TERM ANALYSIS ............................................................................................ 33

4.1 Source Term Study Background ........................................................................................... 33

4.2 The MELCOR Code.............................................................................................................. 35

4.3 MELCOR Modeling Approach ............................................................................................. 37

4.3.1 Plant Models............................................................................................................... 38

4.3.2 Best Modeling Practices ............................................................................................. 45

4.3.3 Radionuclide Modeling............................................................................................... 46

4.3.4 Radionuclide Inventory............................................................................................... 47

5.0 OFFSITE CONSEQUENCE ANALYSES........................................................................ 51

5.1 Weather Sampling ................................................................................................................. 52

5.2 Weather Data......................................................................................................................... 52

5.2.1 Summary of Weather Data.......................................................................................... 53

5.3 Emergency Response Modeling............................................................................................ 55

5.3.1 Base Case Analyses of Emergency Response ............................................................ 57

5.3.2 Sensitivity Analyses of Emergency Response............................................................ 59

5.4 Source Term Evaluation........................................................................................................ 60

5.5 Site-Specific Parameters........................................................................................................ 61

5.6 Non-Site-Specific Parameters ............................................................................................... 63

5.7 Estimating Latent Cancer Fatality Health Effects................................................................. 65

5.8 Risk Metrics Reported........................................................................................................... 68

6.0 RESULTS AND CONCLUSIONS..................................................................................... 69

6.1 Mitigation .............................................................................................................................. 69

6.2 Accident Progression and Radionuclide Release .................................................................. 69

6.3 Offsite Radiological Consequences....................................................................................... 74

6.4 Comparison to NUREG/CR-2239 (the 1982 Siting Study SST1 case)................................. 79

6.5 Sensitivity Analyses on the Size of the Evacuation Zone ..................................................... 82

6.6 Conclusions........................................................................................................................... 84
vii

7.0 REFERENCES .................................................................................................................... 87

APPENDIX A: SOARCA AND THE FUKUSHIMA DAIICHI ACCIDENT...................A-1

APPENDIX B: SOARCA PEER REVIEW COMMITTEE FINAL LETTERS

 AND SOARCA PEER REVIEW COMMENT RESOLUTION

 REPORT...............................................................................B-1

APPENDIX C: SOARCA PUBLIC COMMENTS SUMMARY...............................C-1
viii

FIGURES

Figure 1 The State-of-the-Art Reactor Consequence Analyses process ..................................... 3

Figure 2 SOARCA accident scenario selection and analysis process....................................... 16

Figure 3 Timeline of key nuclear power events and safety studies........................................... 33

Figure 4 MELCOR integration of separate effects codes ......................................................... 36

Figure 5 The Peach Bottom MELCOR nodalization ................................................................ 41

Figure 6 The Surry MELCOR nodalization.............................................................................. 44

Figure 7 Schematic of modeling detail for BWR GNF 10×10 assembly.................................. 49

Figure 8 Peach Bottom—Year 2006—wind-rose and atmospheric stability chart................... 54

Figure 9 Surry—Year 2004—wind-rose and atmospheric stability chart................................. 55

Figure 10 10- and 20-mile radial distances around the Peach Bottom site ................................. 56

Figure 11 Typical evacuation routing ......................................................................................... 58

Figure 12 Iodine releases to the environment for SOARCA unmitigated scenarios 

and the 1982 Siting Study SST1 case ......................................................................... 73

Figure 13 Cesium releases to the environment for SOARCA unmitigated scenarios 

and the 1982 Siting Study SST1 case ......................................................................... 73

Figure 14 Comparison of average individual LCF risk results for SOARCA mitigated and 

unmitigated scenarios to the NRC Safety Goal and to extrapolations of the 

1982 Siting Study SST1.............................................................................................. 81

Figure 15 Mean, individual, LNT, scenario-specific LCF risk for the Peach Bottom 

unmitigated STSBO scenario without RCIC blackstart for people within 

a circular area of specified radius from the plant........................................................ 83

Figure 16 Mean, individual, LNT, scenario-specific LCF risk for the Surry 

unmitigated ISLOCA scenario for people within a circular area 

of specified radius from the plant. .............................................................................. 83
ix

TABLES

Table 1 Summary of Core Damage Frequency from NUREG-1150 .......................................... 7

Table 2 Statistical Summary of Raw Meteorological Data for SOARCA Nuclear Sites.......... 54

Table 3 Deposition Velocities Used in the SOARCA Analyses............................................... 61

Table 4 Peach Bottom Accident Progression Timing Results................................................... 71

Table 5 Surry Accident Progression Timing Results................................................................ 72

Table 6 Peach Bottom Results for Scenarios Assuming LNT Dose-Response Model ............. 76

Table 7 Surry Results for Scenarios Assuming LNT Dose-Response Model........................... 77

Table 8 Peach Bottom Results for Scenarios without Successful Mitigation for LNT and 

Alternative Dose-Response models.............................................................................. 78

Table 9 Surry Results for Scenarios without Successful Mitigation for LNT and 

Alternative Dose-Response Models............................................................................. 79

Table 10 Conditional (i.e., assuming accident occurs), Mean, LNT, Scenario-Specific 

Probabilities of LCF for People within the Specified Radii of the Peach 

Bottom Site................................................................................................................... 80

Table 11 Conditional, Mean, LNT, Scenario-Specific Probabilities of LCF for People 

within the Specified Radii of the Surry Site................................................................. 80

Table 12 Effect of Size of Evacuation Zone on Mean, Individual, LNT, Scenario-Specific 

LCF Risk for People within the Specified Radii of the Plant for the 

Peach Bottom Unmitigated STSBO without RCIC Blackstart and the 

Surry Unmitigated ISLOCA......................................................................................... 82

xi

EXECUTIVE SUMMARY 

The U.S. Nuclear Regulatory Commission (NRC), the nuclear power industry, and the 

international nuclear energy research community have devoted considerable research over the 

last several decades to examining severe reactor accident phenomena and offsite consequences. 

Following the terrorist attacks of 2001, an NRC initiative reassessed severe accident progression 

and offsite consequences in response to security-related events. These updated analyses 

incorporated the wealth of accumulated research and used more detailed, integrated, and bestestimate modeling than past analyses. An insight gained from these security assessments was 

that the NRC needed updated analyses of severe reactor accidents to reflect realistic estimates of 

the more likely outcomes, considering the current state of plant design and operation and the 

advances in understanding of severe accident behavior.

The NRC initiated the State-of-the-Art Reactor Consequence Analyses (SOARCA) project to 

develop best estimates of the offsite radiological health consequences for potential severe reactor 

accidents for two pilot plants: the Peach Bottom Atomic Power Station in Pennsylvania and the 

Surry Power Station in Virginia. Peach Bottom is generally representative of U.S. operating 

reactors using the General Electric boiling-water reactor (BWR) design with a Mark I 

containment. Surry is generally representative of U.S. operating reactors using the 

Westinghouse pressurized-water reactor (PWR) design with a large, dry (subatmospheric) 

containment. SOARCA 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.

The SOARCA project evaluates plant improvements and changes not reflected in earlier NRC 

publications such as NUREG/CR-2239, “Technical Guidance for Siting Criteria 

Development,”[1] NUREG-1150, “Severe Accident Risks: An Assessment for Five U.S. 

Nuclear Power Plants,” [2] and WASH-1400, “Reactor Safety Study: An Assessment of 

Accident Risks in U.S. Commercial Nuclear Power Plants,” [3]. SOARCA includes system 

improvements, improvements in training and emergency procedures, offsite emergency response, 

and security-related improvements, as well as plant changes such as power uprates and higher 

core burnup. To provide perspective between SOARCA results and more conservative offsite 

consequence estimates, SOARCA results are compared to NUREG/CR-2239, “Technical 

Guidance for Siting Criteria Development,” issued in 1982 and referred to in this report as the 

Siting Study [1]. Specifically, SOARCA results are compared to the Siting Study siting source 

term 1 (SST1). SST1 assumes severe core damage, loss of all safety systems, and loss of 

containment after 1.5 hours. The SOARCA report helps the NRC to communicate its current 

understanding of severe-accident-related aspects of nuclear safety to stakeholders, including 

Federal, State, and local authorities, licensees, and the general public.

The SOARCA project sought to focus its resources on the more important severe accident 

scenarios for Peach Bottom and Surry. The project narrowed its approach by using an accident 

sequence’s possibility of damaging reactor fuel, or core damage frequency (CDF), as a surrogate 
xii

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

assessments (PRAs). Core damage sequences from previous staff and licensee PRAs were 

identified and binned into core damage groups. A core damage group consists of core damage 

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

containment or engineered safety feature operability. It is important to note that each core 

damage sequence that belongs to a given core damage group is initiated by a specific cause (for 

example, a seismic event, a fire, or a flood), and that the frequency of each core damage group 

was estimated by aggregating the CDFs of the individual sequences that belong to the group.

This approach was taken to help ensure that the contributions from all core damage sequences 

were accounted for during the sequence selection process. During the consequence analysis, the 

core damage groups for station blackouts were analyzed as if they were initiated by a seismic 

event. This approach was taken because seismically induced equipment failures occur 

immediately following the seismic event, which produces the most severe challenge to the 

plant. 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 10-6 (1 in a 

million) per reactor-year. SOARCA also sought to analyze scenarios leading to an early failure 

or bypass of the containment with a CDF equal to or greater than 10-7 (1 in 10 million) per 

reactor-year, 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. 

The staff used updated and benchmarked standardized plant analysis risk (SPAR) models and 

available plant-specific external events information in the scenario-selection process and 

identified two major groups of accident scenarios for analysis. The first group common to both 

Peach Bottom and Surry includes short-term station blackout (STSBO) and long-term station 

blackout (LTSBO). Both types of SBOs involve a loss of all alternating current (ac) power. The 

STSBO also involves the loss of turbine-driven systems through loss of direct current (dc) 

control power or loss of the condensate storage tank and therefore proceeds to core damage more 

rapidly (hence “short term”). The STSBO has a lower CDF, since it requires a more severe 

initiating event and more extensive system failures. SBO scenarios can be initiated by external 

events such as a fire, flood, or earthquake. SOARCA assumes that an SBO is initiated by a 

seismic event since this is the most extreme case in terms of both the timing and amount of 

equipment that fails. Notwithstanding the SOARCA scenario screening process, SBO scenarios 

are commonly identified as important contributors in PRA because of the common cause of 

failure for both reactor safety systems and containment safety systems. 

SOARCA’s second severe accident scenario group, which was identified for Surry only, is the 

containment bypass scenario. For Surry, two containment bypass scenarios were identified and 

analyzed. The first bypass scenario is a variant of the STSBO scenario, involving a thermallyinduced steam generator tube rupture (TISGTR). The second bypass scenario involves an 

interfacing systems loss-of-coolant accident (ISLOCA) caused by an unisolated rupture of lowhead safety injection piping outside containment. The CDF for the ISLOCA, 3×10-8 (3 in 100 

million) per reactor-year, falls below the SOARCA screening criterion for bypass events but it is 

analyzed for completeness because NUREG-1150 identified ISLOCA, in addition to SBO and 

SGTRs, as principal contributors to mean early and latent cancer fatality risks [2]. This scenario-
xiii

selection process captured the more important internally and externally initiated core damage 

scenarios.

SOARCA’s analyses were performed with two computer codes, MELCOR for accident 

progression and the MELCOR Accident Consequence Code System, Version 2 (MACCS2) for 

offsite consequences. The NRC staff’s preparations for the analyses included extensive 

cooperation from the licensees of Peach Bottom and Surry to develop high-fidelity plant systems 

models, define operator actions including the most recently developed mitigation actions, and 

develop models for simulation of site-specific and scenario-specific emergency planning and 

response. Moreover, in addition to input for model development, licensees provided information 

on accident scenarios from their PRAs. Through tabletop exercises of the selected scenarios 

with senior reactor operators, PRA analysts, and other licensee staff, licensees provided input on 

the timing and nature of the operator actions to mitigate the selected scenarios. The licensee 

input for each scenario was used to develop assumed timelines of operator actions and equipment 

configurations for implementing available mitigation measures which include mitigation 

measures beyond those routinely credited in current PRA models. A human reliability analysis, 

commonly included in PRAs to represent the reliability of operator actions, was not performed 

for SOARCA, but instead tabletop exercises, plant walkdowns, simulator runs and other inputs 

from licensee staff were employed to ensure that operator actions and their timings were 

correctly modeled. 

SOARCA modeled mitigation measures, including those in emergency operating procedures 

(EOPs), severe accident management guidelines (SAMGs), and Title 10 to the Code of Federal 

Regulations (10 CFR) 50.54(hh). The 10 CFR 50.54(hh) mitigation measures refer to additional 

equipment and strategies required by the NRC following the terrorist attacks of September 11, 

2001, to further improve each plant’s capability to mitigate events involving a loss of large areas 

of the plant caused by fire and explosions. To assess the benefits of 10 CFR 50.54(hh) mitigation 

measures and to provide a basis for comparison to the past analyses of unmitigated severe 

accident scenarios, the SOARCA project also analyzed each scenario without 10 CFR 50.54 (hh) 

equipment and procedures. The analysis that credits successful implementation of the 10 CFR 

50.54 (hh) equipment and procedures in addition to actions directed by the EOPs and SAMGs is 

referred to as the mitigated case. The analysis without 10 CFR 50.54(hh) equipment and 

procedures is referred to as the unmitigated case (SAMGs were considered but not implemented 

in the unmitigated case). The unmitigated case of the Surry ISLOCA is an exception to this 

general principle because it was necessary to assume that at least one of the EOP actions failed to 

occur for the scenario to lead to core damage. Chapter 3 of NUREG/CR-7110, Volume 1, 

“SOARCA Peach Bottom Integrated Analysis” and Volume 2, “SOARCA Surry Integrated 

Analysis”, details the specific equipment and operator actions credited for each scenario.

 

For the LTSBO scenarios for both Peach Bottom and Surry (the most likely severe accident 

scenario for each plant considered in SOARCA) analyzed assuming no mitigation, core damage 

begins in 9 to 16 hours, and reactor vessel failure begins at about 20 hours. Offsite radiological 

release due to containment failure begins at about 20 hours for Peach Bottom (BWR) and at 45 

hours for Surry (PWR). The SOARCA analyses therefore show that time may be available for 

operators to take corrective action and get additional assistance from plant technical support 

centers even if initial efforts are assumed unsuccessful. For the most rapid events (i.e., the 
xiv

unmitigated STSBO in which core damage may begin in 1 to 3 hours), reactor vessel failure 

begins at roughly 8 hours, possibly allowing time to restore core cooling and prevent vessel 

failure. In these cases, containment failure and radiological release begins at about 8 hours for 

Peach Bottom and at 25 hours for Surry. For the unmitigated Surry ISLOCA, the offsite 

radiological release begins at about 13 hours and in the other bypass event analyzed, the 

TISGTR, the radiological release begins at about 3.5 hours but is shown by analyses to be 

substantially smaller than the 1982 Siting Study SST1 release.

 

In addition to delayed radiological releases relative to the 1982 Siting Study SST1 case, the 

SOARCA study demonstrates that the amount of radioactive material released is much smaller as 

shown in Figures 1 (Iodine-131) and 2 (Cesium-137) below. The Surry ISLOCA iodine release 

is calculated to be 16 percent of the core inventory, but the results are more generally in the 

range of 0.5 to 2 percent for iodine and cesium for the other scenarios analyzed. By contrast, the 

1982 Siting Study SST1 case calculated an iodine release of 45 percent and a cesium release of 

67 percent of the core inventory. 

Figure ES-1 Iodine release to the environment for SOARCA unmitigated scenarios and the 

1982 Siting Study SST1 case
xv

Figure ES-2 Cesium release to the environment for SOARCA unmitigated scenarios and 

the 1982 Siting Study SST1 case

Past PRAs and consequence studies showed that sequences involving large early releases were 

important risk contributors. For example, the PWR SBO with a TISGTR was historically 

believed to result in a large, relatively early release potentially leading to higher offsite 

consequences. However, MELCOR analysis of Surry performed for SOARCA shows that the 

release is small, because other reactor coolant system piping inside containment (i.e., hot leg 

nozzle) fails soon after the tube rupture and thereby retains the fission products within the 

containment. Additional work would be needed to determine if this result generally applies for 

all types of PWRs.

While this report does not determine the respective likelihoods of the mitigated and unmitigated 

cases of each scenario, the SOARCA results demonstrate the potential benefits of employing 10 

CFR 50.54(hh) mitigation enhancements for the scenarios analyzed. MELCOR analyses were 

used both to confirm the time available to implement mitigation measures and to confirm that 

those measures, once taken, are effective in preventing core damage or significantly reducing 

radiological releases. When successful mitigation is assumed, the MELCOR results indicate no 

core damage for all scenarios except the Surry STSBO and its TISGTR variant. The securityrelated mitigation measures that provide alternative ac power and portable diesel-driven pumps 

are especially helpful in counteracting SBO scenarios. For the Surry STSBO and its TISGTR

variant, the mitigation is sufficient to flood the containment through the containment spray 

system to cover core debris resulting from vessel failure. For the ISLOCA scenario, installed 

equipment unrelated to 10 CFR 50.54(hh) is effective in preventing core damage owing to the 

time available for corrective action.

For scenarios that release radioactive material to the environment, MACCS2 uses site-specific 

weather data to predict the downwind concentration of material in the plume and the resulting 
xvi

population exposures and health effects. The analysis of offsite consequences in SOARCA 

incorporates the improved modeling capability reflected in the MELCOR and MACCS2 codes as 

well as detailed site-specific public evacuation models. These models were developed for each 

scenario based on site-specific emergency preparedness programs and State emergency response 

plans to reflect timing of onsite and offsite protective action decisions and the evacuation time 

estimates and road networks at Peach Bottom and Surry. Scenarios that are assumed to be 

initiated by a seismic event consider the earthquake’s impact on implementing emergency plans 

from loss of infrastructure (i.e., long-span bridges, traffic signals, sirens). 

The unmitigated versions of the scenarios analyzed in SOARCA have lower risk of early 

fatalities than calculated in the 1982 Siting Study SST1 case. SOARCA’s analyses show 

essentially zero risk of early fatalities. Early fatality risk was calculated to be ~ 10-14 for the 

unmitigated Surry ISLOCA (for the area within 1 mile of Surry’s exclusion area boundary) and 

zero for all other SOARCA scenarios. In comparison, 92 early fatalities for Peach Bottom and 

45 early fatalities for Surry were calculated for the SST1 case in the 1982 Siting Study.

SOARCA results indicate that bypass events (e.g., Surry ISLOCA) do not pose a higher 

scenario-specific latent cancer fatality risk than non-bypass events (e.g., Surry SBO). While 

consequences are greater when the bypass scenario happens, this is offset by the scenario being 

less likely to happen. SOARCA reinforces the importance of external events relative to internal 

events and the need to continue ongoing work related to external events risk assessment.

Offsite radiological consequences were calculated for each scenario expressed as the average 

individual likelihood of an early fatality and latent cancer fatality. Tables ES-1 (Peach Bottom) 

and ES-2 (Surry) show, for both mitigated and unmitigated cases, conditional (on the occurrence 

of the core damage scenario) scenario-specific probabilities of a latent cancer fatality for an 

individual located within 10 miles of the plant. Tables ES-1 and ES-2 show the results using the 

linear no-threshold (LNT) dose-response model, which assumes that the health risk is directly 

proportional to the exposure and even the smallest radiation exposure carries some risk. The 

tables also provide the scenario-specific latent cancer fatality risk for an individual located 

within 10 miles of the plant, taking into account the scenario’s core damage frequency. 
xvii

Table ES-1 Offsite Consequence Results for Peach Bottom Scenarios Assuming Linear NoThreshold (LNT) Dose-Response Model

Scenario

Core damage 

frequency 

(CDF)

(per 

reactor-year)*

Mitigated Unmitigated

Conditional 

scenariospecific 

probability of 

latent cancer 

fatality for 

an individual 

located 

within 

10 miles

Scenariospecific risk

(CDF x 

Conditional)

of latent cancer 

fatality for an 

individual 

located within 

10 miles 

(per reactoryear)

Conditional 

scenariospecific 

probability of 

latent cancer 

fatality for an 

individual 

located within 

10 miles

Scenario-specific 

risk

(CDF x 

Conditional)

of latent cancer 

fatality for an 

individual 

located within 

10 miles 

(per reactoryear)

Long-term 

SBO 3×10-6 No Core Damage 9×10-5 ~ 3×10-10 ****

Short-term 

SBO with 

RCIC 

Blackstart**

3×10-7

No Core Damage *** 7×10-5 ~ 2×10-11  

Short-term 

SBO without 

RCIC 

Blackstart

Not Applicable ** * 2×10-4 ~ 6×10-11 ****

* The CDF assumes that 10 CFR 50.54(hh) equipment and procedures were not used.

** Blackstart of the reactor core isolation cooling (RCIC) system refers to starting RCIC without any ac or dc 

control power. Blackrun of RCIC refers to the long-term operation of RCIC without electricity, once it has 

been started. This typically involves using a portable generator to supply power to indications such as reactor 

pressure vessel (RPV) level to allow the operator to manually adjust RCIC flow to prevent RPV overfill and 

flooding of the RCIC turbine. STSBO RCIC blackstart and limited blackrun is credited as an unmitigated case 

for SOARCA purposes because the licensee has included its use in procedures. Past NRC severe accident 

analyses of STSBO scenarios did not credit blackstart of RCIC. A sensitivity calculation without blackstart was 

therefore performed to provide a basis for comparison to past analyses.

*** A scenario with 10 CFR 50.54(hh) mitigation, but without RCIC blackstart was not analyzed.

**** Estimated risks below 1 × 10-7 per reactor year 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.
xviii

Table ES-2 Offsite Consequence Results for Surry Scenarios Assuming LNT DoseResponse Model

Scenario

Core 

damage 

frequency 

[CDF]

(per 

reactoryear)*

Mitigated Unmitigated

Conditional 

scenariospecific 

probability of 

latent cancer 

fatality for an 

individual 

located within 

10 miles

Scenario-specific 

risk

[CDF x 

Conditional]

of latent cancer 

fatality for an 

individual 

located within 

10 miles 

(per reactoryear)

Conditional 

scenariospecific 

probability of 

latent cancer 

fatality for an 

individual 

located within 

10 miles

Scenario-specific 

risk

[CDF x 

Conditional]

of latent cancer

fatality for an 

individual 

located within 

10 miles 

(per reactoryear)

Long-term 

SBO 2×10-5 No Core Damage 5×10-5 ~ 7×10-10 ****

Short-term 

SBO 2×10-6 No Containment Failure ** 9×10-5 ~ 1×10-10  

Short-term 

SBO with 

TISGTR

4×10-7 3×10-4 ** * ~ 1×10-10 **** 3×10-4 ~ 1×10-10  

Interfacing 

systems LOCA 3×10-8 No Core Damage 3×10-4 ~ 9×10-12 ** **

* The CDF assumes that 10 CFR 50.54(hh) equipment and procedures were not used.

** Accident progression calculations showed that source terms in the mitigated case are smaller than in the 

unmitigated case. Offsite consequence calculations were not run, since the containment fails at about 

66 hours. 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 ample time for mitigation through measures transported from offsite. 

*** Containment failure is delayed by about 46 hours in the mitigated case relative to the unmitigated case. 

Rounding to one significant figure shows conditional LCF probabilities of 3×10-4 for both mitigated and 

unmitigated cases, however the original values were 2.8×10-4 for the mitigated case and 3.2×10-4 for the 

unmitigated case.

**** Estimated risks below 1 × 10-7 per reactor year 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.

LCF risks using alternate dose-response models, as well as LCF risks for circular areas out to a 

radius of 50 miles, are also presented. Using a dose-response model that truncates annual doses 

below normal background levels (including medical exposures) results in a further reduction to 

the latent cancer fatality risks (by a factor of 100 for smaller releases and a factor of 3 for larger 

releases). Latent cancer fatality risk calculations are generally dominated by long-term exposure 
xix

to small annual doses (~500 mrem per year corresponding to state return criteria) by evacuees 

returning to their homes after the accident and being exposed to residual radiation over a long 

period of time. SOARCA’s calculated LCF risk results are smaller than extrapolations of 1982 

Siting Study SST1 LCF risk results. However, the difference diminishes when considering 

larger areas, out to a distance of 50 miles from the plant. 

Figure 3 compares SOARCA’s scenario-specific latent cancer fatality risks for an individual 

within 10 miles of the plant to the NRC Safety Goal [72] and to an extrapolation of the 1982 

Siting Study SST11 results. 

Figure ES-3 Comparison of average individual LCF risk results for SOARCA mitigated 

and unmitigated scenarios to the NRC Safety Goal and to extrapolations of 

the 1982 Siting Study SST1 (plotted on logarithmic scale)

The NRC Safety Goal for latent cancer fatality risk from nuclear power plant operation (i.e., 

2x10-6 or two in one million) is set 1,000 times lower than the sum of cancer fatality risks 

resulting from all other causes (i.e., 2x10-3 or two in one thousand). The calculated cancer 

fatality risks from the selected, important scenarios analyzed in SOARCA are thousands of times 

 

1 The Siting Study did not calculate LCF risks. Therefore, to compare the Siting Study SST1 case to LCF 

results for SOARCA, the SST1 source term was put into the MACCS2 offsite consequence code files for 

the Peach Bottom and Surry unmitigated STSBO calculations.
xx

lower than the NRC Safety Goal and millions of times lower than the general U.S. cancer fatality 

risk [73].

Comparisons of SOARCA’s calculated LCF risks to the NRC Safety Goal [72] and the average 

annual US cancer fatality risk from all causes [73] 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 reader should be aware. 

Relative to the safety goal comparison, the safety goal is intended to encompass all accident 

scenarios. SOARCA does not examine all scenarios typically considered in a PRA, even though 

it includes the important scenarios. SOARCA represents a mix of limited PRA models with a 

deterministic treatment of various long-term mitigating features. In fact, any analytical 

technique, including PRAs, will have inherent limitations of scope and method. As a result, 

comparison of SOARCA’s scenario-specific calculated LCF risks to the NRC Safety Goal is 

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

results in the ~ 10-10 range to approximate a summary risk from all scenarios, would yield a 

summary result that is also below the NRC Safety Goal of 2x10-6 or two in one million.

Relative to the U.S. average 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.

The SOARCA analyses show that emergency response programs, implemented as planned and 

practiced, reduce the scenario-specific risk of health consequences among the public during a 

severe reactor accident. Sensitivity analyses of seismic impacts on site-specific emergency 

response (e.g., loss of bridges, traffic signals, and delayed notification) at Peach Bottom and 

Surry do not significantly affect LCF risk. 

In summary, the staff believes SOARCA has achieved its objective to develop a body of 

knowledge regarding detailed, integrated, state-of-the-art modeling of the more important severe 

accident scenarios for Peach Bottom and Surry. SOARCA analyses indicate that successful 

implementation of existing mitigation measures can prevent reactor core damage or delay or 

reduce offsite releases of radioactive material. All SOARCA scenarios, even when unmitigated, 

progress more slowly and release much less radioactive material than the 1982 Siting Study 

SST1 case. As a result, the calculated risks of public health consequences from severe accidents 

modeled in SOARCA are very small.

The SOARCA study was nearing completion when the Fukushima Daiichi accident occurred on 

March 11, 2011. The Fukushima accident has many similarities and differences with some of 

the Peach Bottom severe accident scenarios analyzed in SOARCA. While there are significant 

gaps in information and uncertainties regarding what occurred in the Fukushima reactors, an 

appendix to this report compares and contrasts the SOARCA study and the Fukushima accident

based on currently available information for the following topics: (1) operation of the RCIC 

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

multiunit risk, and (5) spent fuel pool risk. 
xxi

ACKNOWLEDGEMENTS

The SOARCA project and this document are the result of the efforts of many NRC staff and the 

staff of its contractors, Sandia National Laboratories, and dycoda, LLC. The project has also 

benefited greatly from the contributions of senior risk analysts and selected other NRC staff 

members in specific areas of the project. The SOARCA team also recognizes the support 

provided by staff of the licensees of the two pilot plants, Peach Bottom and Surry.

The following individuals were members of the SOARCA project team.

Dorothy Andreas (née Collins)

Jonathan Barr

Terry Brock

Scott Burnell 

Richard Chang

Donald Chung 

Donald Dube 

Tina Ghosh

Anthony Huffert

Chris Hunter

Jocelyn Mitchell

AJ Nosek 

José Pires

Robert Prato

Jason Schaperow

Martin Stutzke

Randolph Sullivan

Charles Tinkler

James Vail

Thomas Weaver

The following individuals provided management oversight:

Kathy Halvey Gibson

Patricia Santiago

Brian W. Sheron

Jennifer Uhle

Jimi Yerokun

The following members of the SOARCA Steering Committee provided senior leadership:

Marc Dapas

William M. Dean

Daniel Dorman

Jack Grobe

Gary Holahan

Brian Holian

Robert Lewis

James Lyons

Cynthia Pederson

Jennifer Uhle
xxii

The external peer review committee provided invaluable feedback on the project. The members 

were:

Ken Canavan (Electric Power Research Institute)

Bernard Clément (Institut de Radioprotection et de Sûreté Nucléaire)

Jeff Gabor (ERIN Engineering)

Robert Henry (Fauske and Associates)

Roger Kowieski (Natural and Technological Hazards Management Consulting, Inc.)

David Leaver (WorleyParsons Polestar)

Bruce Mrowca (Information Systems Laboratories)

Kevin O’Kula (Washington Safety Management Solutions)

John Stevenson (JD Stevenson Consulting Engineering Company)

Karen Vierow (Texas A&M University)

Jacquelyn Yanch (Ioneyes, Inc. and Massachusetts Institute of Technology)
xxiii

ACRONYMS and ABBREVIATIONS

ac

AEC

alternating current

Atomic Energy Commission

ATWS

BEF

BEIR

BWR

anticipated transient without scram

biological effectiveness factor

Biological Effectiveness of Ionizing Radiation

boiling-water reactor

CDF core damage frequency

CFR

CRAC

Code of Federal Regulations

Calculation of Reactor Accident Consequences

Cs

CST

cesium

condensate storage tank

dc direct current

DOE

ECCS

U.S. Department of Energy

emergency core cooling system

ECST

EOF

EOP

EPA

emergency condensate storage tank

emergency operating facility

emergency operating procedure

U.S. Environmental Protection Agency

EPR

EPZ

Evolutionary Power Reactor

emergency planning zone

ESBWR

ETE

FGR

economic simplified boiling-water reactor

evacuation time estimate

Federal guidance report

FR

GNF

Federal Register

Global Nuclear Fuel

gpm

HPCI

gallons per minute

high-pressure coolant injection

HPS Health Physics Society

hr

I

IAEA

ICRP

hour

iodine

International Atomic Energy Agency

International Commission on Radiological Protection

INPO

kg

KI

IPEEE

ISLOCA

Institute of Nuclear Power Operations

kilogram

potassium iodide

individual plant examination of external events

interfacing systems loss-of-coolant accident

LCF latent cancer fatality

LHSI low-head safety injection

LNT linear no-threshold

LOCA loss-of-coolant accident

LOOP loss of offsite power

LTSBO long-term station blackout

LWR light-water reactor
xxiv

MACCS2 MELCOR Accident Consequence Code System, Version 2

m/s

MOX

MTU

MW

meters per second

mixed oxide

metric ton of uranium

megawatts 

MWd megawatt days

NCRP National Council on Radiation Protection and Measurements

NPP nuclear power plant

NRC U.S. Nuclear Regulatory Commission

NRF National Response Framework

ORNL Oak Ridge National Laboratory

ORO offsite response organizations

PGA peak ground acceleration

PRA

PWR

probabilistic risk assessment

pressurized-water reactor

RCIC reactor core isolation cooling

RCS reactor coolant system

RG

RPV

regulatory guide

reactor pressure vessel

RWST refueling water storage tank

SAMG severe accident management guideline

SBO station blackout

SGTR steam generator tube rupture

SGTS

SNL

standby gas treatment system

Sandia National Laboratories

SOARCA State-of-the-Art Reactor Consequence Analyses 

SPAR standardized plant analysis risk

SSE

SST

safe shutdown earthquake

siting source term

STCP Source Term Code Package

STSBO short-term station blackout

Sv

TDAFW

sieverts

turbine-driven auxiliary feedwater

TID

TISGTR

Technical Information Document

thermally induced steam generator tube rupture

TMI

TRANS

TSC

Three Mile Island

transients

technical support center

UFSAR updated final safety analysis report

UO2 uranium dioxide
1

1.0 INTRODUCTION

This document describes the U.S. Nuclear Regulatory Commission’s (NRC’s) state-of-the-art, 

realistic assessment of the accident progression, radiological releases, and offsite consequences 

for important severe accident sequences. 

The overall objective of the State-of-the-Art Reactor Consequence Analyses (SOARCA) project 

is to develop a body of knowledge on the realistic outcomes of severe reactor accidents. The 

results from the SOARCA project to date provide an updated reference of the likely outcomes of 

severe reactor accidents at the Peach Bottom and Surry nuclear power sites, based on the most 

current emergency preparedness and plant capabilities. The NRC also anticipates that the study 

will be a resource for future modeling improvements and verification efforts.

1.1

The evaluation of accident phenomena and offsite consequences of severe reactor accidents has 

been the subject of considerable research. Most recently, with Commission guidance and as part 

of plant security assessments, updated analyses of severe accident progression and offsite 

consequences were completed using the wealth of accumulated research. These analyses are

more detailed (in terms of the fidelity of the representation and resolution of facilities and 

emergency response), realistic (in terms of the use of currently accepted phenomenological 

models and procedures), and integrated (in terms of the intimate coupling between accident 

progression and offsite consequence models). 

Background

The results of those security-related studies confirmed and quantified what was suspected but not 

well-quantified—namely, that some past studies were conservative to the point that predictions 

were not useful for characterizing results. The communication of risk attributable to severe 

reactor accidents should properly consider realistic estimates of the more likely outcomes and 

should reflect both the many improvements and changes to plants and the advances in 

understanding of severe accident behavior.

In addition to the improvements in understanding and calculational capabilities that have resulted 

from these studies, numerous influential changes have occurred in the training of operating 

personnel and the increased use of plant-specific capabilities. These changes include the 

following:

• The transition from event-based to symptom-based emergency operating procedures 

(EOPs) for the boiling-water reactor (BWR) and pressurized-water reactor (PWR) 

designs.

• The performance and maintenance of plant-specific probabilistic risk assessments (PRAs) 

that cover the spectrum of accident scenarios.

• The implementation of plant-specific, full-scope control room simulators to train 

operators.
2

• An industrywide technical basis, owners-group-specific guidance, and plant-specific 

implementation of the severe accident management guidelines (SAMGs).

• Additional safety enhancements, described in Title 10, Section 50.54(hh) of the Code of 

Federal Regulations (10 CFR 50.54(hh)). These enhancements are intended to be used to 

maintain or restore core cooling, containment, and spent fuel pool cooling capabilities 

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

or fire, to include strategies in the following areas: (i) fire fighting; (ii) operations to 

mitigate fuel damage; and (iii) actions to minimize radiological release. For the 

SOARCA scenarios, successful implementation of this equipment and procedures would 

prevent core damage or delay or prevent the release.

• Improved phenomenological understanding of influential processes such as the 

following:

− in-vessel steam explosions 

− Mark I containment drywell shell attack 

− dominant chemical forms for fission products 

− direct containment heating 

− hot-leg creep rupture 

− reactor pressure vessel (RPV) failure and molten core-concrete interactions

Additional changes in plant operation have occurred over time, including the following: 

• power uprates

• higher core burnups

1.2

The overall objective of the SOARCA project is to develop a body of knowledge regarding the 

realistic outcomes of severe reactor accidents. Corresponding and supporting objectives are as 

follows:

Objective

• Incorporate the significant plant improvements and changes not reflected in earlier 

assessments, including system improvements, training and emergency procedures, offsite 

emergency response, and recent security-related enhancements described in 

10 CFR 50.54(hh), as well as plant changes in the form of power uprates and higher core 

burnup.

• Incorporate state-of-the-art integrated modeling of severe accident behavior, which 

includes the insights of several decades of research into severe accident phenomenology 

and radiation health effects.

• Evaluate the potential benefits of recent security-related mitigation improvements in 

preventing core damage and reducing or delaying an offsite release, should one occur.
3

• Enable the NRC to communicate severe-accident-related aspects of nuclear safety to 

stakeholders, including Federal, State, and local authorities; licensees; and the general 

public.

• Update quantification of offsite consequences found in earlier NRC publications, such as 

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

December 1982 [1].

1.3

The approach was to use the detailed, integrated, phenomenological modeling of accident 

progression (reactor and containment thermal-hydraulic and radionuclide response) that is 

embodied in the MELCOR code, coupled with modeling of offsite consequences with the 

MELCOR Accident Consequence Code System, Version 2 (MACCS2) code, to predict the likely 

outcomes for the more significant, albeit still remote, core melt accidents. The basis for the 

selection of the events for analysis included insights from past and current PRAs and from 

research on accident behavior and important failure modes. The selection of events for 

quantification also properly included probability, to focus on more likely and important 

contributors. 

Approach

Figure 1 illustrates the four main elements of SOARCA (i.e., scenario selection, mitigative 

measures analysis, accident progression and source term, and offsite radiological consequences).

Figure 1 The State-of-the-Art Reactor Consequence Analyses process

SOARCA provides a new and useful tool to, at this juncture, focus on specific important events 

and quantify the plant and offsite response rigorously and realistically. This approach can 
4

complement and supplement other analytical methods to efficiently and explicitly address the 

benefits of additional mitigation in further reducing the likelihood of core damage and offsite 

consequences. The offsite consequence analyses were performed on a site-specific basis 

(reflecting site-specific population distributions, weather, and emergency preparedness).

Selection of events considered individual plant examinations,

1 individual plant examinations of 

external events (IPEEEs), standardized plant analysis risk (SPAR) models, and NUREG-1150, 

“Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants,” issued 1990 [2]. 

The plant modeling included information related to system and procedural plant improvements 

that were incorporated as part of the industry’s response to the NRC’s security initiatives 

(e.g., the purchase and development of procedures for diesel-driven pumps in response to 

10 CFR 50.54(hh) requirements), as well as necessary plant information.

1.4 Historical Perspectives

The following sections describe some of the important historical studies that preceded the 

SOARCA project.

1.4.1 WASH-1400 (NUREG-75/014), “Reactor Safety Study: An Assessment of Accident 

Risks in U.S. Commercial Nuclear Power Plants,” 1975

In the summer of 1972, the Atomic Energy Commission (AEC) initiated a major probabilistic 

study, “Reactor Safety Study: An Assessment of Accident Risks in U.S. Commercial Nuclear 

Power Plants” [3]. Professor Norman C. Rasmussen of the Massachusetts Institute of 

Technology served as the study director. Saul Levine of the AEC served as staff director of the 

AEC employees who performed the study with the aid of many contractors and consultants. 

The study team attempted to estimate the potential effects of light-water reactor (LWR) accidents 

on public health and safety. The report analyzed in detail one BWR, Peach Bottom Unit 2, and 

one PWR, Surry Unit 1, to estimate the likelihood and consequences of potential accidents. The 

team chose these plants, because they were the largest plants of each type that were about to start 

operation.

The study’s purpose was to quantify the risks to the general public from commercial nuclear 

power plant (NPP) operation and to compare those risks with nonnuclear risks to provide 

perspective. This required identification, quantification, and phenomenological analysis of a 

wide range of low-frequency, relatively high-consequence scenarios that had not previously been 

considered in much detail. The introduction at this point of the concept of “scenario” is 

significant; as noted above, many design assessments simply look at system reliability (success 

probability), given a design-basis challenge. The review of nuclear plant license applications did 

essentially this, culminating in findings that specific complements of safety systems were 

single-failure proof for selected design-basis events. Going well beyond this, WASH-1400

modeled scenarios leading to large radiological releases from each of the commercial NPPs 

considered. It considered highly complex scenarios involving the success and failure of many 

 1 As requested by the NRC in Generic Letter 88-20, “Individual Plant Examination for Severe Accident Vulnerabilities,” 

dated November 23, 1988, the utilities conducted risk analyses that considered the unique aspects of a particular NPP, 

identifying the specific vulnerabilities of the plant to severe accidents.
5

and diverse systems within a given scenario, as well as operator actions and phenomenological 

events.

The team adapted methods previously used by the U.S. Department of Defense and the National 

Aeronautics and Space Administration to predict the effect of failures of small components in 

large, complex systems. The overall methodology, PRA, is still used today.

The team first identified events that could potentially lead to core damage. It then used event 

trees to delineate possible sequences of successes or failures of systems provided to prevent core 

meltdown or the release of radionuclides, or both. Using fault trees, the team estimated the 

probabilities of system failures from available data on the reliability of system components. 

With these techniques, thousands of possible core melt accident sequences were assessed for 

their occurrence probabilities. Computational models developed as part of the overall effort 

calculated the public health and economic consequences of the identified severe accidents. 

The insights gained from WASH-1400 included (1) “the possible consequences of potential

reactor accidents are predicted to be no larger, and in many cases much smaller, than those of 

nonnuclear accidents,” (2) “the likelihood of reactor accidents is much smaller than that of many

non-nuclear accidents having similar consequences. All non-nuclear accidents examined in this 

study, including fires, explosions, toxic chemical releases, dam failures, airplane crashes, 

earthquakes, hurricanes and tornadoes, are much more likely to occur and can have 

consequences comparable to, or larger than, those of nuclear accidents,” and (3) “non-nuclear 

events are about 10,000 times more likely to produce large numbers of fatalities than nuclear 

plants.”

While the risks from nuclear power appear to be very low, the Reactor Safety Study 

(WASH-1400) did indicate that core melt accidents were more likely than previously thought 

(approximately 5×10-5 per reactor-year for Surry and Peach Bottom2

), and that LWR risks are 

mainly attributable to core melt accidents. The Reactor Safety Study also demonstrated the wide 

variety of accident sequences (initiators and ensuing equipment failures or operator errors or 

both) that can cause core melt. In particular, the report indicated that, for the plants analyzed, 

accidents initiated by transients or small loss-of-coolant accidents (LOCAs) were more likely to 

cause core melt than the traditional large design-basis LOCAs.

In addition to providing some quantitative perspective on severe accident risks, other significant 

WASH-1400 results helped increase the application of PRAs in the commercial nuclear power

arena. They showed, for example, that some of the more frequent, less severe initiating events 

(e.g., “transients”) lead to severe accidents at higher expected frequencies than do some of the 

less frequent, more severe initiating events (e.g., very large pipe breaks). This led to the 

beginning of the understanding of the level of design detail that a PRA must include, if the 

scenario set is to support useful findings (e.g., consideration of support systems and 

environmental conditions). Following the severe core damage event at Three Mile Island in 

1979, application of these insights gained momentum within the nuclear safety community, 

 2 This value is derived from the following statement in the WASH-1400 Executive Summary: “The [probability of 

melting the core] value obtained was about one 1 in 20,000 per reactor per year.”
6

leading eventually to a PRA-informed reexamination of the allocation of licensee and regulatory

safety resources. In the 1980s, this process led to some significant adjustments to safety 

priorities at NPPs; since the 1990s, the NRC has refocused its regulations on areas of plant safety 

where that attention is more risk important. 

1.4.2 NUREG/CR-2239, “Technical Guidance for Siting Criteria Development,” 1982

The NRC contracted with Sandia National Laboratories (SNL) to develop a technical guidance 

report for siting future reactors [1]. The agency requested guidance on (1) criteria for population 

density and distribution surrounding future sites and (2) standoff distances of plants from offsite 

hazards.

Because the work was primarily focused toward the development of generic siting criteria, 

uncoupled from specific plant design, five types of accidents, with assumed representative 

radiological source terms, were imposed on each plant in the 91-site study. The accidents or

“siting source term events” (SST events) were to be derived from the previous Reactor Safety 

Study (WASH-1400) [3], and each SST event would be assumed identical regardless of plant 

design. 

(1) SST1—Severe core damage. All safety systems and containment are lost after 1.5 hours. 

(2) SST2—Severe core damage. Containment systems (e.g., sprays, suppression pools) 

function to reduce radioactive release, but containment leakage is large after 3 hours.

(3) SST3—Severe core damage. Containment systems function, but there is small 

containment leakage (1 percent per day) after 1 hour.

(4) SST4—Modest core damage. Containment systems function but there is small 

containment leakage after 1⁄2 hour.

(5) SST5—Limited core damage. Containment functions as designed with minimal leakage.

The early fatality results for most of the 91 sites were similar because of the low population 

density close to the sites. Using the extremely large and rapid SST1 radiological source term 

with a population density of 50 persons per square mile resulted in 47 to 140 early fatalities and 

730 to 860 latent cancer fatalities (LCFs). For the release represented by SST2 events, the mean

values from typical plants were zero early fatalities and 95 to 140 LCFs.

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

Plants,” 1990

NUREG-1150 [2] documents the results of an extensive NRC-sponsored PRA. The study 

examined five plants representative of classes of reactor and containment designs to give an 

understanding of risks for these particular plants. Selected insights regarding the classes of 

plants were also obtained in the study. The improved PRA methodology used in the 

NUREG-1150 study greatly enhanced the understanding of risk at NPPs and is considered a 

significantly updated and improved revision to the Reactor Safety Study [3]. One improvement 
7

was the specific inclusion of an uncertainty estimate for the core damage frequency (CDF) and 

source term portions of the study. This uncertainty estimate was based on extensive use of 

expert elicitation. For the offsite consequence portion of the study, random weather sampling 

addressed the uncertainty in health effects caused by weather variability. 

The following five NPPs were analyzed in NUREG-1150: 

(1) Unit 1 of the Surry Power Station, a Westinghouse-designed, three-loop PWR reactor in a 

large, dry, subatmospheric containment building located near Williamsburg, VA

(2) Unit 1 of the Zion Nuclear Power Plant, a Westinghouse-designed, four-loop PWR 

reactor in a large, dry containment building located near Chicago, IL

(3) Unit 1 of the Sequoyah Nuclear Power Plant, a Westinghouse-designed, four-loop PWR 

reactor in an ice condenser containment building located near Chattanooga, TN

(4) Unit 2 of the Peach Bottom Atomic Power Station, a General Electric-designed, BWR-4 

reactor in a Mark I containment building located near Lancaster, PA

(5) Unit 1 of the Grand Gulf Nuclear Station, a General Electric-designed, BWR-6 reactor in 

a Mark III containment building located near Vicksburg, MS

The various accident sequences that contribute to the CDF from internal initiators can be 

grouped by common factors into categories. NUREG-1150 uses the accident categories depicted 

in Table 1 below: station blackout (SBO), anticipated transients without scram (ATWS), other 

transients (TRANS), interfacing system LOCAs (SG/IF Sys), and other LOCAs. The selection 

of such categories is not unique but merely a convenient way to group the results. 

Table 1 Summary of Core Damage Frequency from NUREG-1150

Plant 

Name

Internal Initiators External

Initiators

SBO ATWS TRANS SG/IF Sys LOCA†

Core 

Damage

Total/yr

Fire & 

Seismic

Surry 2.7×10-5 1.6×10-6 2.0×10-6 3.4×10-6 6.0×10-6 4.0×10-5 2.6×10-5

Peach 

Bottom 2.2×10-6 1.9×10-6 1.4×10-7 - 2.6×10-7 4.5×10-6 2.3×10-5

† The LOCA category shown here includes LOCAs that are initiated by pipe break events. Transient-induced 

LOCAs are included under the other categories.

1.5

The central focus of the SOARCA project was to introduce the use of a detailed, best estimate, 

self-consistent quantification of scenarios based on current scientific knowledge and plant 

capabilities. The essence of the analysis methodology is the application of the integrated severe 

accident progression modeling tool, the MELCOR code. The analysis used an improved offsite 

consequence (MACCS2) code, including both improved code input and updated 

Scope
8

scenario-specific emergency response. Because the priority of this work was to bring more 

detailed, best estimate, and consistent analytical modeling to bear in determining realistic 

outcomes of severe accident scenarios, the benefits of this state-of-the-art modeling could most 

efficiently be demonstrated by applying these methods to a set of the more important severe 

accident scenarios. Thus, the project elected to limit its analysis to a set of important accident 

scenarios considering both likelihood and potential consequences. The scenarios that were 

eventually selected (e.g., SBO, interfacing systems loss-of-coolant accident (ISLOCA), 

thermally induced steam generator tube rupture (TISGTR)) are, in fact, scenarios that were also 

considered to be important in recent and past probabilistic assessments. 

The following several classes of accident events were not considered as part of the SOARCA 

project:

• multiunit accidents 

• low-power and shutdown accidents

• extreme seismic events that lead directly to gross containment failure with simultaneous 

reactor core damage 

• spent fuel pool accidents

• security events

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 developed to date do not 

explicitly consider multiunit accidents, because the NRC policy is to apply the Commission’s 

“Safety Goals for the Operation of Nuclear Power Plants” (51 Federal Register (FR) 28044) [4]

and subsidiary risk acceptance guidelines (see Regulatory Guide (RG) 1.174, “An Approach for 

Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to 

the Licensing Basis” [5]) on a “per reactor” basis. Therefore, multiunit accidents were not 

evaluated in the SOARCA project. The results of the unmitigated scenario analyses in SOARCA 

suggest that consideration of multiunit events would not substantially alter the study findings 

regarding low individual risk, but explicit analysis would be required to confirm the conclusion. 

Low-power and shutdown accidents are potentially significant, because the plant configuration is 

altered—the containment may be open and the reactor safety systems may be realigned. 

However, offsetting mitigating attributes include a potentially much smaller decay heat level and 

low pressure that allows for easier cooling of the reactor fuel. In this area, SOARCA has 

focused on the accidents that historically have received the most attention—the accidents 

initiated at full power. Also, one of the objectives was to provide an updated quantification of 

risk from past studies such as the Siting Study [1], and that study similarly was confined to 

full-power reactor events.

The SOARCA study excluded extreme seismic events that involve failure of the containment and 

lead to core damage. Seismic fragility quantification for these extreme and rare seismic events, 
9

in particular quantification of the size of a hole or amount of leakage, is currently subject to 

considerable uncertainty. More research is needed before undertaking a realistic, best estimate 

analysis of such rare events. 

Spent fuel pool accidents can contribute to overall risk associated with nuclear reactors, because 

significant quantities of spent fuel are stored onsite in such pools. Past NRC studies, including 

NUREG-1738, “Technical Study of Spent Fuel Pool Accident Risk at Decommissioning Nuclear 

Power Plants,” issued February 2001 [62], would suggest that risk from the most severe spent 

fuel pool accidents is low, yet the consequences of the release of a large inventory of cesium (Cs) 

and other radioisotopes could be serious. Since that time, the NRC has undertaken substantial 

analytical and experimental research to improve the modeling of spent fuel pool accidents, as 

well as research to identify significant improvements to spent fuel pool safety, as part of the 

NRC’s security-related research following the terrorist attacks of September 11, 2001. Based on 

the results of this research, the NRC concludes that spent fuel pool risk, which was assessed very 

conservatively in past studies such as NUREG-1738, is now much lower, based on both the new 

physical safety improvements required by the NRC and the improved modeling capability. 

Therefore, when developing the SOARCA project, the NRC elected to exclude spent fuel pool 

accidents from its scope. 

The NRC did not include security events as part of SOARCA to avoid providing any specific 

information that may materially assist in planning or carrying out a terrorist attack on an NPP. 

However, the NRC has stated that the security-related studies conducted after September 

11, 2001, led it to conclude that previous risk studies used conservative radionuclide source 

terms and that plant improvements plus improved modeling would confirm that radionuclide 

releases and early fatalities were substantially smaller than suggested by earlier studies.

Offsite consequences of severe nuclear reactor accidents could include economic and 

environmental damage in addition to harmful effects on human health. SOARCA calculates 

offsite consequences in terms of the risks of human fatalities for the specific scenarios. These 

risks are quantified as the individual risk of an early fatality and the individual risk of a latent 

cancer fatality. This enables comparison of SOARCA’s results to the NRC Safety Goal [72] and 

to the 1982 Siting Study’s results [1]. 

1.6

In the selection of important sequences, the SOARCA project ideally would have included those 

sequences found to be important to risk as demonstrated by a full-scope Level 3 PRA, which is 

an assessment of risk of offsite consequences in the event of a severe accident causing release of 

radioactive material to the environment. In practice, that was not feasible, because no current 

full-scope Level 3 PRAs (considering both internal and external events) were generally available 

to draw upon. However, the preponderance of Level 1 PRA information, combined with insights 

on severe accident behavior, is available on dominant core damage sequences, especially internal 

event sequences. This information, combined with the NRC’s understanding of containment 

loadings and failure mechanisms, and together with radionuclide release, transport, and 

deposition, allows the use of CDF as a surrogate criterion for risk. Thus, for SOARCA, the 

project team elected to analyze sequences with a CDF greater than 10-6 per reactor-year. In 

addition, the SOARCA team included sequences that have an inherent potential for higher 

Basis of Accident Selection
10

consequences (and risk) with a lower CDF (i.e., those with a frequency greater than 10-7 per 

reactor-year). Such sequences would be associated with events involving containment bypass or 

leading to an early failure of the containment. By adopting these criteria, the SOARCA team is

reasonably assured that the more probable and important core melt sequences will be captured. 

Further, SOARCA includes certain scenarios that had CDFs lower than the screening criteria, 

because of their historical significance. Thus, the selection of scenarios has a more generic 

application to plants with designs similar to Peach Bottom and Surry. 

1.7

An important objective of the SOARCA project was to assess the impact of severe accident 

mitigative features and reactor operator actions in mitigating an accident. This was done by 

evaluating in detail the operator actions and equipment that may be available (including 

10 CFR 50.54(hh) equipment).

Mitigated and Unmitigated Cases

Early in the project (2007), SOARCA staff visited the Peach Bottom Atomic Power Station and 

the Surry Power Station. During the visits, tabletop exercises were conducted for each scenario. 

Participants included plant senior reactor operators and PRA analysts. SOARCA staff provided 

initial and boundary conditions, elicited how plant staff would respond, and, through the tabletop 

exercises, developed a timeline of operator actions for each scenario. These assessments of 

mitigative measures were qualitative but, nonetheless, consisted of detailed scenario-specific 

consideration of systems and operations, based on licensee-identified mitigative measures from 

EOPs, SAMGs, 10 CFR 50.54(hh) measures, assistance from the technical support center (TSC),

and other severe accident guidelines that are applicable to and determined to be available during 

a specific scenario. The assessment of mitigation systems provided the basis for the assumptions 

on availability, capability, and timing used as input into the MELCOR analyses. For scenarios 

involving a seismic initiator, operator response times were lengthened to reflect the severity of 

the seismic event. 

A traditional human reliability assessment has not been performed to quantify the probabilities of 

plant personnel succeeding in implementing these measures. Therefore, each scenario was 

analyzed twice: a “mitigated” case assuming mitigative equipment was available and operable 

and operators were completely successful in implementing mitigative actions; and an 

“unmitigated” case assuming mitigation was not available, was not implemented, or was not 

effective. This report does not determine the respective likelihoods of the mitigated and 

unmitigated cases for each scenario. 

The NRC issued 10 CFR 50.54(hh) requiring plant licensees to possess the equipment, develop 

the strategies, and train plant personnel to implement these mitigative measures. The 

10 CFR 50.54(hh) measures are the result of a major effort by industry and the NRC in the 

2004–2008 timeframe to develop means to mitigate events involving a loss of large areas of the 

plant caused by fire and explosions. These mitigation measures were implemented by each plant 

on a per site basis rather than a per reactor basis, however some licensees have indicated plans to 

purchase additional equipment for the other unit. These measures are new and diverse and 

include the following major elements:

• procedures for manually operating turbine-driven injection (reactor core isolation cooling 

(RCIC) and turbine-driven auxiliary feedwater (TDAFW)) systems
11

• portable diesel-driven pumps for injecting into the reactor coolant system (RCS) (BWR) 

and steam generators (PWR)

• alternative means to depressurize

• portable power supplies for critical instrumentation such as reactor vessel water level

The assessment of mitigation measures has continued to receive attention since the initial 

assessment conducted with plant staff. The SOARCA team conducted additional site visits and 

system walkdowns in 2007, 2010, and 2011, with licensee personnel specifically reviewing the 

mitigation steps. The team used the results of accident progression calculations to characterize 

anticipated changes in plant conditions and describe the signatures of measurable parameters. It 

then estimated the time needed to assemble necessary personnel, tools, and equipment; align and 

start components; and establish a desired operating condition. SOARCA staff conducted 

followup site visits in June and August 2010 to explicitly address RCIC blackstart and blackrun 

for short-term station blackout (STSBO) and manual operation of TDAFW. The site visits 

included a review of RCIC blackstart and blackrun procedures, additional tabletop exercises to 

refine the PWR STSBO timeline, plant walkdowns of equipment areas, and detailed reviews of 

procedures. For the ISLOCA scenario, the licensee also had reactor operators use EOPs in a 

plant simulator to ensure timing for operator actions to be used in the SOARCA MELCOR 

calculations was accurate and reasonable. 

For each scenario and the mitigation measures identified, the team conducted detailed accident 

progression analyses to assess the efficacy of those measures. For each scenario, it also 

performed accident progression and offsite consequence analyses, assuming key mitigative 

measures were not taken, to demonstrate the relative importance and significance of those 

measures and to allow comparison of offsite consequence predictions with earlier studies. 

For each scenario, the project identified applicable mitigative measures that are potentially 

available (not eliminated by initial conditions). The systems and operations analyses were based 

on the initial conditions and anticipated subsequent failures to do the following:

• verify the availability of the primary system 

• determine the availability of support systems and equipment

• determine time estimates for implementation

Based on these scenario specifications, the team used MELCOR to determine the effectiveness 

of those mitigative measures that are expected to be available at a given time.
12

1.8 Uncertainty Analysis

The SOARCA project included a number of sensitivity studies to examine issues associated with 

accident progression, mitigation, and offsite consequences for the accident scenarios of interest. 

The objective of these sensitivity studies was to examine specific issues and ensure the 

robustness of the conclusions documented in this report. Single sensitivity studies, however, do 

not form a complete picture of the uncertainty associated with accident progression and offsite 

consequence modeling. Such a picture requires a more comprehensive and integrated evaluation 

of modeling uncertainties.

A follow-on uncertainty study will evaluate the impact of uncertainty by randomly sampling 

distributions for key model parameters that were considered to have a potential impact on the 

offsite consequences. The intended purpose of this uncertainty study is to develop insight into 

the overall uncertainty of the SOARCA results on scenario-specific risk to the combined and 

integrated uncertainty in accident progression (MELCOR) and offsite health effects (MACCS2) 

modeling. By addressing key MELCOR and MACCS2 modeling uncertainties in an integrated 

fashion, the SOARCA team believes it will further its understanding of the importance of this 

modeling on risk and thereby reveal where improvements in understanding are likely to be of 

benefit. (It will not address uncertainty in the scenario frequency.) Of principal interest is a 

comparison of the mean value, as determined by the uncertainty analysis, with the best estimate 

value of scenario-specific risk contained in this report.

1.9 Structure of NUREG-1935 and Supporting Documents

The SOARCA project is documented in multiple reports. This volume, NUREG-1935, describes 

the approach and procedures used in the study and summarizes the project results and 

conclusions. NUREG/CR-7110, Volumes 1 and 2, contain detailed descriptions of the 

plant-specific SOARCA analyses and results for the Peach Bottom and Surry plants, 

respectively. Because this volume and the NUREG/CR reports rely on highly technical 

explanations, an information brochure (NUREG/BR-0359, “Modeling Potential Reactor 

Accident Consequences”) was developed as a plain-language summary of SOARCA’s methods, 

results, and conclusions.

The SOARCA team assembled a panel of independent, external technical experts from industry, 

consulting, academia, and research laboratories to review the SOARCA analyses and assure their 

technical accuracy. The 11 members of the committee possess technical expertise in the fields of 

severe accident phenomenology and modeling; plant design, operation, and maintenance; 

mitigation measures; offsite emergency planning, preparedness, and response; radiological health 

consequences; seismic and structural analysis; and probabilistic risk assessment applications.

In addition to assuring technical accuracy, the committee also assessed whether the project’s 

conclusions were supported by the underlying technical work. The SOARCA team provided 

draft reports of NUREG-1935 and NUREG/CR-7110 Volumes 1 and 2 to the peer review 

committee at various points and held meetings with the members of the committee in July 2009, 

September 2009, March 2010, October 2010, and December 2011. During some of these 

meetings, NRC staff explained how peer reviewer comments were considered and addressed. 

The final letters from the individual members of the peer review committee are provided in 
13

Appendix B to this document. Individual letters rather than a consensus report were provided so 

that each member’s points of view could be fully expressed. In addition, Appendix B includes 

the NRC letter which provided resolutions of open peer review comments from the March 2010 

and October 2010 meetings.

The SOARCA project was nearly at the end of its peer review when the Fukushima Daiichi 

accident occurred in Japan on March 11, 2011. This accident presented real information 

regarding the progression of severe accidents and many insights with potential parallels to 

SOARCA’s analysis of SBO scenarios at Peach Bottom, a similarly designed plant. The 

SOARCA team developed Appendix A to this volume which qualitatively compares and 

contrasts specific accident phenomena based on information available to date. As additional 

information becomes available, the NRC will continue to review it for lessons learned and 

insights potentially applicable to nuclear plants in the United States. 

NUREG-1935 was released as a draft for public comments from January 31, 2012 through 

February 29, 2012. Comments related to the SOARCA project covered a wide range of topics. 

Appendix C to this document provides a summary of the different questions and comments 

received related to SOARCA along with NRC responses. The comments are related to the 

following general areas of the SOARCA project: project scope, scenario selection, MELCOR 

and accident progression analysis, emergency response analysis, and MACCS2 and offsite 

consequence analysis.

15

2.0 ACCIDENT SCENARIO SELECTION

An accident sequence begins with the occurrence of an initiating event (e.g., a loss of offsite 

power, a LOCA, or an earthquake) that perturbs the steady-state operation of the NPP. The 

initiating event challenges the plant’s control and safety systems, the failure of which could 

cause damage to the reactor fuel and result in the release of radioactive fission products. 

Because an NPP has numerous diverse and redundant safety systems, many different accident 

sequences are possible, depending on the type of initiating event that occurs, the amount of 

equipment that fails, and the nature of the operator actions involved. 

One way to systematically identify possible accident sequences is to develop accident sequence 

logic models using event tree analysis, as is done in PRAs. Pathways through an event tree 

represent accident sequences. Typically, the analysis is divided into two parts: (1) a Level 1 

PRA that represents the plant’s behavior from the occurrence of an initiating event until core 

damage occurs and (2) a Level 2 PRA that represents the plant’s behavior from the onset of core 

damage until radiological release occurs. The development of accident sequence logic models 

requires detailed information about the plant and the expertise of engineers and scientists from a 

wide variety of technical disciplines. As a result, the construction of accident sequence logic 

models is a complex and time-consuming activity.

The NRC and NPP licensees have already completed many PRAs. However, because of the 

improvements in PRA technology and plant capabilities and performance, this study gave more 

importance to the most current PRA information. 

2.1

Figure 1 illustrates the overall process used to identify and characterize accident scenarios for the 

SOARCA project. The SOARCA team selected scenarios from the results of existing PRAs. 

Some of these existing PRAs model accident sequences to the point of radiological release 

(i.e., they are Level 2 PRAs); however, the majority of existing PRAs are limited to the onset of 

core damage (i.e., Level 1 PRAs). The team identified core damage sequences from previous 

staff and licensee PRAs and separated them into core damage groups. A core damage group 

consists of core damage sequences that have similar timing for important severe accident 

phenomena and similar containment or engineered safety feature operability. The groups were 

screened according to their approximate CDFs to identify those that were the most significant. 

Finally, the accident scenario descriptions were augmented by assessing the status of 

containment systems (which are not typically modeled in Level 1 PRAs).

Approach
16

Figure 2 SOARCA accident scenario selection and analysis process

The scope of analyses using MELCOR and MACCS2 was generally confined to scenarios based 

on the following CDF screening guidelines:

• 10-6 per reactor-year for most scenarios

• 10-7 per reactor-year for scenarios that are known to have the potential for higher 

consequences (e.g., containment bypass scenarios such as steam generator tube rupture 

(SGTR) and ISLOCA initiators)

To accomplish this, the project grouped the release characteristics so that they are representative 

of scenarios binned into those groups. In addition, the groups are sufficiently broad to include 

the potentially risk significant but lower frequency scenarios. As a result of limitations in 

available Level 2 analyses and models, the team selected and screened the scenarios using CDF 

per reactor-year as the criterion, rather than radionuclide release frequency.

The application of the screening criteria to the available Level 1 PRA information for the pilot 

plants resulted in the identification of two basic types of scenarios: SBOs and bypass scenarios. 

This result presents certain advantages with respect to the inherent adequacy of the criteria and 

of the scope of scenarios. First, SBO scenarios are representative of a broad class of events in 

PRA—loss of heat removal events. Selection of SBO events in SOARCA ensures that the 

project covers that broader class of transients involving a loss of heat removal, and further, 

including an STSBO reasonably bounds the radionuclide release time and consequences of that 

class of accidents (which could include other events, such as loss of service water or loss of 

component cooling water but which develop more slowly). Also, for the PWR, the SBO 

includes, in part, the effect of a small LOCA by considering reactor coolant pump seal leakage. 

Additionally, selecting SBO sequences for analysis meant including the effects of loss of 

containment heat removal (fan coolers) and loss of containment spray systems (which are all 

electrically powered) to remove airborne radionuclides. Thus, the nonbypass sequences also 
17

result in containment failure, which would not be the case for all other transients involving such 

loss of heat removal in a typical PRA. Therefore, while SOARCA used CDF for screening, in 

effect, the CDF in these cases also represents the radionuclide release frequency. 

While the study did not include medium or large loss-of-inventory accidents—because of their 

very low frequency—it should be noted that such internal events are well below the screening 

criteria for the BWR and comfortably below the screening criterion for the PWR. For Peach 

Bottom, the medium and large LOCAs had CDFs of 2×10-9 and 1×10-9 per reactor-year. For 

Surry, the medium and large LOCAs had frequencies of 6×10-8 and 7×10-10 per reactor-year. 

Only a fraction of these sequences would have resulted in containment failure, because there 

may not have been a loss of containment heat removal. Since the Surry analyses included an 

ISLOCA sequence, it can also be argued that they reasonably bounded the radionuclide release 

time and consequences of events involving a LOCA inside containment for that plant.

The timing of a severe accident’s offsite release has a major impact on both early and LCF risks. 

In this respect, the team examined candidate SOARCA sequences with the timing of both core 

damage and containment failure in mind. As part of this consideration, it addressed, for the 

Peach Bottom plant, an additional sequence, the STSBO, even though it fell below the screening

criterion. The STSBO frequency is roughly an order of magnitude lower than the LTSBO 

(3×10-7 per reactor-year versus 3×10-6 per reactor-year); however, the STSBO has a more prompt 

radiological release and a slightly larger release over the same interval of time. The initial 

qualitative assessment of the STSBO concluded that it would not have greater risk significance 

than the LTSBO, because, while it has a more prompt release (8 hours versus 20 hours), the 

release is delayed beyond the time needed for successful evacuation. To demonstrate the points 

regarding risk versus frequency for lower frequency events, the study nonetheless included a 

detailed analysis of the STSBO. In a related fashion, the study included an ISLOCA sequence 

for Surry, even though it fell below the screening criterion of 1×10-7 per reactor-year for bypass 

scenarios. Past studies (e.g., NUREG-1150) cited this scenario as important, and it has the 

potential for larger releases because of its direct release outside the containment. 

Finally, the team routinely considered core damage initiators and phenomenological containment 

failure modes in SOARCA that were considered in the past, except for those that were excluded 

by extensive research (alpha mode failure, direct containment heating, and gross failure without 

prior leakage). The detailed analysis includes modeling behavior (including radionuclide 

transport and release) associated with long-term containment pressurization, Mark I liner failure, 

induced SGTR, hydrogen combustion, and core concrete interactions. 

SOARCA does not include analysis of an extreme earthquake that directly results in a large 

breach of the RCS (large LOCA), a large breach of the containment, and an immediate loss of 

safety systems. Given the considerable uncertainties in the quantification of seismic loads and 

seismic fragilities, in particular the quantification of the size of a hole or the amount of leakage, 

more research is needed to perform a best estimate analysis. In addition, it would not be 

sufficient to perform a nuclear plant risk evaluation of this event without also assessing the 

concomitant nonnuclear risk associated with such a large earthquake. This assessment would 

have to include an analysis of the impact on public health of an extremely large earthquake—
18

larger than that generally considered in residential or commercial construction codes—to provide 

the perspective on the relative risk posed by operation of the plant. 

Additionally, SOARCA considered whether the seismic events evaluated for Surry could cause 

liquefaction-induced settlements large enough to result in containment failure at containment 

penetrations. A review of previous work related to liquefaction at the Surry site and preliminary 

analyses assessed the potential for liquefaction-induced soil deformations. According to 

NUREG/CR-4550 [68], liquefaction is expected to occur for a seismic event greater than the 

safe-shutdown earthquake (SSE) at Surry. Estimated liquefaction-induced settlements provided 

in NUREG/CR-4550 range between 2 and 4 inches for a peak ground acceleration of 0.3 to 0.4 g. 

Using geotechnical data provided in the Surry updated final safety analysis report [69] and the 

original geotechnical investigation report by Dames and Moore [70], analyses for the SOARCA 

study resulted in similar settlement estimates in the vicinity of the containment structure, 

auxiliary building, and turbine building for a peak ground acceleration of 0.4 g. These estimated 

settlements are considered to be a mean estimate. A site examination performed by engineers for 

the NUREG/CR-4550 study of the piping systems and cable penetrations going from the 

auxiliary, safety area, service, and turbine buildings into the containment indicated that such 

displacements were not likely to cause failure. NUREG/CR-4550 did not provide the basis for 

this assessment, and this study did not include additional analyses on piping systems to confirm 

this assessment. Additional settlement analyses were performed for a peak ground acceleration 

of 0.75 g, which is associated with an event having an annual frequency of occurrence on the 

order of 1x10-6 to 1x10-7

. At this ground motion level, mean settlement estimates increase to 

between 4 and 8 inches [71]. The effects of this magnitude of settlement on piping systems have 

not been assessed in SOARCA. Because of the considerable uncertainties in the quantification 

of these effects for this magnitude of settlement estimates, more research is needed to perform a 

best-estimate analysis.

In summary, SOARCA addresses the more likely (though still remote) and important sequences 

that are understood to compose much of the severe accident reactor risk from nuclear plants. 

NRC staff conclude that the general methods of SOARCA (i.e., detailed, consistent, 

phenomenologically based, sequence-specific, accident progression analyses) are applicable to 

PRA methodology and should be the focus of improvements in that regard. 

2.2 Scenarios Initiated by Internal Events

The study identified scenarios initiated by internal events and the availability of containment 

systems for these scenarios using the NRC’s plant-specific SPAR models, licensee PRAs, and 

NUREG-1150 [2]. The SPAR models support the NRC’s oversight of licensed commercial 

NPPs and have been developed and maintained under a formal quality assurance program. The 

Peach Bottom SPAR model has been peer reviewed against staff-endorsed industry consensus 

PRA standards. Both the Surry and Peach Bottom licensee PRAs have been peer reviewed 

against the same standards. In addition, the SPAR model accident sequence results (including 

the sequence minimal cut sets) are periodically compared to the results from licensee PRAs 

under the Mitigating System Performance Index Program, which is part of the NRC’s Reactor 

Oversight Process. As a result, both the qualitative and quantitative results from the Surry and 
19

Peach Bottom SPAR models are in reasonable agreement with the corresponding licensee PRAs. 

Specific comparisons are discussed below.

The following process determined the scenarios for further SOARCA analyses:

• Candidate accident scenarios were identified in analyses using plant-specific SPAR 

models (Version 3.31).

− Initial Screening. Screened-out sequences with a CDF less than 10-8

, eliminating

4 percent of the overall CDF for Peach Bottom and 7 percent of the overall CDF 

for Surry.

− Sequence Evaluation. Identified and evaluated the dominant cutsets for the 

remaining sequences. Determined system and equipment availabilities and 

accident sequence timing.

− Scenario Grouping. Grouped sequences with similar times to core damage and 

equipment availabilities into scenarios.

• Containment systems availabilities for each scenario were assessed using system 

dependency tables that delineate the support systems required for performance of the 

target front-line systems and from a review of existing SPAR model system fault trees.

• Core damage sequences from the licensee PRA model were reviewed and compared with 

the scenarios determined by using the SPAR models. Differences were resolved during 

meetings with licensee staff.

• The screening criteria (CDF less than 10-6 for most scenarios and less than 10-7 for 

containment bypass sequences) were applied to eliminate scenarios from further analyses.

This process provides the basic characteristics of each scenario. However, it is necessary to have 

more detailed information about each scenario than is contained in a PRA model. Capturing the 

additional scenario details requires further analysis of system descriptions and a review of 

procedures. This review includes the analysis of mitigation measures beyond those treated in 

current PRA models. Mitigation measures treated in SOARCA include the plant-specific EOPs, 

SAMGs, and 10 CFR 50.54(hh) mitigation measures. Section 0 describes the mitigation 

measures assessment process used to determine what measures would be available and the 

associated timing to implement them.

2.3

As explained in Section 2.1, the SOARCA team considered and selected accident scenarios 

(sequence groups, rather than individual sequences) based on both likelihood and potential 

consequences. The team identified core damage sequences from previous staff and licensee 

PRAs and separated them into core damage groups. It then screened the groups (not individual 

sequences) according to their approximate CDFs to identify the most significant ones. Since 

Scenarios Initiated by External Events
20

core damage groups (i.e., scenarios) were considered, many individual lower order sequences 

would be captured in the aggregation into groups.

External events include internal flooding and fire; seismic events; extreme wind-, tornado-, and 

hurricane-related events; and similar events that may apply to a specific site. The external event 

scenarios developed for SOARCA analysis were derived from a review of past studies, such as 

the NUREG-1150 study [2], IPEEE submittals, and other relevant generic information. Detailed 

sequence characteristics are more difficult to specify for external event scenarios because of the 

general lack of external event PRA models industrywide. As a result, the SOARCA external 

event scenarios are heuristically based (i.e., experience based), as opposed to the internal event 

scenarios, which were developed through more formal, rigorous PRA methods. 

These scenarios were initiated by a seismic, fire, or flooding event. The mitigation measures 

assessment for each of these scenarios assumed that the initiator was a seismic event, because it 

was judged to be limiting. Seismic initiators are considered to be limiting for two principal 

reasons. First, they are more likely to result in the near immediate failure of systems, whereas, 

fire and flood would be expected to result in delayed failures. Secondly, a seismic event may be 

more likely than a fire or flood to fail passive components, such as water tanks. Additionally, 

seismic initiators may be more likely to have sitewide and offsite impacts. 

No attempt was made to match the frequencies of the external event scenarios to the actual 

sequence frequencies in any of the input information sources, because much of the available 

quantitative risk information on external events is dated. For example, since the publication of 

input information sources, new seismic hazard estimates have been developed. As a result, the 

estimated frequencies of the external event scenarios were based on expert judgment that 

considered the impact of changes in seismic data and methods on the published external-event 

PRA results. Care was taken to ensure that the external event scenario selection maintained the 

relative importance of external events CDF versus internal events CDF.

2.4

The SOARCA team selected four accident scenarios for the Surry plant (two initiated by internal 

events and two initiated by external events). The following sections identify each selected 

accident scenario, provide its representative CDF, and summarize the accident scenario in terms 

of its initiating event, equipment failures, and operator errors.

Accident Scenarios Selected for Surry

2.4.1 Surry Internal Event Scenarios

Two internal event scenarios for Surry met the criteria for further analysis.

(1) Initiating Event: Spontaneous SGTR

Representative CDF: 5×10-7 per reactor-year (SPAR)

Scenario Summary: This scenario is initiated by a spontaneous rupture in one steam 

generator tube. The operators fail to (1) isolate the faulted steam generator,

(2) depressurize and cool down the RCS, and (3) refill the refueling water storage tank 
21

(RWST) or cross-connect to the unaffected unit’s RWST. Auxiliary feedwater, 

high-pressure injection, low-pressure injection, and containment spray are available, if 

needed. However, high-pressure recirculation, low-pressure recirculation, and the 

recirculation sprays will be unavailable as a result of lack of water in the containment 

sump.

Comparison with Licensee PRA: The licensee PRA calculates a CDF of 1×10-6 per 

reactor-year for this scenario. The conditional core damage probabilities are virtually 

identical for the SPAR analysis (1.4×10-4

) and for the licensee PRA (1.5×10-4). The 

difference in the calculated CDFs is mainly attributable to the difference in initiating 

event frequency. Because both the SPAR model and licensee-calculated CDFs for this 

scenario are above the 1×10-7 per reactor-year threshold for containment bypass 

scenarios, this scenario was retained for further analysis.

(2) Initiating Event: ISLOCA in the Low-Head Safety Injection System

Representative Frequency: 3×10-8 per reactor-year (SPAR)

Scenario Summary: This scenario is initiated by a common-cause failure of both 

low-head safety injection (LHSI) inboard isolation check valves. The open pathway 

pressurizes and ruptures a section of the low-pressure piping outside the containment, 

which opens a containment bypass LOCA. This sequence group consists of the bypass 

LOCA, followed by operator failures to refill the RWST or cross-connect to the 

unaffected unit’s RWST. The ability to inject using the LHSI is not possible because of 

the pipe rupture. The high-head injection system remains available, because the pumps 

are in a separate location. Core damage occurs because of RWST depletion and operator 

failure to refill the RWST or cross-connect to the unaffected unit’s RWST.

Comparison with Licensee PRA: The ISLOCA scenario analyzed in SOARCA is a 

catastrophic failure of both of the inboard isolation check valve disks within the LHSI 

piping, together with failure to refill the RWST or to cross-connect to the unaffected 

unit’s RWST. For this ISLOCA scenario, the NRC’s SPAR model calculated a CDF of 

3x10-8 per reactor-year, and the NRC’s initial understanding was that the licensee’s PRA 

calculated a CDF of 7x10-7 per reactor-year. SOARCA analyses originally included this 

scenario because the licensee’s PRA for Surry included an ISLOCA frequency of 7x10-7

per reactor-year, and it has been commonly identified as an important contributor in 

PRA. 

During Surry site visits on January 19, 2011, and October 26, 2011, the NRC staff 

learned that the licensee‘s current PRA model has the following two ISLOCA scenarios: 

• scenario one: catastrophic failure of one check valve, leak-by of the second check 

valve, and the motor-operated isolation valve being unable to close 

• scenario two: catastrophic failure of two check valves 
22

Scenario one would result in a leak between 50–300 gallons per minute (gpm) from the 

RCS. Anything less than 50 gpm would be mitigated by a relief valve on the low 

pressure side of the LHSI injection line; pipe rupture would not occur. The frequency of 

the catastrophic failure of one check valve and the leak-by of the second check valve is 

1x10-6 per reactor-year. When compounded by all the potential failure modes (including 

operator error and mechanical or electrical failures) of the motor-operated valve, that 

lowers the frequency of scenario one to 7 x10-7 per reactor-year. This frequency does not 

include any consideration of averting core damage by refilling or cross-connecting 

RWSTs. This is a significant conservatism.

Scenario two would result in a leak above 300 gpm from the RCS. The licensee‘s current 

PRA model assumes that the probability for the catastrophic failure of both isolation 

check valves is approximately 3 x10-8 per reactor-year. As with scenario one, this 

frequency does not include consideration of averting core damage by refilling or 

cross-connecting RWSTs. Scenario two does not meet the SOARCA screening criterion 

of 1×10-7 per reactor-year for a bypass event. However, the team elected to retain it,

because it has been commonly identified as an important contributor in PRA.

2.4.2 Surry External Event Scenarios

Two external event scenarios for Surry met the criteria for further analysis.

(1) Initiating Event: Seismic-initiated LTSBO

Representative Frequency: 1×10-5 to 2×10-5 per reactor-year

Scenario Summary: This scenario is initiated by an earthquake of 0.3–0.5 g peak ground 

acceleration (PGA). The seismic event results in loss of offsite power (LOOP) and 

failure of onsite emergency alternating current (ac) power, resulting in an SBO event 

where neither onsite nor offsite ac power are recoverable. All systems dependent on ac

power are unavailable, including the containment systems (containment spray and fan 

coolers). The TDAFW system is available initially. Eventually, loss of the TDAFW 

occurs because of battery depletion and the resulting loss of direct current (dc) power for 

sensing and control. The loss of pump seal cooling will cause a reactor coolant pump 

seal to leak.

(2) Initiating Event: Seismic-Initiated STSBO

Representative Frequency: 1×10-6 to 2×10-6 per reactor-year

Scenario Summary: This scenario is initiated by an earthquake of 0.5–1.0 g PGA. The 

seismic event results in a LOOP and failure of onsite emergency ac power, resulting in an

SBO event where neither onsite nor offsite ac power are recoverable. All systems 

dependent on ac power are unavailable, including the containment systems (containment 

spray and fan coolers). The seismic event also results in a loss of dc power, resulting in 

the loss of automatic control of the TDAFW system. The earthquake ruptures the 
23

emergency condensate storage tank (ECST), which is conservatively assumed to empty 

immediately, rendering the TDAFW system initially unavailable. This scenario is 

referred to as the STSBO, since the site loses all power, even the batteries, and therefore 

all of the safety systems become quickly inoperable in the “short term.”

(3) Initiating Event: Seismic-Initiated STSBO with Induced SGTR 

Representative Frequency

[7]

: 3×10-7 to 5×10-7 per reactor-year. The representative 

frequency for this event is estimated to be 3.75×10-7 per reactor-year, based on an 

assumed conditional tube failure probability of 0.25, selected from NUREG-1570, “Risk 

Assessment of Severe Accident-Induced Steam Generator Tube Rupture,” issued 

March 1998 .

Scenario Summary

2.5

: An additional seismic-initiated STSBO scenario involves a variation 

that considers the conditional likelihood of a thermally induced steam generator tube 

rupture (TISGTR). 

The SOARCA team selected two accident scenarios for the Peach Bottom plant (both initiated by 

a seismic event). In addition, SOARCA included a mitigation assessment for the Loss of Vital 

AC Bus E-12 scenario. The following sections identify each selected accident scenario, provide 

its representative CDF, and summarize the scenario in terms of its initiating event, equipment 

failures, and operator errors.

Accident Scenarios Selected for Peach Bottom

2.5.1 Peach Bottom Internal Event Scenarios

The Loss of Vital AC Bus E-12 was initially estimated to have a frequency above the SOARCA 

screening criterion of 1×10-6 per reactor-year and was therefore analyzed. However, after further 

review of the SPAR model and comparison with the licensee’s PRA, the team determined that 

the scenario had a CDF below the screening criteria. Because the MELCOR analysis provided 

unique insights into the mitigation and response of the plant for this internal event sequence, the 

team retained the MELCOR analysis.

2.5.2 Peach Bottom External Event Scenarios

(1) Initiating Event: Seismic-Initiated LTSBO

Representative Frequency: 1×10-6 to 5×10-6 per reactor-year

Scenario Summary: This scenario is initiated by an earthquake of 0.3–0.5 g PGA. The 

seismic event results in a LOOP, failure of onsite emergency ac power, and failure of the 

Conowingo Dam power line, resulting in an SBO event where neither onsite nor offsite 

ac power are recoverable. All systems dependent on ac power are unavailable, including 

the containment systems (containment spray). The turbine-driven injection systems—

high-pressure coolant injection (HPCI) or RCIC, or both—are available until battery 

depletion. 
24

(2) Initiating Event: Seismic-Initiated STSBO 

Representative Frequency: 1×10-7 to 5×10-7 per reactor-year

Scenario Summary: This scenario is initiated by an earthquake of 0.5–1.0 g PGA. The 

seismic event results in a LOOP, failure of onsite emergency ac power, and failure of the 

Conowingo Dam power line, resulting in an SBO event where neither onsite nor offsite 

ac power are recoverable. All systems dependent on ac power are unavailable, including 

the containment systems (containment spray). In addition, HPCI and RCIC are initially 

assumed to be unavailable because of the loss of dc power. The larger earthquake 

ruptures the condensate storage tank (CST). The earthquake causes the fire water system 

to fail. This scenario is referred to as the short-term SBO since the site loses all power, 

even the batteries, and therefore all of the safety systems become quickly inoperable in 

the “short term.”

Note

2.6

: The STSBO scenario does not meet the SOARCA screening criterion of 1×10-6

per reactor-year; however, the team retained the scenario for analysis to assess the risk 

importance of a lower frequency, potentially higher consequence scenario. This type of 

scenario has been a risk-important severe accident scenario in past PRA studies. The 

SOARCA study analyzed two variations of this scenario, one with and one without RCIC 

blackstart. 

The results of existing PRAs indicate that the likelihood of an NPP accident sequence that 

releases a significant amount of radioactivity is very small, owing to the diverse and redundant 

barriers and numerous safety systems in the plant, the training and skills of the reactor operators,

testing and maintenance activities, and the regulatory requirements and oversight of the NRC. In 

addition, it is important to recognize that CDFs of NPPs have decreased over the years. Several 

reasons exist for these decreases:

Generic Factors

• Utilities have completed plant modifications intended to remedy concerns raised in 

earlier PRAs.

• Plants exhibit better performance as evidenced by reductions in initiating event 

frequencies, improvements in equipment reliability, and higher equipment availability. 

NPP equipment has become more reliable and available because of improved 

maintenance practices motivated by the Maintenance Rule (10 CFR 50.65, 

“Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power 

Plants”) [8].

• The NRC has issued new regulations, such as the Anticipated Transient without Scram 

(ATWS) Rule (10 CFR 50.62, “Requirements for Reduction of Risk from Anticipated 

Transients without Scram (ATWS) Events for Light-Water-Cooled Nuclear Power 

Plants”) [9] and the SBO Rule (10 CFR 50.63, “Loss of All Alternating Current 

Power”) [10] that directly affect the likelihood of certain types of accidents. Although 
25

the NRC issued the ATWS Rule and the SBO Rule before it completed 

NUREG-1150 [2], it did not address the impact of these rules on risk in NUREG-1150.

• PRA methodologies have improved, allowing a more realistic assessment of risk. In this 

category, improvements in common-cause failures analysis are noteworthy.

As a result, risk estimates reflect the impacts of constantly changing plant operational, 

regulatory, and PRA technology environments. Any attempt to identify significant accident 

sequences should be viewed as a “snapshot” of the plant at the time the analysis was completed.

27

3.0 MITIGATIVE MEASURES ASSESSMENT

The overall objective of the SOARCA project is to develop a body of knowledge regarding the 

realistic outcomes of severe reactor accidents. Included within this objective is to provide

insight into the effectiveness and benefits of mitigation measures currently employed at 

operating reactors. Section 2.0 describes the PRA information sources, including the NRC’s 

SPAR models, licensees’ PRA models, NUREG-1150, and additional expert judgment that this 

study used to identify risk-important sequence groups leading to core damage and containment 

failure or bypass. This section describes the methods that determined the mitigation measures 

that would be available and the associated timing to implement them. This includes mitigation 

measures beyond those treated in current PRA models. Mitigation measures treated in SOARCA 

include the licensee’s EOPs, SAMGs, and 10 CFR 50.54(hh) mitigation measures. It is expected 

that the licensee’s emergency response organization would implement these measures in 

accordance with the approved emergency plan.

3.1

In preparation for the detailed, realistic modeling of accident progression and offsite 

consequences, the SOARCA project staff had extensive cooperation from the licensees to 

develop high-fidelity plant systems models; define operator actions, including the most recently 

developed mitigative actions; and develop models to simulate site-specific and scenario-specific 

emergency planning. In addition to input for model development, licensees provided 

information from their own PRAs on accident scenarios. Through tabletop exercises (with senior 

reactor operators, PRA analysts, and other licensee staff) of the selected scenarios, licensees 

provided input on the timing and nature of the operator actions to mitigate the selected scenarios. 

The licensee input for each scenario was used to develop timelines of operator actions and 

equipment lineup or setup times for implementing the available mitigation measures. This 

includes mitigation measures beyond those treated in current PRA models. 

Site-Specific Mitigation Strategies

The SOARCA team developed the timelines for implementing the mitigation measures directed 

in plant-specific procedures and mobilizing support organizations after discussing each scenario 

with licensee personnel who have experience in operations, engineering, and facility 

management. The team developed these timelines through multiple site visits and system 

walkdowns in 2007, 2010, and 2011, with licensee personnel specifically reviewing the steps to 

implement mitigation. Results of preliminary accident progression calculations were used to 

characterize anticipated changes in plant conditions and describe the signatures of measurable 

parameters. Estimates were then made for the time needed to assemble necessary personnel, 

tools, and equipment; align and start components; and establish a desired operating condition. 

For the ISLOCA scenario, where the timing of operator actions was judged to be important to the 

results, the licensee performed plant simulator runs with reactor operators to ensure that the 

timing for key actions was as realistic as possible. 

Mitigation measures treated in SOARCA include EOPs, SAMGs, and 10 CFR 50.54(hh)

mitigation measures. The 10 CFR 50.54(hh) mitigation measures refer to additional equipment 

and strategies required by the NRC following the terrorist attacks of September 11, 2001, to 
28

further improve severe accident mitigation capability. NRC inspectors completed the 

verification of licensee implementation (i.e., equipment, procedures, and training) of 

10 CFR 50.54(hh) mitigation measures in December 2008. These mitigation measures are for 

use during scenarios involving large fires and explosions. One such measure is portable, 

self-powered equipment, including generators and diesel-driven pumps. Portable generators 

provide electrical power to equipment that gives critical indications, such as the reactor vessel 

water level. Portable generators also provide electrical power needed to operate safety relief 

valves. Portable diesel-driven pumps provide a diverse and independent means of injecting 

water into the RCS and steam generators. Another such measure is starting and controlling, 

without electrical control power, the plant’s existing turbine-driven injection systems, including 

the RCIC and TDAFW systems.

To quantify the benefits of the mitigation measures and to provide a basis for comparison to past 

analyses of unmitigated severe accident scenarios, the project team also analyzed the scenarios 

assuming that the events proceed as unmitigated by key available onsite mitigation measures,

ultimately leading to core damage and an offsite release. This NUREG refers to these as 

“unmitigated scenarios,” because they are not effectively mitigated by onsite resources. This 

report does not determine the respective likelihoods of the mitigated and unmitigated cases of 

each scenario. 

3.1.1 Scenarios Initiated by External Events

Scenarios identified in SOARCA included both externally and internally initiated events. The 

externally initiated events included events for which seismic, fire, extreme wind, and flooding 

initiators were grouped together. 

The PRA screening identified the following scenarios that were initiated by external seismic, 

fire, or flooding events: 

• Peach Bottom LTSBO: 1×10-6 to 5×10-6

/reactor-year

• Surry LTSBO: 1×10-5 to 2×10-5/reactor-year

• Surry STSBO: 1×10-6 to 2×10-6/reactor-year

• Surry STSBO with TISGTR: 3×10-7 to 5×10-7

/reactor-year

The mitigation measures assessment for each of these scenarios assumed that the initiator was a 

seismic event, because it was judged to be limiting. Seismic initiators are considered to be 

limiting for two principal reasons. First, seismic initiators are more likely to result in the near 

immediate failure of systems, whereas, fire and flood would be expected to result in delayed 

failures. Secondly, a seismic event may be more likely than a fire or flood to fail passive 

components, such as water tanks. Additionally, seismic initiators may be more likely to have 

sitewide and offsite impacts. 

It is important to note that, although it is not included in the above list, the seismically induced 

Peach Bottom STSBO was also retained for analysis. With a frequency of 1×10-7 to 

5×10-7

/reactor-year, this scenario does not explicitly meet the SOARCA screening criterion. 

Nonetheless, it was retained to assess the risk importance of a lower frequency, potentially 
29

higher consequence scenario. The STSBO has also been an important event in many past PRAs 

and is limiting in many transients. 

Seismic events considered in SOARCA result in loss of offsite and onsite ac power and, for the 

more severe seismic events, loss of dc power. Under these conditions, the turbine-driven RCIC 

and TDAFW systems are important mitigation measures. BWR SAMGs include starting RCIC 

without electricity to cope with SBO conditions. This is known as RCIC blackstart. The 

10 CFR 50.54(hh) mitigation measures have taken this a step further and also include long-term 

operation of RCIC without electricity (RCIC blackrun), using a portable generator to supply 

power to indications, such as the RPV level indication, to allow the operator to manually adjust 

RCIC flow to prevent RPV overfill and flooding of the RCIC turbine. Similar procedures have 

been developed for PWRs for TDAFW. For the Peach Bottom and Surry LTSBO scenarios, 

RCIC and TDAFW can be used to cool the core until battery exhaustion. In addition, blackstart 

procedures can be used for the Peach Bottom STSBO scenario. After battery exhaustion, 

blackrun of RCIC and TDAFW systems can continue to cool the core. The study used 

MELCOR calculations to demonstrate core cooling under these conditions. 

Seismic PRAs for Peach Bottom and Surry do not describe general plant damage and 

accessibility. The damage was assumed to be widespread and accessibility to be difficult, 

consistent with the unavailability of many plant systems. 

The seismic initiating event for the SBO accident scenarios might rupture the CST, which is the 

primary water reservoir for RCIC. However, the Peach Bottom CST is surrounded by a 

reinforced concrete dike or moat, which would retain water drained from the CST. Therefore, 

suction from the CST would not be interrupted by a loss of CST integrity. Makeup to the CST 

would likely be available from the cooling water tower basin (3.55 million gallons or 13,438 

cubic meters), or the Susquehanna River; the diesel-driven portable pump 

(i.e., 10 CFR 50.54(hh) equipment) or other mobile equipment could be used. 

For the Surry LTSBO, the TDAFW pump is available until the ECST empties. The ECST 

initially supplies the TDAFW pump but has finite resources (i.e., it empties in 5 hours). 

However, the team estimated that the operators would have sufficient time, access, and resources 

to make up water for injection into the ECST. The low-pressure injection and safety-related 

containment spray piping were judged not likely to fail for this scenario. The integrity of this 

piping provided a connection point for a portable, diesel-driven pump to inject into the RCS. 

Licensee staff estimated that transporting the pump and connecting it to plant piping would take 

about 2 hours, leading to vessel injection at 3.5 hours, or 2 hours after the operators and support 

staff recommended the action. Consequently, the cooling water would be supplied to the steam 

generators for RCS heat removal. The team assumed that operators would eventually provide 

makeup water to the ECST. 

One Surry STSBO assumption was that the ECST would fail and an alternative reservoir would 

be available within 8 hours; using a fire truck or portable pump to draw from the discharge canal. 

The low-pressure injection and containment spray safety-related piping were judged not likely to 

fail, based primarily on NUREG/CR-4334, “An Approach to the Quantification of Seismic 

Margins in Nuclear Power Plants,” issued in 1985 [11], to help extrapolate the potential viability 
30

of safety-related piping after a 1.0 g event. This conclusion also considered related studies,

including a 2007 German study, “Seismic PSA of the Neckarwestheim 1 Nuclear Power 

Plant” [12], that physically simulated ground motion equal to 1 g on an existing plant. The 

integrity of this piping provided a connection point for a portable, diesel-driven pump to inject 

into the RCS or into the containment spray systems. Licensee staff estimated that transporting 

the pump and connecting it to plant piping would take about 2 hours. However, for the STSBO, 

this mitigation measure was conservatively estimated to take 8 hours, owing to the higher level 

of damage. Because the installation time was beyond the estimated time to fuel damage and 

vessel failure (3 hours to core damage, 7 hours to lower head failure), the containment spray 

system was the preferred mitigation measure. A better understanding of the effect of large 

seismic events on general plant conditions would be helpful in reducing the uncertainty in 

availability and accessibility for mitigation measures. If accessibility was not significantly 

impaired and delay in using the portable pump was limited to 2 hours, then core damage could be 

averted.

The 10 CFR 50.54(hh) mitigation measures include portable equipment (such as portable power 

supplies to supply indication, portable diesel-driven pumps, and portable air bottles to open 

air-operated valves), together with procedures to implement these measures under severe 

accident conditions. Surry’s portable equipment and fire truck are stored onsite in a one-story, 

multibay garage. Some of Peach Bottom’s portable equipment is stored in an open bay in the 

water treatment building and some is stored outside under a tarp. The mitigated cases assumed 

that this equipment survived the seismic event and could be successfully implemented.

The SOARCA team estimated the time to implement individual mitigation measures based on 

licensee input for each scenario; these estimates take into account the plant conditions following 

the seismic event. Also, for portable equipment, the time estimates reflect exercises run by 

licensee staff that provided actual times to move the equipment into place and were adjusted 

(increased) to account for the larger seismic event. The time estimates for staffing the TSCs and 

the emergency operating facilities (EOFs) were based on regulatory requirements and the 

potential for additional delays resulting from the possible effect of the seismic event on roads 

and bridges.

The mitigation measures assessment noted the possibility of bringing in offsite equipment 

(e.g., fire trucks, pumps, and power supplies from sister plants or from contractors), but it did not 

quantify the types, amounts, and timing of this equipment arriving and being implemented. This 

equipment is also judged to be effective in mitigating an environmental release (by flooding core 

debris) after it begins. Section 3.2 provides additional information on equipment available 

offsite and time estimates for transporting this equipment.

Because the SOARCA project did not analyze multiunit accidents, the mitigation measures

assessment for external events assumed that the operators only had to mitigate an accident at one 

reactor, even though Peach Bottom and Surry are two-unit sites. It also assumed minimum 

staffing and that half of the onsite operators mitigate the damaged unit. Peach Bottom had 

voluntarily arranged to provide redundant 10 CFR 50.54(hh) equipment to mitigate both units

simultaneously; however, SOARCA did not examine this. 
31

3.1.2 Scenarios Initiated by Internal Events

The PRA screening identified the following scenarios that were initiated by internal events:

• Surry interfacing systems loss-of-cooling accident (ISLOCA): 3×10-8/reactor-year 

• Surry spontaneous SGTR: 5×10-7

/reactor-year

These scenarios result in core damage as a result of assumed operator errors. For the ISLOCA, 

the operators fail to refill the RWST or cross-connect to the other unit’s RWST. For the 

spontaneous SGTR, the operators fail to (1) isolate the faulted steam generator, (2) depressurize 

and cool down the RCS, and (3) refill the RWST or cross-connect to the unaffected unit’s 

RWST.

The SPAR model and the licensee’s PRA concluded that these two events proceed to core 

damage as a result of the above-postulated operator errors. However, these PRA models do not 

appear to have credited the significant time available for the operators to correctly respond to 

events. They also do not appear to credit technical assistance from the TSC and the EOF. For 

the ISLOCA, the realistic analysis of thermal-hydraulics presented in NUREG/CR-7110, 

Volume 2, estimated 6 hours until the RWST is empty and 13 hours until fission product release 

begins, providing time for the operators to correctly respond. The ISLOCA time estimates are 

based on a double-ended pipe rupture. These estimates would be longer for smaller break sizes. 

Also, if the operators throttle high-head safety injection to match decay heat, the time to empty 

the RWST and the beginning of core damage would be extended by an additional 24 hours. For 

the SGTR, the realistic analysis of thermal-hydraulics showed from 24 to 48 hours until core 

damage begins. Therefore, based on realistic time estimates by which the technical assistance is 

received from the TSC and the EOF, it was highly likely that the operators would correctly 

respond to the events. These time estimates considered indications that the operators would have 

of the bypass accident, operator training on plant procedures for dealing with bypass accidents 

and related drills, and assistance from the TSC and EOF, which were estimated to be fully 

staffed and operational by 1 to 1.5 hours into the event.

The mitigation measures assessment for internal events also included 10 CFR 50.54(hh) 

mitigation measures, but these were subsequently shown to be redundant to the wide variety of 

equipment and indications available for mitigating the ISLOCA and SGTR. ISLOCA and SGTR 

are internal events that involve few equipment failures and are controlled by operator errors.

The PRA screening for Peach Bottom initially identified the Loss of Vital ac Bus E12 scenario as 

exceeding the SOARCA screening criterion of 1×10-6

/reactor-year. However, a simplifying 

modeling assumption was subsequently found in the SPAR model, and the scenario frequency 

was determined to be below the SOARCA screening criterion. By the time the issue was 

discovered, the mitigation measures assessment and the MELCOR analysis were complete. The 

MELCOR analysis described in NUREG/CR-7110, Volume 1, demonstrated that this scenario 

did not result in core damage, even without crediting 10 CFR 50.54(hh) mitigation measures, 

contrary to the more conservative treatment in SPAR. Nevertheless, this report describes the 
32

mitigation measures assessment and the MELCOR analysis for this scenario to demonstrate the 

benefit of a detailed review of success criteria using integrated thermal-hydraulic analysis.

3.2

Many resources at the State, regional, and national level would be available to mitigate an NPP 

accident. The staff reviewed available resources and emergency plans and determined that 

adequate mitigation measures could be brought onsite within 24 hours and be connected and 

functioning within 48 hours.

Truncation of Releases

Concurrent with the NRC and industry response, the National Response Framework (NRF) 

would establish a coordinated response of national assets. As described in the 

Nuclear/Radiological Incident Annex to the NRF, the NRC is typically the Coordinating Agency 

for incidents occurring at NRC-licensed facilities. As Coordinating Agency, the NRC has 

technical leadership for the Federal Government’s response to the incident. Under an established 

agreement with the NRC, the U.S. Department of Homeland Security would be the Coordinating 

Agency for an event in which a general emergency is declared. The NRF is exercised 

periodically and provides access to the full resources of the Federal Government. The NRC has 

an extensive, well-trained, and exercised emergency response capability and has onsite resident 

inspectors. These onsite inspectors are equipped and available to provide firsthand knowledge of 

accident conditions. The NRC would activate the incident response team at the NRC regional 

office and Headquarters. The focus of the NRC response is to ensure that public health and 

safety are protected and to assist the licensee with the response by working with the Department 

of Homeland Security to coordinate the national response. Concurrently, the NRC regional 

office would send a site team to staff positions in the reactor control room, TSC, and EOF to 

support the response. The EOF and TSC are assumed manned and operational in roughly 1–

2 hours, depending on the accident scenario. The NRC performs an independent assessment of 

the actions taken or proposed by the licensee to confirm that such actions will arrest the accident.

Both Surry and Peach Bottom are supported by an offsite EOF. The emergency response 

organization at the EOF has access to fleetwide emergency response personnel and equipment, 

including the 10 CFR 50.54(hh) mitigation measures and equipment from sister plants. These 

assets, as well as those from neighboring utilities and State preparedness programs, could be 

brought to bear on the accident if needed. Every licensee participates in full onsite and offsite 

exercises every 2 years where response to severe accidents and coordination with offsite

response organizations (OROs) is demonstrated and inspected by the NRC and the Federal 

Emergency Management Agency. In addition, the Institute for Nuclear Power Operations and 

the Nuclear Energy Institute would activate their emergency response centers to assist the site as 

needed. 

All of the described resources would be available to the site to mitigate the accident. Although 

some of these efforts would be ad hoc, knowledgeable personnel and an extensive array of 

equipment would be available and were considered in the conclusion that radiological releases 

would be truncated within 48 hours except for the Surry LTSBO sequence, which was truncated 

at 72 hours. 
33

4.0 SOURCE TERM ANALYSIS

Section 4.1 describes some background in key studies for regulatory and probabilistic 

applications. Figure 3 shows a timeline of key events and NRC studies in the evolution of 

nuclear safety technology, as well as the key source term studies cited in the timeline that 

preceded the SOARCA program (also discussed in Section 4.1 below). Section 4.2 contains a 

history of the severe accident source term codes developed by the NRC and the scope of the 

MELCOR code. The MELCOR code is the culmination of the NRC research and code 

development of severe accident phenomena for source term evaluations. Section 4.3 presents the 

MELCOR modeling approach used in the SOARCA analyses. This includes the development of 

the plant models, the best practices approaches to important but uncertain phenomena and 

equipment performance, recent advances in source term models, and the methods used to 

calculate the radionuclide inventories.

Figure 3 Timeline of key nuclear power events and safety studies

4.1

The Reactor Safety Study (WASH-1400)

Source Term Study Background

[3], was the first systematic attempt to provide realistic 

estimates of public risk from potential accidents in commercial NPPs. The 1975 study included 

analytical methods for determining both the probabilities and consequences of various accident 

scenarios. The study used event trees and fault trees to define important accident sequences and 

to quantify the reliability of engineered safety systems and contained a list of nine PWR and five 

BWR source terms. All the accidents that were believed to contribute significantly to the overall 

core melt frequency were grouped, or “binned,” into the source term categories. The 
34

WASH-1400 source terms included characterizations of accident timing, the release duration 

(e.g., puff or sustained release), and the energy of the release for plume loft considerations. The 

description of radioactivity used eight chemical categories. The 54 most health-significant 

isotopes were used in health consequence calculations. 

The WASH-1400 methodology used to predict the health effects from the source term was based 

on the newly developed Calculation of Reactor Accident Consequences (CRAC) code [18] that 

calculated the atmospheric dispersion and health consequences. However, an integrated tool for 

the calculation of the source term did not exist. The estimation of the source term used the best 

analytic procedures available at the time. When ample data were available, a model for the 

phenomenon was included as realistically as possible, but when data were lacking, consideration 

of the phenomenon was omitted. The resultant source terms reflected uncertainties and poor 

understanding of applicable phenomena. Uncertainties in accident frequencies were accounted 

for by adding 10 percent of the likelihood of each release category into the next larger and the 

next smaller category.

Subsequently, the NRC documented the technical basis for source terms in NUREG-0772, 

“Technical Bases for Estimating Fission Product Behavior during LWR Accidents,” issued 

June 1981 [19]. NUREG-0772 assessed the assumptions, procedures, and available data for 

predicting fission product behavior. Four conclusions of the NUREG-0772 study were (1) a new 

definition of the chemical form of iodine (I) (i.e., cesium iodide (CsI) was the dominant form), 

(2) the potential retention of CsI within the vessel or containment versus elemental iodine, (3) the 

inclusion of in-vessel retention, and (4) the role of containment engineering safety features 

(e.g., sprays, suppression pools, and ice condensers). However, NUREG-0772 based much of 

the quantitative assessment on scoping calculations that were applicable only to specific 

conditions. In particular, it conducted the examination of fission product behavior in different 

regions of the plant with different accidents in parallel with limited consideration of integral 

effects. The NRC examined the potential impact of the NUREG-0772 findings on reactor 

regulation and documented the results in draft NUREG-0771, “Regulatory Impact of Nuclear 

Reactor Accident Source Term Assumptions,” issued June 1981 [20].

The NUREG-0771 and NUREG-0772 studies formed the basis for the designation of five 

accident groups as representative of the spectrum of potential accident conditions documented in 

NUREG-0773, “The Development of Severe Accident Source Terms: 1957–1981,” issued 

November 1982 [21]. In 1982, the NRC issued the NUREG/CR-2239 siting study [1] using the 

NUREG-0773 source terms. It determined that the five source terms adequately spanned the 

range of possible source terms. The source terms, developed from separate effects computer 

code analyses that were performed in 1978, were used to calculate accident consequences at 

91 U.S. reactor sites using site-specific population data and a mixture of site-specific and 

regionally specific meteorological data. An objective of the SOARCA study is to update this 

study.

In response to emerging regulatory needs, Battelle Columbus Laboratories conducted a study, 

“Radionuclide Release Under Specific LWR Accident Conditions,” published in 1985 [22], that 

developed and modified a number of separate effects severe accident computer codes based on 

emerging severe accident research. The codes, coupled together to form a code suite, could 
35

calculate a complete accident sequence. The new Source Term Code Package (STCP) code [22]

calculated the source terms for about 25 specific sequences for five operating plants. Although 

the STCP was a significant step forward in deterministic severe accident analysis, the code suite 

had some significant shortcomings. Because the code represented the linkage of many separate 

code modules, the data transfer and feedback effects were not always handled consistently. The 

technical basis for the models in the STCP is in NUREG-0956, “Reassessment of the Technical 

Bases for Estimating Source Terms,” issued July 1986 [16]. The results from the STCP 

calculations supported the NUREG-1150 PRA [2], along with expert judgment and simplified 

algorithms for sequence-specific source terms.

The NUREG-1150 PRA was an 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 CDF and 

source term caused by an incomplete understanding of reactor systems and severe accident 

phenomena at that time. NUREG-1150 therefore used 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, structural response, and source term issues. 

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. 

4.2

The MELCOR code, a fully integrated, engineering-level computer code, has, as its primary 

purpose, to model the progression of accidents in LWR NPPs, as well as in nonreactor systems 

(e.g., spent fuel pool, dry cask). Current uses of MELCOR include estimation of fission product 

source terms and their sensitivities and uncertainties in a variety of applications. MELCOR is a 

modular code comprising three general types of packages: (1) basic physical phenomena 

(i.e., hydrodynamics, heat and mass transfer to structures, gas combustion, aerosol and vapor 

physics), (2) reactor-specific phenomena (i.e., decay heat generation, core degradation, ex-vessel 

phenomena, sprays, and engineering safety systems), and (3) support functions 

(thermodynamics, equations of state, other material properties, data-handling utilities, and 

equation solvers). As a fully integrated code, MELCOR models all major systems of a reactor 

plant and their important coupled interactions. 

The MELCOR Code

Figure 4 shows the MELCOR code integration of models for important phenomena previously 

treated in separate effects codes. 
36

Figure 4 MELCOR integration of separate effects codes

The scope of MELCOR includes the following:

• thermal-hydraulic response of the RCS, reactor cavity, containment, and confinement 

buildings

• core uncovery (loss of coolant), fuel heatup, cladding oxidation, fuel degradation (loss of 

rod geometry), and core material melting and relocation

• heatup of reactor vessel lower head from relocated core materials and the thermal and 

mechanical loading and failure of the vessel lower head and transfer of core materials to 

the reactor vessel cavity

• core-concrete attack and ensuing aerosol generation

• in-vessel and ex-vessel hydrogen production, transport, and combustion

• fission product release (aerosol and vapor), transport, and deposition
37

• behavior of radioactive aerosols in the reactor containment building, including scrubbing 

in water pools and aerosol mechanics in the containment atmosphere, such as particle 

agglomeration and gravitational settling

• the impact of engineered safety features on thermal-hydraulic and radionuclide behavior

Most MELCOR models are mechanistic, and the use of parametric models is limited to areas of 

high phenomenological uncertainty where no consensus exists concerning an acceptable 

mechanistic approach. Current use of MELCOR often includes uncertainty analyses and 

sensitivity studies. To facilitate this, many of the mechanistic models have been coded with 

optional adjustable parameters. This does not affect the mechanistic nature of the modeling, but 

it does allow the analyst to easily address questions of how particular modeling parameters affect 

the course of a calculated transient. MELCOR does not use core radioactive nuclide inventories;

rather, it uses masses and decay heats of chemical element groups. Appropriate code 

calculations for specific fuel and core design are carried out to the burnup of interest to provide 

the initial core inventories for MELCOR severe accident analysis (see Section 4.3.1).

After the completion of Version 1.8.1 in 1991, the NRC commissioned a peer review using 

recognized experts from national laboratories, universities, and the MELCOR user community 

[61]. The charter of the MELCOR peer review committee was to (1) provide an independent 

assessment of the MELCOR code through a peer review process, (2) determine the technical 

adequacy of the MELCOR code for the complex analyses it is expected to perform, and (3) issue 

a final report describing its technical findings. The committee offered a set of major findings 

that covered the various physics model numerics, missing models, modeling deficiencies, code 

assessment, and documentation. The NRC incorporated the findings into the research plan that 

governed the subsequent code development.

In 2000, the NRC began reducing the number of codes that it actively maintained by 

consolidating the CONTAIN, SCDAP/RELAP5, and VICTORIA code functionality and models 

into MELCOR. The assessment of MELCOR parity with CONTAIN showed that MELCOR 

results are comparable to CONTAIN. A comprehensive parity study of the MELCOR code with 

SCDAP/RELAP5 is ongoing. The assessment of fission product chemistry and transport is 

currently supported by foreign experiments (especially those from the Phebus facility in France). 

Hence, the scope of the evaluation of parity of the MELCOR to the VICTORIA code not only 

includes the phenomena treated in VICTORIA but also new experimental findings.

4.3

Section 4.3.1 presents a high-level description of MELCOR models used for the SOARCA 

project. Existing MELCOR models for Surry and Peach Bottom were updated to current 

state-of-the-art modeling practices, as well as the latest version of the MELCOR code 

(Version 1.8.6). More detailed information describing the plant models is in the plant-specific 

analysis reports (i.e., NUREG/CR-7110, Volumes 1 and 2, for Peach Bottom and Surry, 

respectively). 

MELCOR Modeling Approach 

The modeling and prediction of accident progression and radiological release in a severe 

accident requires the integration of a number of phenomenological models to address a range of 
38

thermal-hydraulic, materials, structural, and fission product behavior, as well as models for 

component (e.g., safety relief valve) behavior. Section 4.3.2 describes the procedure to define 

the best practices approach to modeling important and uncertain phenomena. NUREG/CR-7008, 

“Best Practices for Simulation of Severe Accident Progression at Nuclear Power Plants” [6], 

provides a more detailed description of the best practices modeling approach. At the beginning 

of the SOARCA project, an independent review of MELCOR best practices modeling provided 

greater assurance of the technical soundness of the analytical modeling [42]. The NRC used that 

review to identify and incorporate subsequent modeling insights and improvements before the 

start of plant analyses. Moreover, members of the SOARCA peer review committee 

recommended additional sensitivity analyses to explore specific modeling issues that were 

viewed as both uncertain and potentially important to risk. These analyses, discussed in detail in 

NUREG/CR-7110, Volumes 1 and 2, help confirm that the modeling of MELCOR best practices 

is sound.

Section 4.3.3 summarizes some recent changes to the modeling of radionuclide release and 

cesium speciation, which is important to the source term results. Finally, Section 4.3.4 describes 

the methodology for calculating the radionuclide inventory.

4.3.1 Plant Models

The SOARCA program updated the MELCOR models for Peach Bottom and Surry to the most 

recent version of the MELCOR code.3

 The scope of the models included the following:

• detailed five-ring reactor vessel models

• representation of the RCS (and secondary system through the main steam isolation valve 

for Surry)

• representation of the primary containment

• representation of the Peach Bottom reactor building and the Surry Safeguards and 

Auxiliary Buildings, and ventilation and filter systems, which were radionuclide 

pathways in the ISLOCA scenario

• representation of the emergency core cooling systems (and the auxiliary feedwater 

system for Surry)

• representations of the emergency portable water-injection systems

 3 All SOARCA calculations used MELCOR Version 1.8.6. 
39

The best practices updates to each input deck specified the following new models for both plants 

for these important but uncertain phenomena or equipment responses:

• Safety relief valve failure modeling addressing stochastic and high-temperature failure 

modes.

• An additional thermomechanical fuel collapse model for heavily oxidized fuel following 

molten Zircaloy breakout.

• Enhanced lower plenum coolant debris heat transfer that recognizes breakup and 

multidimensional cooling effects not present in the one-dimensional countercurrent 

flooding model in older versions of MELCOR (e.g., [23]).

• Updated, plant-specific chemical element masses and decay heats (see Section 4.3.4).

• A new Oak Ridge National Laboratory (ORNL) Booth chemical element release model 

and new cesium speciation model (see Section 4.3.3).

• A new turbulent deposition model for aerosol deposition in piping systems. 

NUREG/CR-7110, Volume 2 discusses this new model and its validation. 

• Vessel failure based on gross failure [24] using the improved one-dimensional creep 

rupture model with the new hemispherical head model and radial heat transfer between 

lower head conduction node segments. A more complete discussion of this model is 

presented in NUREG/CR-7008 [6] and the MELCOR manual [31]. A penetration failure 

model was not used because the timing differences between gross lower head failure and 

penetration failure with the available penetration model are not significant to the overall 

accident progression (i.e., minutes difference). Also, Sandia lower head failure tests 

showed that gross creep rupture of the lower head was measured to be the most likely 

mechanism for vessel failure [24].

• Enhanced ex-vessel core debris heat transfer that reflects multidimensional effects and 

rates measured in MACE tests [25].

Sections 4.3.1.1 and 4.3.1.2 summarize the SOARCA program’s recent enhancements to the 

MELCOR Peach Bottom and Surry models, respectively. 

4.3.1.1 Peach Bottom MELCOR Model

Brookhaven National Laboratory originally developed the Peach Bottom MELCOR plant model 

for code Version 1.8.0. J. Carbajo at ORNL subsequently adopted the model to study differences 

in fission product source term behavior predicted by MELCOR 1.8.1 and those generated for use 

in NUREG-1150 [2] using the STCP [26]. Starting in 2001, SNL has made considerable 

refinements to the BWR/4 core nodalization to support the developmental assessment and release 

of MELCOR 1.8.5. These refinements concentrated on the spatial nodalization of the reactor 
40

core (both in terms of fuel and structural material and hydrodynamic volumes) used to calculate 

in-vessel melt progression.

Subsequent work in support of several NRC research programs has motivated further refinement 

and expansion of the BWR/4 model in four broad areas. The first area involved the addition of 

models to represent a wide spectrum of plant design features, such as safety systems, to broaden 

the capabilities of MELCOR simulations to apply to a wider range of severe accident sequences. 

These enhancements include the following:

• modifications of modeling features needed to achieve steady-state reactor conditions 

(recirculation loops, jet pumps, steam separators, steam dryers, feedwater flow, control 

rod drive hydraulic system, main steamlines, turbine/hotwell, core power profile) 

• new models and control logic to represent coolant injection systems (RCIC, HPCI, 

residual heat removal, low-pressure core spray) and supporting water resources 

(e.g., CST with switchover)

• new models to simulate reactor vessel pressure management (safety relief valves, safety 

valves, automatic depressurization system, and logic for manual actions to effect a 

controlled depressurization if torus water temperatures exceed the heat capacity 

temperature limit)

The second area focused on the spatial representation of the containment and the reactor 

building. The drywell portion of containment has been subdivided to distinguish thermodynamic 

conditions internal to the pedestal from those within the drywell itself. Also, refinements have 

been added to the spatial representation and flow paths within the reactor building. A 

containment failure model is included that accounts for leakage around the drywell head flange, 

leakage caused by elevated drywell temperature, and leakage caused by drywell melt-through 

(see NUREG/CR-7110, Volume 1, Section 4.6). The third area focused on bringing the model 

up to current “best practice” standards for MELCOR 1.8.6 (see Section 4.3.2). The fourth area 

of model improvements included a new radionuclide inventory and decay heat based on the 

recent plant operating history (see Section 4.3.4).

Although not new for SOARCA, the MELCOR Peach Bottom model includes a multiregion 

ex-vessel debris spreading model. The debris spreads according to its temperature relative to the 

solidus and liquidus temperatures of the concrete and the debris height. If the debris spreads 

against the drywell liner steel wall, and if the debris temperature is above the carbon steel 

melting temperature, the liner will fail.

The potential for creep rupture of a BWR main steam line (i.e., piping or RPV nozzle) was added to 

the Peach Bottom model developed for SOARCA. 

NUREG/CR-7110, Volume 1, more fully describes the MELCOR Peach Bottom model. Figure 

5 shows the MELCOR nodalization diagrams for Peach Bottom. 
41

Figure 5 The Peach Bottom MELCOR nodalization
42

4.3.1.2 Surry MELCOR Model

In 1988, Idaho National Engineering Laboratory originally generated the Surry MELCOR model 

applied in this study. SNL periodically updated it (1990 to present) to test new models, 

advancing the state of the art in modeling PWR accident progression and providing support to 

the NRC for analyses of various issues that could affect operational safety. Significant changes 

were made during the last several decades in the approach to modeling core behavior and core 

melt progression, as well as the nodalization and treatment of coolant flow within the RCS and 

reactor vessel. In 2002, the reactor vessel and RCS nodalization were updated using the 

SCDAP/RELAP5 Surry model to include a five-ring vessel nodalization and countercurrent 

hot-leg representation for natural circulation flow [27]. The current MELCOR Surry model is a 

culmination of these efforts. 

In preparation for the SOARCA analyses described in this report, the model was further refined 

and expanded in three areas. The first area is an upgrade to core modeling in MELCOR 

Version 1.8.6. These enhancements include the following:

• a hemispherical lower head model that replaces the flat-bottom cylindrical lower head 

model

• new models for the core former and shroud structures that are fully integrated into the 

material degradation modeling, including separate modeling of debris in the bypass 

region between the core barrel and the core shroud

• models for simulating the formation of molten pools in both the core and lower plenum, 

crust formation, convection in molten pools, stratification of molten pools into metallic 

and oxide layers, and partitioning of radionuclides between stratified molten pools

• a reflood quench model that separately tracks the component quench front and the 

quenched and unquenched temperatures

• a control rod aerosol release model

• addition of the new ORNL Booth radionuclide release model for modern high-burnup 

fuel

The second area focused on the addition of user-specified models to represent a wide spectrum 

of plant design features and safety systems to broaden the ability of MELCOR to apply to a 

wider range of severe accident sequences. These enhancements included the following:

• update of the containment leakage model to include nominal leakage and leakage caused 

by containment overpressure (see NUREG/CR-7110, Volume 2, Section 4.8)

• update of core degradation modeling practices
43

• models of the individual primary system and secondary system relief valves with failure 

logic for rated and degraded conditions

• update of the containment flooding characteristics

• heat loss from the reactor to the containment

• separate motor driven auxiliary feed water and TDAFW models with control logic for 

plant automatic and operator cooldown responses

• new TDAFW models for steam flow, flooding failure, and performance degradation at 

low pressure

• nitrogen discharge model for accumulators

• update of the fission product inventory, the axial and radial peaking factors, and an 

extensive fission product tracking control system

• improvements to the modeling of natural circulation in the hot leg and steam generator 

and the potential for creep rupture

NUREG/CR-7110, Volume 2, more fully describes the MELCOR Surry model. Figure 6 shows 

the MELCOR nodalization diagrams for Surry.
44

Figure 6 The Surry MELCOR nodalization
45

4.3.2 Best Modeling Practices

The SOARCA project’s integrated modeling of the accident progression and offsite 

consequences uses both state-of-the-art computational analysis tools and best modeling practices 

drawn from the collective body of knowledge on severe accident behavior generated over the 

past several decades of research. 

The MELCOR 1.8.6 computer code embodies much of this knowledge and was used for the 

accident and source term analysis. MELCOR includes capabilities to model the two-phase 

thermal-hydraulics, core degradation, fission product release, transport, deposition, and 

containment response. The SOARCA analyses include operator actions and equipment 

performance issues as prescribed by the sequence definition and mitigative actions. The 

MELCOR models are constructed using plant data, and the operator actions were developed 

based on tabletop exercises during site visits. The code models and user-specified modeling 

practices represent the current best practices.

While much has been learned through extensive research, uncertainties exist in understanding 

phenomena associated with severe accident progression and radionuclide transport. Consistent 

with the stated objective of SOARCA, phenomena were modeled using realistic characterization 

of phenomena and events. The accident progression analysts developed a list of key uncertain 

phenomena that can have a significant effect on the progression of the accident. Plant-specific 

reports for Peach Bottom (NUREG/CR-7110, Volume 1) and Surry (NUREG/CR-7110, 

Volume 2) outlined each issue and identified a modeling approach or base case values. 

NUREG/CR-7008 [6] discusses the specific modeling practices.

The SOARCA project excluded several early containment failure modes of historical interest 

because of their assessed low likelihood of occurrence. These include the following:

• Alpha mode containment failure

[28]

would be caused by an in-vessel steam explosion during 

melt relocation that simultaneously fails the vessel and the containment. A group of 

experts in this field, referred to as the Steam Explosion Review Group, concluded, in a 

position paper published by the Nuclear Energy Agency Committee on the Safety of 

Nuclear Installations , that the alpha-mode failure issue for Western-style reactor 

containment buildings can be considered resolved from a risk perspective, having little or 

no significance to the overall risk from an NPP.

• Direct containment heating

[29]

would be caused by containment failure in PWR 

containments. NRC research has shown that an early failure of the PWR RCS caused by

high-temperature natural circulation will likely depressurize the RCS before vessel 

failure. Importantly, extensive NRC testing and analyses have also shown that, in the 

unlikely event of a high-pressure vessel failure, early containment failure caused by

direct containment heating is very unlikely, with some variation depending on plant 

design . In the case of Surry, the research concluded that no feasible likelihood exists 

of failing the containment.
46

• Early containment failure

[30]

would be caused by drywell liner melt-through in a wet cavity 

in Mark I containments (e.g., Peach Bottom). Through a detailed assessment of the issue, 

the research 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 .

At the start of the SOARCA project, a panel of experts reviewed the proposed modeling 

approach for SOARCA analyses during a public meeting sponsored by the NRC on August 21–

22, 2006, in Albuquerque, NM. The panel examined the best modeling practices for the 

application of MELCOR to realistically evaluate accident progression and source term. The 

panel also reviewed a set of code enhancements and considered the SOARCA project in general.

4.3.3 Radionuclide Modeling

The radionuclide modeling was updated in the Peach Bottom and Surry models to apply a more 

mechanistic radionuclide release model (i.e., the ORNL-Booth model [31]) based on assessments 

of recent radionuclide release tests. These assessments identified an alternative set of Booth 

diffusion parameters recommended by ORNL (ORNL-Booth) [32] that produced significantly 

improved release signatures for cesium and other fission product groups. Some adjustments to 

the scaling factors in the ORNL-Booth model were made for selected fission product groups 

including uranium dioxide (UO2), molybdenum (Mo), and ruthenium (Ru) to gain better 

comparisons with FPT-1 data [33]. The adjusted model, referred to as “Modified ORNL-Booth,” 

was subsequently compared to original ORNL VI fission product release experiments and to 

more recently performed VERCORS tests [34], and the comparisons were as favorable or better 

than the original CORSOR-M MELCOR default release model. These modified ORNL-Booth 

parameters were introduced into the MELCOR code as new defaults for the SOARCA project. 

Although the analysis of the FPT-1 test with the ORNL-Booth parameters obtained significant 

improvements in release behavior, some additional modification to the MELCOR release model 

was pursued. Evidence from the Phebus experiments increasingly indicates that the dominant 

chemical form of released Cs is that of Cs2MoO4. This is based on deposition patterns in the 

Phebus experiment, where Cs is judged to be in aerosol form at 700 degrees Celsius, which 

explains deposits in the hot upper plenum of the Phebus test section and deposition patterns in 

the cooler steam generator tubes. In recognition of response, a Cs2MoO4 radionuclide class was 

defined with the vapor pressure Cs2MoO4 and the release coefficients developed for Cs. The Mo 

vapor pressure is so exceedingly low that the net release is limited by the vapor pressure 

transport term. Because there is significantly more Mo than Cs in the radionuclide inventory, 

only a portion of the Mo was added to the new Cs2MoO4 radionuclide class.

There are 69 isotopes in the treatment of consequences considered in the MACCS2 analysis, as 

described in NUREG/CR-7110, Volume 1. These isotopes are grouped into a set of nine 

chemical classes in the MELCOR analyses that generated the source terms used in the SOARCA 

analyses. Since release fractions are calculated by MELCOR at the level of chemical classes, it 

is reasonable and useful to examine how these same chemical classes influence the evaluation of 

risk. Volumes 1 and 2 of NUREG/CR-7110 discuss the importance of chemical classes.
47

The radionuclide input was reconfigured to (1) represent the dominant form of Cs as Cs2MoO4, 

(2) represent the dominant form of iodine (I) as CsI, and (3) represent the gap inventories 

consistent with the NUREG-1465 recommendations [14]. The MELCOR radionuclide transport, 

deposition, condensation and evaporation, and scrubbing models were all activated. The model 

for chemisorption of Cs to stainless steel was activated. In addition, the hygroscopic coupling of 

the steam or fog condensation or evaporation thermal-hydraulic solutions to the airborne aerosol 

size and mass was also activated [31].

4.3.4 Radionuclide Inventory 

One important input to MELCOR is the initial mass of the radionuclides in the fuel and their 

associated decay heat [31]. These values are important to the timing of initial core damage and 

the location and concentration of the radionuclides in the fuel. The radioisotopes in a nuclear 

reactor come from three primary sources: (1) fission products, which are the result of fissions in 

either fissile or fissionable material in the reactor core, (2) actinides, which are the product of 

neutron capture in the initial heavy metal isotopes in the fuel, and (3) radioactive decay of these 

fission products and actinides. Integrated computer models such as the TRITON sequence in 

SCALE exist to capture all of these interrelated physical processes, but they are intended 

primarily as reactor physics tools [35]. As such, their standard output does not provide the type 

of information needed for SOARCA. Therefore, this report describes a method for deriving the 

needed information. It is important to note that no changes to the physics codes were needed. 

The method described here merely extracts additional output from the TRITON sequence and 

combines it in a way that makes it useful for the SOARCA project.

4.3.4.1 Methods

Reactor physics codes implicitly account for both of the physical parameters of interest for 

SOARCA (i.e., decay heat power and radionuclide inventories), but they do not provide a 

mechanism to easily extract and combine these results. This section will describe the tools used 

to calculate the radioisotopic inventory and a new code developed to properly combine these 

results for use in the SOARCA calculations. The results were combined in a manner so as to 

capture actual plant operating data.

The TRITON sequence from SCALE 5.1 was used to develop input data for MELCOR. 

TRITON allows detailed two-dimensional calculations of reactor fuel, including the ability to 

deplete fuel to a user-defined level of accuracy. TRITON accurately models curvilinear surfaces 

such as cylindrical fuel rods and allows the fuel to be burned down to the subpin-cell level. 

There is no requirement to perform any homogenization of the two-dimensional geometry. 

TRITON allows for accurate depletion of highly self-shielded fuel such as poison pins. For more 

information, refer to the SCALE documentation [36].

The BLEND3 code was developed from previous work performed by ORNL, and its capabilities 

were extended for this study. BLEND3 uses the reactor-specific fuel loading from three 

different cycles, the nodal exposure, and the assembly-specific power data from the licensee to 

derive node-averaged radioisotopic inventories. TRITON uses generic fuel assembly data and 

ties them to specific reactor operating conditions. Then, BLEND3 performs the following tasks. 

First, for a given node, BLEND3 identifies which specific power ORIGEN output files are 
48

assigned to the specified input power. Second, for three different cycles of fuel, BLEND3 

interpolates a radioisotopic inventory from the relevant ORIGEN output files. Finally, using the 

input volume fractions for the three different cycles of fuel, BLEND3 creates a new, 

volumetrically averaged ORIGEN output file for the node for the specified input conditions.

The PRISM module from SCALE 5.1 was then used to drive ORIGEN decay calculations using 

the newly created averaged ORIGEN output files as input. PRISM is a SCALE utility module 

that allows the user to automate the execution of a series of SCALE calculations. 

4.3.4.2 Peach Bottom Model

The Peach Bottom model is based on the Global Nuclear Fuel (GNF) 10×10 (GE-14C) fuel 

assembly. The GNF 10×10 is representative of a limiting fuel type actually being used in 

commercial BWRs. Figure 7 illustrates the GNF 10×10 model. The axial nodalization of the 

core is designed, in part, to account for changes in material composition and mass along the axial 

length of a typical fuel assembly. For example, some BWR fuel assembly designs (modern 

10×10 assemblies, for example) incorporate fuel rods of different lengths within a single 

assembly. As a result, the amount of UO2 and other constituents can differ at the top and bottom 

of an assembly. Discrete locations of fuel rod spacers along the axial height of an assembly also 

affect local Zircaloy mass. The distribution of material mass within the axial nodalization of the 

core takes these variations into account.

At nine different specific power histories, 27 different TRITON runs were performed to model 

three different cycles of fuel. The specific power histories ranged from 2 megawatt-days per

metric ton of uranium (MWd/MTU) to 45 MWd/MTU to cover all expected BWR operational 

conditions. For times before the cycle of interest, an average specific power of 25.5 MWd/MTU 

was used. For example, for second cycle fuel, the fuel was burned for its first cycle using 

25.5 MWd/MTU, allowed to decay for an assumed 30-day refueling outage, and then nine

different TRITON calculations were performed with specific powers ranging from 

2 to 45 MWd/MTU). The BLEND3 code was then applied to each of the 50 nodes in the 

MELCOR model using the average specific powers and volume fractions. Once new libraries 

for each of the 50 nodes in the model were generated, the final step in the procedure was to 

deplete each node for 48 hours. The decay heats, masses, and specific activities as a function of 

time were processed and applied as input data to MELCOR to define decay heat and the 

radionuclide inventory. The SOARCA application, in keeping with the intent of using best 

estimate approaches, based the Peach Bottom fuel analysis of decay power and radionuclide 

inventories on the assumption that the accident occurs at a point midway in a recent fuel cycle.
49

Figure 7 Schematic of modeling detail for BWR GNF 10×10 assembly

4.3.4.3 Surry Model

Previously, detailed input was developed for Surry in a separate NRC program investigating the 

source term from high-burnup uranium fuel. This study used the same methodology as the Peach 

Bottom model (Section 4.3.4.2) but extended the burn-up of the lead assembly to the licensing 

limit (i.e., above current best-estimate practices). Based on comparisons to the Peach Bottom 

decay heat, the best-estimate, midcycle decay power for a recent Surry fuel cycle is expected to 

be about 17–18 percent lower than that used in the SOARCA MELCOR analyses for Surry.

4.3.4.4 Evaluation of the Results

Very few measurements of decay heat exist, and those that do are not directly relevant to this 

study. Therefore, the discussion of the decay heat predictions will be limited to a comparison to 

previously published work. RG 3.54, “Spent Fuel Heat Generation in an Independent Spent Fuel 

Installation” [37] summarizes a source of decay heat predictions, and results from RG 3.54 will 

be used to assess the predictions in the current study. Decay heat for two decay times will be 

used as a check on the consistency of the results presented in this study. By interpolation of 

tables in RG 3.54 for a specific power of 27 MW/MTU, decay powers at 1 and 2 years following 

shutdown of 9.3 W/kgU and 5.1 W/kgU, respectively, are calculated. Using the results from the 

Peach Bottom calculations, the corresponding decay powers are 8.92 W/kgU and 4.734 W/kgU. 

The maximum difference between results is about 8 percent, which is considered acceptable 
50

given the best estimate nature of the SOARCA study compared to the methods used to generate 

the tables in RG 3.54.

A quantitative discussion of the radioisotopic predictions presented in this study would be of 

limited use, given the cycle-specific nature of this work. However, it is beneficial to discuss the 

relevant SCALE assessment. Specifically, the TRITON module has been assessed by 

M.D. DeHart and S.M. Bowman [38], S.M. Bowman and D.F. Gill [39], and I. Germina and 

I.C. Gauld [40]. These assessment reports use data from Calvert Cliffs, Obrigheim, San Onofre, 

and Trino Vercelles PWRs. The third report [40] summarized comparisons to decay heat 

measurements from four different BWR assemblies.
51

5.0 OFFSITE CONSEQUENCE ANALYSES

MACCS2 [41] is a consequence analysis code for evaluating the impacts of atmospheric releases 

of radioactive aerosols and vapors on human health and the environment. It includes all of the 

relevant dose pathways: cloudshine, inhalation, groundshine, and ingestion. Because it is 

primarily a PRA tool, it accounts for the uncertainty in weather that is inherent to an accident 

that could occur at any point in the future. WinMACCS is a user-friendly front end to MACCS2 

that facilitates selection of input parameters and sampling of uncertain input parameters, and it 

performs postprocessing of results. The final SOARCA calculations use WinMACCS 

Version 3.6. MACCS2 is still the computational engine underlying WinMACCS.

The SOARCA offsite consequence predictions used MACCS2 Version 2.5. This version 

includes a number of improvements to the original MACCS2 code, which can be categorized as 

follows:

• atmospheric transport and dispersion modeling improvements (e.g., morning and 

afternoon mixing heights, alternative Briggs plume rise model, and alternative long-range 

plume spreading model)

• capability to describe wind directions in 64 compass directions (instead of 16)

• increased limits on several input parameters (e.g., a limit of 200 plume segments instead 

of the previous limit of 4)

• up to 20 emergency-phase cohorts (instead of the original limit of 3) to describe 

variations in emergency response by segments of the population

• enhancements in the treatment of evacuation speed and direction to better reflect the 

spatial and temporal response of individual cohorts

• capability to run on a cluster of computers instead of an individual processor

• addition of several new options for LCF dose response (i.e., user-input yearly truncation 

value, user-input yearly truncation value with a lifetime restriction, and a piece-wise 

linear model)

An expert panel reviewed the MACCS2 code and modeling choices in August 2006, before 

specific work on Surry and Peach Bottom began. This expert panel review and the NRC staff 

recommendations influenced much of the development undertaken specifically to support the 

SOARCA work [42]. 

Subsequent parts of this chapter describe specific aspects of the consequence modeling in 

SOARCA that depart from previous studies such as NUREG-1150 [2].
52

5.1

The weather sampling strategy adopted for SOARCA uses the nonuniform weather-binning 

approach in MACCS2. This approach, which allows the user to specify a different number of 

random samples to be chosen from each bin, has been available since MACCS2 was first 

released

Weather Sampling

[41] but was not commonly used in the past. Weather binning is an approach used in 

MACCS2 to categorize similar sets of weather data based on windspeed, stability class, and the 

occurrence of precipitation. The SOARCA project chose this sampling strategy to improve the 

statistical representation of the weather, as is further discussed below.

The standard way of defining weather bins originated in the NUREG-1150 [2] analyses. A set of 

16 weather bins differentiates stability classes and wind speeds. An additional 20 weather bins 

include all weather trials in which rain occurs before the initial plume segment travels a distance 

of 32 kilometers (20 miles). The bins differentiate rain intensity and the distance the plume 

travels before rain begins. The parameters used to define the rain bins are the same as those used 

in NUREG-1150 and documented in the MACCS2 User’s Manual [41]. Because the strategy 

provides for weighting the particular trials chosen (based on the number of samples in the bin 

and the number of samples requested), the particular choice of a binning strategy is not important 

(provided a sufficient number of samples is chosen). However, a well-chosen binning strategy 

will reduce the number of samples required for adequate statistical precision. The binning 

strategy used in NUREG-1150 and for SOARCA ensures that the rain cases, which are only a 

fraction of the full year’s data, are adequately sampled, with the weighting factors used in the 

code accounting for the prevalence in the weather record.

For the nonuniform weather sampling strategy approach for SOARCA, the number of trials 

selected from each bin is the maximum of 12 trials and 10 percent of the number of trials in the 

bin. Some bins contain fewer than 12 trials. In those cases, all of the trials within the bin are 

used for sampling. This strategy results in roughly 1,000 weather trials for both Peach Bottom 

and Surry.

Previous calculations, such as NUREG-1150, used about 125 weather trials, including an 

additional strategy—rotation—to account for the probability that the wind might have been 

blowing in a different direction when the release began. This strategy uses wind-rose data 

constructed from the annual weather file to determine the probability that the wind might have 

been in any of the compass directions. The strategy used at the time of NUREG-1150 leveraged 

the weather data to get 125×16=1,750 results for the computational price of 125, but at a cost 

that the individual results are not truly independent. For the strategy chosen here, the trials are 

independent. 

MACCS2 does not allow the use of rotation in concert with the network evacuation option; 

therefore, rotation was not an option for SOARCA. The strategy adopted for SOARCA was a 

compromise between obtaining adequate statistical significance and keeping central processing 

unit time at a reasonable level.

5.2

The SOARCA project used 1 year of hourly meteorological data for each site (8,760 data points 

Weather Data
53

per site for each meteorological parameter). This was primarily accomplished through a 

cooperative effort, with the licensee using onsite meteorological tower observations. Each 

licensee provided 2 years of weather data. The licensees measured and reported hourly 

precipitation directly. Temperature measurements at two elevations on the site meteorological 

towers provided stability class data. The project based the specific year of data chosen for each 

reactor on data recovery (greater than 99 percent being desirable) and proximity to the target 

year for SOARCA, which is 2005. Different trends (e.g., wind-rose pattern and hours of 

precipitation) between the years were estimated to have a relatively minor (less than 25 percent) 

effect on the final results. The next subsection discusses the specific details of the weather data.

For the weather record years and the particular data used in SOARCA, the recovery of data was 

in excess of 90 percent. The missing data were bridged over using the hourly records before and 

after by employing “Procedures for Substituting Values for Missing NWS [National Weather 

Service] Meteorological Data for Use in Regulatory Air Quality Models,” dated 

July 7, 1992 [43]. The meteorological data parameters were formatted for the MACCS2 

computer code.

The NRC staff used the methodology described in NUREG-0917, “Nuclear Regulatory 

Commission Staff Computer Programs for Use with Meteorological Data,” issued July 1982 [44]

to perform quality assurance evaluations of all meteorological data presented. Further review 

used computer spreadsheets. The NRC staff ensured a joint data recovery rate in the 90th

percentile, which is in accordance with RG 1.23, “Meteorological Monitoring Programs for 

Nuclear Power Plants,” Revision 1, issued March 2007 [45] for the wind speed, wind direction, 

and atmospheric stability parameters. In addition, it evaluated atmospheric stability to determine 

if the time of occurrence and duration of reported stability conditions were generally consistent 

with expected meteorological conditions (e.g., neutral and slightly stable conditions 

predominated during the year with stable and neutral conditions occurring at night and unstable 

and neutral conditions occurring during the day). The mixing height data came from the 

U.S. Environmental Protection Agency’s (EPA’s) SCRAM database4 (using data from the years 

1984–1992). Data needed for MACCS2 includes 10-meter wind speed, 10-meter wind direction 

in 64 compass directions, stability class (using the Pasquill-Gifford scale and representative 

values of 1–6 for stability classes A–F/G (see Section 5.2.1)), hourly precipitation, and diurnal 

(morning and afternoon) seasonal mixing heights. 

All of the SOARCA consequence analyses included boundary weather, but it was imposed 

beyond the outer boundary (50 miles or ~80.5 kilometers) for which results are reported. Thus, 

the choice of boundary weather had no influence on the consequence results that are reported. 

Appendices in the companion Peach Bottom and Surry reports contain the specific parameters 

chosen to describe the boundary weather.

5.2.1 Summary of Weather Data

Table 2 presents a summary of the meteorological statistical data and shows that the annual 

average ground-level wind speeds were generally low, ranging from 2.02 to 2.27 meters per 

second (m/s) at Surry and 2.12 to 2.17 m/s at Peach Bottom. The atmospheric stability 

 4 The EPA SCRAM Web site is http://www.epa.gov/scram001/mixingheightdata.htm.
54

 

0.00%

2.00%

4.00%

6.00%

8.00%

10.00%

12.00%

N

NNE

NE

ENE

E

ESE

SE

SSE

S

SSW

SW

WSW

W

WNW

NW

NNW

 

-50

0

50

100

150

200

250

300

350

400

0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24

Hour of the Day

Number of Occurrences

Unstable

Neutral

Stable

frequencies were consistent with expected meteorological conditions. The neutral and slightly 

stable conditions predominated during the year, with stable and neutral conditions occurring at 

night and unstable and neutral conditions occurring during the day. 

Figure 8 and Figure 9 show the wind direction (direction the wind blows toward) and 

atmospheric stability (unstable,

5 neutral,6 and stable7

) data for the years that were actually used 

in the consequence analyses (i.e., 2006 for Peach Bottom and 2004 for Surry). The MACCS2 

calculations used the Pasquill-Gifford stability classes. These classes were only parsed into 

unstable (A–C), neutral (D), and stable (E–F) conditions for Figure 8 and Figure 9 for 

comparisons with expected weather patterns.

Table 2 Statistical Summary of Raw Meteorological Data for SOARCA Nuclear Sites

Parameter

Peach Bottom Surry

Year 2005 Year 2006 Year 2001 Year 2004†

Average Wind Speed (m/s) 2.17 2.12 2.02 2.27

Yearly Precipitation (hr) 588

(6.7%)

593

(6.8%)

388

(4.4%)

521

(5.9%)

Atmospheric 

Stability (%)

Unstable 21.43 20.56 7.09 3.94

Neutral 63.97 62.34 69.67 77.59

Stable 14.60 17.10 23.24 18.47

Joint Data Recovery (%) 97.53 99.25 99.58 99.24 † Year 2004, used in the Surry meteorological analysis, is a leap year (8,784 total hourly data points versus 8,760 hourly data points for 

a regular annual period).

Figure 8 Peach Bottom—Year 2006—wind-rose and atmospheric stability chart

 5 This corresponds to Pasquill-Gifford stability classes A, B, and C. 6 This corresponds to Pasquill-Gifford stability class D. 7 This corresponds to Pasquill-Gifford stability classes E, F, and G.
55

 

0.00%

2.00%

4.00%

6.00%

8.00%

10.00%

12.00%

N

NNE

NE

ENE

E

ESE

SE

SSE

S

SSW

SW

WSW

W

WNW

NW

NNW

 

-50

0

50

100

150

200

250

300

350

400

0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24

Hour of the Day

Number of Occurrences

Unstable

Neutral

Stable

Figure 9 Surry—Year 2004—wind-rose and atmospheric stability chart

5.3

An objective of the SOARCA project was to model emergency response in a more detailed and 

realistic manner using site-specific emergency planning information. The analysis included 

modeling of the timing of onsite and offsite decisions and implementation of protective actions 

applied to multiple population segments (called cohorts). Advances in consequence modeling—

specifically the development of WinMACCS—made it easier to integrate protective action

decision timing and response of the public into the consequence analysis, resulting in an 

evolutionary advancement over previous studies. 

Emergency Response Modeling

Emergency response programs for NPPs are designed to protect public health and safety in the 

unlikely event of a radiological accident. These emergency response programs are developed, 

tested, and evaluated and are in place as an element of the NRC’s defense in depth policy. 

Detailed plans for onsite and offsite response are approved by the NRC and the Federal 

Emergency Management Agency respectively. 

Offsite response organization emergency plans are required to include detailed evacuation plans 

for the 10-mile emergency planning zone (EPZ) [46]. Site-specific information was obtained 

from ORO emergency response plans to support development of timelines for protective action 

implementation. Site specific planning elements were modeled, for example whether evacuation

of schools follows declaration of a site area emergency or a general emergency. The SOARCA 

project integrated response plan elements and a best estimate of protective action decision timing 

that was based upon actual biennial exercise history. Specific population cohorts were identified 

and their evacuation timing modeled. This detailed modeling was undertaken for the SOARCA 

project to improve the overall fidelity of the consequence analyses.

Figure Figure 10 shows the 10- and 20-mile radial distances around the Peach Bottom site.
56

Figure 10 10- and 20-mile radial distances around the Peach Bottom site

The SOARCA project assessed the ORO protective action decision-making process as detailed in 

emergency plans and developed a best estimate of implementation of those decisions by ORO 

populations within the 10-mile EPZs. The project also assessed possible variations of emergency 

response for the two sites studied, including evacuation and sheltering of population groups 

beyond the EPZ to a distance of 20 miles from the plants. As discussed in 

NUREG-0654/FEMA-REP-1, Revision 1, “Criteria for Preparation and Evaluation of 

Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants,” 

issued November 1980 [46], detailed planning within 10 miles would provide a substantial base 

for expansion of response efforts, should this prove necessary. Any response beyond the EPZ 

was expected to be limited to areas where dose projections indicate protective actions are 

necessary. State or county response agencies would inform the OROs of these projections in 

areas beyond the EPZ. Protective actions would be implemented in these areas beyond the EPZ

in an ad hoc manner, which means that such actions would follow the existing local all-hazards 

emergency response plans. 

For dose calculation purposes, evacuees are treated in the model as traveling to a point 30 miles 

from the site. This treatment is consistent with previous calculations (e.g., NUREG-1150) where 

evacuees moved 10 miles beyond the evacuation zone, at which point they were assumed to 

receive no further dose. Previous analyses chose the evacuation zone to represent the EPZ and 

did not consider shadow evacuation. The SOARCA analyses consider a shadow evacuation 

beyond the EPZ out to a distance of 20 miles from the plant. Thus, evacuating to a 30-mile 

radius results in the outermost evacuees traveling 10 miles beyond their initial location.

A shadow evacuation is the voluntary (self-initiated) evacuation of members of the public from 

areas that are not under official evacuation orders and typically occurs when a large scale 
57

evacuation is ordered. Shadow evacuations are often reported and observed, but there is little 

quantitative data available regarding these evacuations. SOARCA models a shadow evacuation 

of 20 percent of the public residing in the 10- to 20-mile area beyond the EPZ [48] based on data 

from a national telephone survey of residents of EPZs. In the survey, about 20 percent of people 

who had been previously asked to evacuate had also evacuated for situations in which they were 

asked not to evacuate (e.g., shadow evacuation). The size of the shadow evacuation is greatly 

affected by the emergency messaging used by the local authorities and could be larger or smaller. 

A shadow evacuation can delay the evacuation of people closer to the plant increasing their risk 

of exposure. SOARCA modeled shadow evacuations to improve realism.

The initiating event for many of the accident scenarios considered by SOARCA is a large 

earthquake close to the plant site. For this event, it was assumed that severe damage would be 

generally localized (e.g., 30–40 kilometers from the site). The SOARCA team considered the 

effects of such an earthquake on emergency response capabilities onsite and offsite as well as the 

evacuation speed of the public. However, considerable uncertainty exists in characterizing the 

impacts of an earthquake, and the SOARCA project therefore addressed the earthquake effects in 

a separate analysis. A consequence analysis was performed for the accident sequences for each 

site, and a single seismic analysis was performed for the more challenging accident postulated 

for each site.

The study performed a limited and conservative seismic analysis of local infrastructure, which 

may affect evacuation activities for each site. The seismic analysis indicated that long-span 

bridges, typically bridges crossing the river and interstate roadway crossings (at Surry), close to 

each site are unlikely to survive the earthquake and are assumed to be impassable throughout the 

emergency response. The study also assumed that some smaller bridges and road crossings 

would fail, as well as some roadways where underlying soils could slide off into adjacent 

waterways. Residential and commercial structures would be damaged but generally would 

survive the earthquake. The local electrical grid is assumed to be out of service through 

automatic shutdown or equipment failures; however, it is not expected that power would be out 

within the entire 10-mile (314-square-mile) EPZ. A limited backup power system is in place for 

the sirens at Peach Bottom, allowing some sirens to operate. Backup power is available for the 

sirens at Surry and those sirens are assumed to operate. OROs would perform route alerting to 

notify the population of the need to take protective actions in areas where sirens are not 

functional. Route alerting consists of emergency responders driving through neighborhoods 

using loudspeakers or going door to door to notify residents of the emergency and is a routine 

and effective method of informing the public [47]. Response parameters that may be affected by 

an earthquake (e.g., mobilization of the public, evacuation speed, shielding) were adjusted to 

reflect the potential impact.

5.3.1 Base Case Analyses of Emergency Response

The SOARCA project used WinMACCS to develop and model a case for each accident sequence 

that resulted in a radioactive release to the environment that would invoke protective actions. 

Initial protective actions at Surry, for which Supplement 3 to NUREG-0654/FEMA-REP-1, 

Revision 1 [63] provides guidance, would likely include evacuation of the 2-mile zone around 

the NPP and of a 5-mile downwind keyhole and would expand to 10 miles, if necessary, based 

on dose projections. Pennsylvania implements a 360-degree, 10-mile evacuation. For 
58

consistency in approach, the analyses for both sites included 360-degree evacuation of the public 

residing within the 10-mile EPZ. FigureFigure 11 is an example of evacuation routing. In 

addition, the analyses included a 20 percent shadow evacuation of the public residing in the 10-

to 20-mile area beyond the EPZ [48]. The population beyond 20 miles was not assumed to 

evacuate, although this segment of the population is relocated if projected doses exceed EPA 

guidelines.

The project established six cohorts for each site, each of which represents a discrete segment of 

the population that has different response characteristics. The use of six cohorts provides greater 

fidelity in the treatment of emergency response than was possible for previous studies, because 

MACCS2 previously allowed a maximum of three cohorts. As a general assumption, the 

accident scenario was assumed to occur during school hours, and one cohort was established for 

schoolchildren within the EPZ. Other cohorts included the general public within the EPZ, 

special facilities within the EPZ, the evacuation tail, shadow evacuees, and a nonevacuating

cohort. The nonevacuating cohort represents a small fraction (0.5 percent in this case) of the 

population who may choose not to evacuate when directed to do so. 

The SOARCA project used the evacuation time estimates (ETEs) provided by the licensees to 

develop speeds for evacuating cohorts. For NPPs, Appendix E to 10 CFR 50.47, “Emergency 

Plans,” requires that licensees provide an analysis of the time required to evacuate transient and 

permanent residents for various sectors and distances within the EPZ. Developed by licensees to 

support this requirement, an ETE is a tool that gives emergency managers information on how 

long it may take to evacuate a portion or all of the EPZ. Using this information, emergency 

managers can decide if evacuation is the most appropriate protective action for a specific 

accident. The site-specific ETEs were used to establish the evacuation-related input parameters 

for MACCS2. 

Figure 11 Typical evacuation routing

Using WinMACCS, the parameters related to emergency planning protective actions were input

to the consequence modeling. WinMACCS permits temporal and spatial elements of sheltering

and evacuation to be modeled and allows for the movement of multiple cohorts and 
59

accommodates speed and direction variations for each evacuating cohort. To develop the input, 

a review of the evacuation routes determined the likely directions in which evacuees would 

travel. The evacuation area was mapped onto a grid with 64 compass sectors and 15 radii that

formed the basis for the network evacuation model. All accident sequences at each site had the 

same evacuation network. Response timing and evacuation speed parameters were created 

specifically for each accident sequence. A newly developed option in WinMACCS and 

MACCS2 allowed an adjustment within grid elements to increase or decrease speeds based on 

expected traffic congestion. In addition, for cases where the hourly meteorological data included 

precipitation, the speeds of all evacuating cohorts were reduced. For dose modeling purposes, all 

evacuees were assumed to travel to a point 30 miles from the site. This distance accounts for the 

fact that the assembly sites are at some distance from the plant. Whether doses are actually 

received by the evacuees during any part of their travel is a complicated function of the direction 

of travel and the times and the directions of the plumes as they are released from the plant. Each 

plume disperses in a straight line in its own downwind direction.

5.3.2 Sensitivity Analyses of Emergency Response

After completion of the base case analysis, the following three variations were conducted as 

sensitivity analyses: 

(1) Evacuation to a distance of 16 miles from the plant. This analysis assessed the complete 

evacuation of 16 miles around the plant. It assumed the members of the public in the 16-

to 20-mile zone would shelter.

(2) Evacuation to a distance of 20 miles from the plant. This analysis developed an ETE for 

the 20-mile area to provide realistic modeling parameters for the movement of the public.

(3) Delay in implementing protective actions. This analysis included an assumption that 

there could be a 30-minute delay in implementing protective actions by the public. This 

sensitivity study assumes that a delay could occur in notification to offsite authorities, 

notification from offsite authorities to the public, receipt of the warning by the public, or 

for other reasons. The analysis assumed that cohorts take 30 minutes longer to start 

implementing protective actions than in the base case analysis.

The first two sensitivity studies used the Oak Ridge Evacuation Modeling System to create 

additional ETEs for each site to establish speeds and mobilization parameters for movement of 

the public residing between 10 and 20 miles. This system developed ETEs for the general public 

within the 10- to 20-mile zone. 

The development of the parameters for the sensitivity analyses showed that, for the larger 

evacuation areas, the travel speeds were typically slower than in the baseline analyses. This was 

because of the additional vehicle load on the roadway network in the more populated areas, such 

as Lancaster, PA. For the third sensitivity analysis, which assessed a delay in implementation, 

the speeds remained unchanged. The effects of these variations in emergency response on latent 

cancer fatality risk calculations are discussed further in Section 6.5.
60

5.4

A source term evaluation for each of the accident scenarios used MELMACCS

Source Term Evaluation 

[49], which reads

a MELCOR plot file and extracts information useful for source term definition for MACCS2. 

MELMACCS requires the selection of a number of user options. The following paragraphs 

describe the specific choices made for SOARCA.

The first set of choices is related to the chemical groups or classes to be included in the analysis. 

Here, the analyses included the standard set of fission product groups (i.e., the xenon, cesium, 

barium, iodine, tellurium, ruthenium, molybdenum, cerium, and lanthanum groups). A related 

quantity defining the burnup to be assumed when calculating the fission product inventory 

depends on the plant type. In an effort to provide a best estimate fission product inventory for 

Peach Bottom, SOARCA used an ORIGEN calculation to estimate the inventory at midcycle for 

which peak rod burnup is estimated to be 49 MWd/kg. MELMACCS used these data to specify 

the inventory for MACCS2, and the MACCS2 input is, therefore, consistent with the MELCOR 

calculation. Surry used a previously available fission product inventory based on the regulatory 

limit of burnup (65 MWd/kg for the peak fuel rod). This inventory is conservative.

A set of parameters define the ground elevation (grade) in the MELCOR reference frame, the 

height of the building from which release occurs, and the initial plume dimensions. The 

SOARCA MELCOR analyses use reactor shutdown as the reference time, so the time of accident 

initiation is always set to zero in the MELMACCS input.

MELMACCS calculates aerosol deposition velocities based on the geometric mean diameter of 

each aerosol bin as defined in the MELCOR analysis. Expert elicitation data form the basis for 

the deposition velocities, using the median value of the combined distribution from the 

experts [50]. This report applies the MELMACCS equation to determine the deposition 

velocities. The equation accounts for the dependence of deposition velocity on surface 

roughness, wind speed, and aerodynamic particle diameter. MELMACCS also accounts for 

independent particle size distributions for each of the chemical groups reported by MELCOR. 

Thus, the aerosol size distribution and the resulting deposition velocity distribution are generally 

different for each chemical group. This is accounted for by assigning different mass fraction 

distributions across the aerosol bins for each chemical group in the MACCS2 input. 

Typical values for surface roughness and mean wind speed, 0.1 m and 2.2 m/s, respectively, are 

used as inputs in MELMACCS. This value of surface roughness is commonly used for 

consequence analyses in the United States and is consistent with the value in NUREG-1150 [2]. 

However, a value that is more representative of the terrain at the two sites is also investigated as 

a sensitivity study and the results documented in the reports for the two sites. The specific 

weather files determined the mean wind speeds used in the consequence analyses. Table 3

displays the deposition velocities in SOARCA analyses for both Peach Bottom and Surry.
61

Table 3 Deposition Velocities Used in the SOARCA Analyses

Bin

#

Median Diameter

(μm)

Deposition Velocity

(m/s)

1 0.15 5.35×10-4

2 0.29 4.91×10-4

3 0.53 6.43×10-4

4 0.99 1.08×10-3

5 1.8 2.12×10-3

6 3.4 4.34×10-3

7 6.4 8.37×10-3

8 11.9 1.37×10-2

9 22.1 1.70×10-2

10 41.2 1.70×10-2

MELCOR results include the relative quantities of aerosols contained in each size bin listed in 

the table. MACCS2 uses this information, plus the deposition velocities in the table, to 

determine the rate of depletion of aerosols from the plume. Generally, the larger aerosols deposit 

more quickly and so are depleted more rapidly from the plume. The peak in the aerosol size 

distribution is usually a few micrometers (μm), which corresponds to a deposition velocity of a 

few millimeters per second.

Finally, significant releases were broken up into 1-hour plume segments. MACCS2 allows 

plume segments to travel in only one compass direction based on weather data. More plume 

segments can better represent plume transport and dispersion caused by possible changes in the 

weather (such as the wind direction) during the release. Longer plume segments were sometimes 

used for trivial releases, such as those where the segment content is a very small fraction of the 

total release. Finer resolution of these releases was not necessary to maintain the fidelity of the 

calculation. The MELCOR analyses provided the amount of each chemical element group in 

each aerosol bin for each plume segment.

5.5

The SOARCA project took weather data for each site from meteorological archives provided by 

each plant (see Section

Site-Specific Parameters

5.2). It then processed the raw data into 64 compass sectors to use the 

angular resolution capabilities in WinMACCS 3.6 and MACCS2 2.5. 

SECPOP2000 [51] initially created site files for 16 compass sectors, which is the only angular 

resolution supported by that code. WinMACCS then interpolated these site files onto the 

64-compass-sector grid used for the consequence analyses. The granularity of the population 

data for 16 compass directions is maintained for the 64 compass directions data. The 

SECPOP2000 population data were also scaled by a factor of 1.0533 to account for the average 

U.S. population growth between the years 2000 and 2005. 2010 U.S. Census data was not used 

because most calculations were already completed by the time it was released. Changes in 
62

population over the last decade are not expected to have a significant impact on any of the 

calculated individual latent cancer fatality risks reported in Chapter 6.

Consequence analyses used the standard approach of evaluating accidents in the following two 

phases:

(1) Emergency phase. This phase begins with the initiating event and continues for about 

1 week. The release from the plant and plume transport through the MACCS2 grid 

occurs during this phase. This phase also includes emergency response (i.e., evacuation 

and relocation of the population to reduce exposures and doses). The project chose the 

length of this phase to ensure that all plumes can exit the calculational mesh during the 

period, because certain assumptions about doses (e.g., that all late phase doses are small 

enough to warrant applying the dose and dose rate effectiveness factor) could be 

questionable if the early phase was made too short.

(2) Long-term phase. This phase is the period following the emergency phase and continues 

for 50 years. Three actions take place during the long-term phase. Land that is 

contaminated above the level that is allowable for habitation is decontaminated and 

potentially interdicted for an additional period. During this time, the land is not available 

for human habitation. Land that cannot be restored to habitability is condemned, in 

which case the residents do not return during the long-term phase. 

Both sites needed a choice for surface roughness, which affects both vertical dispersion and 

deposition velocities. A generic value of 10 centimeters for surface roughness was selected for 

the consequence analyses, just as it had been in NUREG-1150 and most other previous studies. 

However, sensitivity studies also evaluated the effect of site-specific surface roughness. The 

reports for each of the sites document these sensitivity studies.

The project evaluated shielding factors applied to evacuation, normal activity, and sheltering for 

each relevant dose pathway (i.e., inhalation, deposition onto skin, cloudshine, and groundshine)

for each site based on values used in NUREG-1150 [2] and NUREG-6953, Volume 1, “Review 

of NUREG-0654, Supplement 3, ‘Criteria for Protective Action Recommendations for Severe 

Accidents,’” issued December 2007 [52]. A review of the discussion of shielding in the 

NUREG-1150 documentation suggests that the factors the authors considered were adequate for 

SOARCA purposes. One departure from the NUREG-1150 values is for normal activity. The 

SOARCA project reevaluated each of the normal activity values, assuming that the average 

person spends 19 percent of the day outdoors and 81 percent of the day indoors [63]. It 

evaluated the value for each of the pathways as a linear combination of 19 percent of the value 

for evacuation and 81 percent of the value for sheltering. 

For dose calculations, the project modeled evacuees as traveling to a distance of 30 miles from 

the plant. In addition, it relocated the nonevacuating cohort and the public beyond the EPZ from 

areas where the projected dose during the emergency phase exceeded a set of two upper bounds. 

These bounds were based on dose levels published by EPA, which are 1 to 5 rem. SOARCA

used the upper limit of this range (5 rem) to trigger hot-spot relocation for both Surry and Peach 

Bottom and used the lower limit of this range (1 rem) to trigger normal relocation for Surry,
63

while it used 0.5 rem for Peach Bottom, to be consistent with the Pennsylvania habitability 

criterion.

MACCS2 performs hot-spot relocation first and normal relocation second. The choices of times 

associated with normal and hot-spot relocation depended on the specific accident scenario,

because they are based on plume arrival. The scenario-specific time for completion of the 

relocation includes the time for response personnel to identify the involved area, for them to 

notify the residents within that area that relocation is necessary, and for the residents to remove 

themselves from the area. Because the timing of relocation is keyed to plume arrival, there is 

always a period of exposure before initiation of relocation. Volumes 1 and 2 of 

NUREG/CR-7110, discuss the specific choices for the parameters controlling the exposure 

period.

Site-specific values determine long-term habitability. Most States adhere to EPA guidelines that 

allow a dose of 2 rem in the first year and 500 millirem (mrem) per year thereafter. The EPA 

recommendation that has traditionally been implemented in MACCS2 is 4 rem over 5 years 

(2 rem in the first year + 4 years × 0.5 rem) of exposure, and this study adopts that convention. 

MACCS2 cannot explicitly use the EPA recommendation because MACCS2 accepts only one 

dose and one time period. Some States, like Pennsylvania, have a stricter habitability criterion

(i.e., 0.5 rem/year beginning in the first year). Thus, the habitability or return criterion is site 

specific, as Volumes 1 and 2 of NUREG/CR-7110 discuss further.

Some States have distributed potassium iodide (KI) tablets to people who live near commercial 

NPPs. KI has been distributed within the EPZ at the Peach Bottom and Surry sites. The purpose 

of the KI is to saturate the thyroid gland with iodine so that further uptake of iodine by the 

thyroid is diminished. If taken at the right time, KI can nearly eliminate doses to the thyroid 

gland from inhaled radioiodine. Ingestion of KI is modeled for half of the residents near plants 

where KI has been distributed by the State or local government. A further assumption is that 

most residents do not take KI at the optimal time (from shortly before to immediately after plume 

arrival), so the efficacy is only 70 percent (i.e., the thyroid dose from inhaled radioiodine is 

reduced by 70 percent).

Other site-specific parameters include farmland and nonfarmland values. These values are also

scaled from NUREG-1150 values using the Consumer Price Index as the basis for price 

escalation. A scaling factor of 1.09 accounts for inflation between the years 2002 and 2005. 

Land values influence the decision to decontaminate, interdict, or condemn land. If the assessed 

cost of decontamination is higher than the land value, the assumption is the land would be 

condemned. Because the public would not be allowed to return to condemned land and, 

therefore, no dose would be received, the land values did have a second-order effect on the 

predicted long-term health consequences.

5.6

The SOARCA analyses do not treat the ingestion of contaminated food and water. The 

reasoning is that abundant supplies of food and water are available in the United States and can 

be distributed to areas affected by a reactor accident. 

Non-Site-Specific Parameters
64

Much of the non-site-specific data used for consequence analysis in SOARCA are taken from 

reports that document a joint NRC/Commission of the European Communities expert elicitation 

study [50]. These data include atmospheric dispersion parameters, dry deposition velocities, wet 

deposition parameters, and acute health-effect parameters. In all cases, the point-value 

consequence analyses in SOARCA use median values extracted from the elicitation study [50]. 

The SOARCA analyses based the dose conversion factors on Federal Guidance Report (FGR) 

-13, “Cancer Risk Coefficients for Environmental Exposure to Radionuclides: Updates and 

Supplements,” CD Supplement, Revision 1, issued April 2002 [53]. This guidance report also 

recommended changes to the biological effectiveness factors (BEFs) for alpha radiation for two 

of the organs used to estimate latent cancer health effects, to be consistent with the evaluation of 

risk factors for cancers associated with those organs. The two organs are bone marrow and 

breast; for these organs, the BEFs for alpha radiation were changed from the standard value of 20 

to 1 and 10, respectively. Doses to these organs are used to evaluate occurrences of leukemia 

and breast cancer, respectively. The choice of BEFs for these tissues is dictated by 

EPA 402-R-93-076, “Estimating Radiogenic Cancer Risks,” issued June 1994 [54]. 

A February 2009 ORNL memo, “Risk Coefficients for SOARCA Project” [55], also 

recommended using dose to the pancreas as a surrogate for dose to soft tissue to estimate residual 

cancers. The reason for the choice of the pancreas dose coefficient for the “residual” cancer sites 

is because it serves as a reasonable surrogate for the residual group for both external radiation 

fields and the intake of radionuclides. Because MACCS2 does not currently read the data for the 

pancreas from the dose conversion factor file, a workaround was created. Values of the dose 

coefficients for the pancreas were copied into the organ called bladder wall. Thus, residual 

cancers are associated with the organ called bladder wall, which actually contains data for the 

pancreas. 

The SOARCA study applied a dose and dose rate effectiveness factor to all doses in the late 

phase of the offsite consequence calculation and to those doses in the early phase that were less 

than 20 rem to the whole body. This factor, which appears in the denominator, accounts for the 

fact that protracted low doses are perceived to be less effective in causing cancer than acute 

doses. The dose and dose rate effectiveness factor for all cancers except for the breast was 2.0, 

and for the breast, it was 1.0.

The January 2012 ORNL memorandum [74] also recommended risk factors for latent health 

effects that come from the National Research Council’s Committee on the Biological Effects of 

Ionizing Radiations (BEIR) V report, “Health Effects of Exposure to Low Levels of Ionizing 

Radiation,” issued 1990 [56], and are consistent with the modified dose conversion factor file 

described in the preceding paragraph. These risk factors include seven organ-specific cancers 

plus residual cancers that are not accounted for directly. In 2009, the National Research Council 

released the BEIR VII report, an additional study of the biological effects of ionizing radiation. 

No one-to-one correspondence exists between the cancers reported in BEIR VII and those in the 

earlier BEIR V report. Therefore, the dose coefficients of tissues of the body in FGR-13 may or 

may not be consistent with the BEIR VII cancer sites. Thus, the SOARCA staff decided to await 

EPA’s review of BEIR VII and subsequent update of FGR-13 before implementing BEIR VII 

risk coefficients.
65

Values from NUREG-1150 [2] provide the basis for decontamination parameters, which consist 

of two levels of decontamination, just as in NUREG-1150. The cost parameters associated with 

decontamination are adjusted to account for inflation using the Consumer Price Index. This 

report does not consider costs associated with a reactor accident; however, these parameters do 

affect decisions on whether contaminated areas can be restored to habitability and therefore 

affect predicted doses and risk of health effects.

5.7

Experts generally agree that it is difficult to characterize cancer risk because of the low statistical 

precision associated with relatively small numbers of excess cases at low doses. This limits the 

ability to estimate trends in risk. From an epidemiological standpoint, the number of LCFs 

attributable to radiation exposure from accidental releases from a severe accident would not be 

statistically detectable above the normal rate of cancer fatalities in the exposed population 

(i.e., the excess cancer fatalities predicted are too few to allow the detection of a statistically 

significant difference in the cancer fatalities expected from other causes among the same 

population). For example, in 2006, the World Health Organization estimated that 

16,000 European cancer deaths would be attributable to radiation released from the 1986 

Chernobyl NPP accident, but these predicted numbers are small relative to the several hundred 

million cancer cases that are expected in Europe through 2065 from other causes. Moreover, the 

World Health Organization concluded that “it is unlikely that the cancer burden from the largest 

radiological accident to date could be detected by monitoring national cancer statistics.” 

Estimating Latent Cancer Fatality Health Effects

New findings have been published from analyses of fractionated or chronic low-dose exposure to 

low, linear energy transfer radiation. In particular, these recent findings included a study of 

nuclear workers in 15 countries, studies of persons living in the vicinity of the Techa River in the 

Russian Federation who were exposed to radioactive waste discharges from the Mayak 

Production Association, a study of persons exposed to fallout from the Semipalatinsk nuclear test 

site in Kazakhstan, and studies in regions with high natural background levels of radiation. 

Cancer risk estimates in these studies are generally derived from the Japanese atomic bomb data. 

The most recent results from analyzing these data are consistent with a linear or linear-quadratic 

dose-response relationship of all solid cancers together and with a linear-quadratic dose-response 

relationship for leukemia. A linear-quadratic form for a dose model has a dependence on the 

square of the dose, as well as on the dose itself.

In the absence of additional information, the International Commission on Radiological 

Protection (ICRP), the National Academy of Sciences, and the United Nations Scientific 

Committee on the Effects of Atomic Radiation have each indicated that the current scientific 

evidence is consistent with the hypothesis that an LNT dose-response relationship exists between 

exposure to ionizing radiation and the development of cancer in humans.

Conversely, in “Dose-effect relationships and estimation of the carcinogenic effects of low doses 

of ionizing radiation,” dated March 30, 2005 [57], the French National Academy of Medicine 

advocates the following on page 1: 
66

A linear no-threshold relationship (LNT) describes well the relation between the 

dose and the carcinogenic effect in this dose range (0.2 to 3 Sv) [to the whole 

body] where it could be tested. However, the use of this relationship to assess by 

extrapolation the risk of low and very low doses deserves great caution. Recent 

radiobiological data undermine the validity of estimations based on LNT in the 

range of doses lower than a few dozen mSv which leads to the questioning of the 

hypotheses on which LNT is implicitly based.

Although the French National Academy of Medicine raises doubts about the validity of using 

LNT to evaluate the carcinogenic risk of low doses (less than 100 millisieverts (mSv) (10 rem)), 

and particularly for very low doses (less than 10 mSv (1 rem)), it did not articulate the exact 

value that should be ascribed to a dose threshold.

Ultimately, MACCS2 converts external and internal exposures to individual members of the 

public from collective organ dose to LCFs. The LNT model raises the concern that the 

summation of very small exposures may inappropriately attribute LCFs to individuals receiving 

these exposures. Organizations such as ICRP and the Health Physics Society (HPS) consider it 

to be an inappropriate use of these exposures. While the National Council on Radiation 

Protection and Measurements (NCRP) supports the LNT model, it recommends binning 

exposures into ranges and considering those ranges separately. Moreover, in situations involving 

very small exposures to large populations, ICRP and NCRP have noted that the most likely 

number of excess health effects is zero when the collective dose to such populations is equivalent 

to the reciprocal of the risk coefficient (about 20 person-sieverts (Sv) (2,000 person-rem)). 

Nevertheless, issues remain related to assessing public exposure, estimating offsite 

consequences, and communicating these assessments to the public. Several organizations such 

as ICRP have addressed this issue. In its most recent recommendations (ICRP Report 103, “The 

2007 Recommendations of the International Commission on Radiological Protection,” approved 

March 2007 [58]), ICRP stated the following:

Collective effective dose is an instrument for optimization, for comparing 

radiological technologies and protection procedures. Collective effective dose is 

not intended as a tool for epidemiological studies, and it is inappropriate to use it 

in risk projections. This is because the assumptions implicit in the calculation of 

collective effective dose (e.g., when applying the LNT model) conceal large 

biological and statistical uncertainties. Specifically, the computation of cancer 

deaths based on collective effective doses involving trivial exposures to large 

populations is not reasonable and should be avoided. Such computations based 

on collective effective dose were never intended, are biologically and statistically 

very uncertain, presuppose a number of caveats that tend not to be repeated when 

estimates are quoted out of context, and are an incorrect use of this protection 

quantity.

Although ICRP provided qualitative guidance on situations where collective dose should not be 

used, it did not provide guidance on when these concepts actually are, and are not, appropriate, 

nor did it clearly articulate the boundaries within which the calculations are valid, as well as the 

dose ranges for which epidemiological and cellular or molecular data provide information on the 
67

health effects associated with radiation exposure. ICRP did note, however, that when ranges of 

exposures are large, collective dose may aggregate information inappropriately and could be 

misleading for selecting protective actions.

The National Academy of Sciences reported the following [56]:

The magnitude of estimated risk for total cancer mortality or leukemia has not 

changed greatly from estimates in past reports such as Biological Effects of 

Ionizing Radiation (BEIR) and recent reports of the United Nations Scientific 

Committee on the Effects of Atomic Radiation (UNSCEAR) and ICRP. New 

data and analyses have reduced sampling uncertainty, but uncertainties related to 

estimating risk for exposure to low doses and dose rates and to transporting risks 

from Japanese A-bomb survivors to the U.S. population remain large.

The National Academy of Sciences goes on to conclude that “current scientific evidence is 

consistent with the hypothesis that there is a linear, no-threshold dose-response relationship 

between exposure to ionizing radiation and the development of cancer in humans.”

Many groups acknowledge the uncertainties associated with estimating risk for exposure to low 

radiation doses. One important question that remains is what health consequences, if any, are 

attributable to very low radiation exposure. In its most recent recommendations (ICRP 

Report 103 [58]) described above, ICRP warned that the computation of cancer deaths based on 

collective effective doses involving trivial exposures is not reasonable and should be avoided. 

However, the report did not explicitly provide a quantitative range for which exposures should 

not be considered. However, in its 2007 Report 104, “Scope of Radiological Protection Control 

Measures” [59], ICRP concludes that the radiation dose that is of no significance to individuals 

should be in the range of 20–100 microsieverts (μSv) (2–10 mrem) per year whole body dose. 

The International Atomic Energy Agency has stated that an individual dose is likely to be 

regarded as trivial if it is on the order of some several millirem per year.

Alternatively, HPS developed a position paper, “Radiation Risk in Perspective,” revised 

August 2004 [60], to specifically address quantitative estimation of health risks. This position 

paper concludes that quantitative estimates of risk should be limited to individuals receiving a 

whole body dose greater than 0.05 Sv (5 rem) in 1 year or a lifetime dose greater than 0.1 Sv 

(10 rem) in addition to natural background radiation. HPS also concluded that risk estimates 

should not be conducted below these doses. The position paper further states that low dose 

expressions of risk should be only qualitative, discuss a range of possible outcomes, and 

emphasize the inability to detect any increased health detriment.

The LNT model provides a viewpoint that is consistent with the NRC regulatory approach, and 

past analyses using the MACCS2 code have assumed an LNT dose-response model. The NRC is 

neither changing nor contemplating changing radiation protection standards and policy as a result 

of an approach taken in the SOARCA study to characterize offsite health consequences for low 

probability events. Still, the NRC can use different approaches for different applications. 

Therefore, the SOARCA analyses consider a range of dose truncation values ranging from LNT 

to a dose truncation level based on the HPS position that there is a dose below which, because of 
68

uncertainties, a quantified risk should not be assigned, which is 5 rem/year with a lifetime dose 

limit of 10 rem. 

The SOARCA analyses also considered two additional dose truncation levels. One is the 

10 mrem/year dose truncation value suggested in ICRP Report 104 [59]; the other is U.S. 

average background radiation combined with average annual medical exposure as a dose 

truncation level (abbreviated as US BGR), which is 620 mrem/year. Results for three of these 

four dose truncation levels are reported for each of the accident scenarios considered in the 

SOARCA study. The results for the 10-mrem/year dose truncation levels were calculated but are 

not included in the report because the results are very similar to LNT and are also always slightly 

less than the LNT results. 

5.8

The statistic that is chosen to convey the likelihood of LCFs resulting from an accident at an 

NPP is the mean, population-weighted individual risk. This value is more meaningful than the 

predicted number of LCFs in the sense that it may be compared with cancer fatality rates that 

have other causes. Individual risks can be presented as conditional risks (i.e., as if the accident 

had taken place) or as absolute risks (i.e., accounting for the likelihood of the accident occurring 

per year of reactor operation). The latter definition of risk is more useful, because it conveys the 

full meaning of risk, which is probability (or frequency) times consequence. 

Risk Metrics Reported

The term “population-weighted” in the preceding paragraph carries the meaning of the effect of 

population distribution along with wind-rose probabilities of the predicted risk. This statistic is 

simply the number of predicted fatalities divided by the population within a specified region. 

The use of the word “mean” is intended to convey that the results are weighted averages over the 

annual weather trials used in the analysis. The work presented in this report considers 

uncertainty in the weather. Subsequent work will explore the effect on the predictions of 

uncertainties in other input parameters.

The mean, population-weighted individual risks range from 0 to 50 miles. The 0- to 10-mile 

range represents the population within the EPZ. Analyses of severe accident mitigation and 

severe accident mitigation design alternatives generally use the range from 0 to 50 miles. 
69

6.0 RESULTS AND CONCLUSIONS

To assess the benefits of the various mitigative measures and to provide a basis for comparison 

to past analyses of unmitigated severe accident scenarios, the SOARCA project treated the 

selected scenarios in two separate and distinct manners. In the first, it analyzed scenarios that 

included an assessment of reasonable mitigation measures for which procedures and equipment 

(and training) exist. Alternatively, if adequate time exists, this would suffice for implementation 

in lieu of fully developed procedures and training (especially for simple actions), e.g., refilling 

water storage capabilities. In the second manner, it assumed that the key or vital measures 

necessary to prevent core damage or to mitigate radiological release were not taken to compare 

them with previous analyses of unmitigated scenarios. This comparison could reveal the benefits 

of improved severe accident phenomenological understanding and modeling. 

6.1

While this report does not determine the respective likelihoods of the mitigated and unmitigated 

cases of each scenario, the SOARCA results demonstrate the potential benefits of employing 10 

CFR 50.54(hh) mitigation enhancements for the scenarios analyzed. MELCOR analyses were 

used both to confirm the time available to implement mitigation measures and to confirm that 

those measures, when successfully implemented, are effective in preventing core damage or 

significantly reducing radiological releases. When successful mitigation is assumed, the 

MELCOR results indicate no core damage for all scenarios except the Surry STSBO and its 

TISGTR variant. The security-related mitigation measures that provide alternative ac power and 

portable diesel-driven pumps are especially helpful in counteracting SBO scenarios. For the 

Surry STSBO and its TISGTR variant, the mitigation is sufficient to flood the containment 

through the containment spray system to cover core debris resulting from vessel failure. For the 

ISLOCA scenario, installed equipment unrelated to 10 CFR 50.54(hh) is effective in preventing 

core damage owing to the time available for corrective action. Mitigation results are included in 

Tables 4 through 7.

Mitigation

6.2

An important result of the MELCOR analyses was that the select severe accidents proceed much 

more slowly than the SST1 case from the 1982 Siting Study. The reasons for this are threefold: 

(1) research and development of better phenomenological modeling has produced a much more 

protracted and delayed core degradation transient with substantial delays of reactor vessel 

failure, (2) all aspects of accident scenarios receive more realistic treatment, which includes 

more complete modeling of plant systems and often yields delays in core damage and 

radiological release, and (3) the scope of SOARCA focuses on the more likely and important 

accident scenarios, while past treatments included less likely accident progressions. In general, 

the bounding approaches in past simplified treatments used qualitative logical models. In 

SOARCA, where specific self-consistent scenarios are analyzed in an integral fashion using 

MELCOR, the result is that accident conditions or attributes that contribute to a more severe 

response in one area may produce an ameliorating effect in another area.

Accident Progression and Radionuclide Release
70

For the LTSBO scenarios for both Peach Bottom and Surry (the most likely severe accident 

scenario for each plant considered in SOARCA) analyzed assuming no mitigation, core damage 

begins in 9 to 16 hours, and reactor vessel failure begins at about 20 hours. Offsite radiological 

release due to containment failure begins at about 20 hours for Peach Bottom (BWR) and at 45 

hours for Surry (PWR). The SOARCA analyses therefore show that time may be available for 

operators to take corrective action and get additional assistance from plant technical support 

centers even if initial efforts are assumed unsuccessful. For the most rapid events (i.e., the 

unmitigated STSBO in which core damage may begin in 1 to 3 hours), reactor vessel failure 

begins at roughly 8 hours, possibly allowing time to restore core cooling and prevent vessel

failure. In these cases, containment failure and radiological release begins at about 8 hours for 

Peach Bottom and at 25 hours for Surry. For the unmitigated Surry ISLOCA, the offsite 

radiological release begins at about 13 hours and in the other bypass event analyzed, the 

TISGTR, the radiological release begins at about 3.5 hours but is shown by analyses to be 

substantially smaller than the 1982 Siting Study SST1 release.

Table 4 and Table 5 provide key accident progression timing results for SOARCA scenarios. 

Table 4 shows the same times for lower head failure and start of the release to the environment, 

because drywell shell melt-through occurs about 15 minutes after lower head failure. 
71

Table 4 Peach Bottom Accident Progression Timing Results

* Blackstart of the reactor core isolation cooling (RCIC) system refers to starting RCIC without any ac or dc 

control power. Blackrun of RCIC refers to the long-term operation of RCIC without electricity, once it has 

been started. This typically involves using a portable generator to supply power to indications such as reactor 

pressure vessel (RPV) level to allow the operator to manually adjust RCIC flow to prevent RPV overfill and 

flooding of the RCIC turbine. STSBO RCIC blackstart and limited blackrun is credited as an unmitigated case 

for SOARCA purposes because the licensee has included its use in procedures. Past NRC severe accident 

analyses of STSBO scenarios did not credit blackstart of RCIC. A sensitivity calculation without blackstart was 

therefore performed to provide a basis for comparison to past analyses.

** A scenario with 10 CFR 50.54(hh) mitigation, but without RCIC blackstart was not analyzed.

Scenario

Mitigated Unmitigated

Time to 

start of 

core 

damage 

(hours)

Time to 

lower 

head 

failure 

(hours)

Time to start 

of release to 

environment 

(hours)

Time to 

start of 

core 

damage 

(hours)

Time to 

lower 

head 

failure 

(hours)

Time to start 

of release to 

environment 

(hours)

Long-term 

SBO No Core Damage 9 20 20

Short-term 

SBO with 

RCIC 

Blackstart*

No Core Damage** 7 17 17

Short-term 

SBO without 

RCIC 

Blackstart

Not Applicable** 1 8 8
72

Table 5 Surry Accident Progression Timing Results

* Although the time at which release to the environment starts is the same in the mitigated and unmitigated 

cases, containment failure is delayed by about 46 hours in the mitigated case compared to the unmitigated 

case.

The SOARCA study also demonstrated that the magnitude of the environmental radionuclide 

release is likely to be much smaller than the SST1 source term, again as a result of (1) extensive 

research and improved modeling, (2) integrated and more complete plant simulation, and (3) the 

SOARCA project’s focus on the more likely severe accident scenarios, while past treatments 

included less likely accident progressions. Historically important radionuclides have included 

the more volatile fission products (i.e., those released in greater quantity from the overheated 

fuel) such as iodine and cesium. These two radionuclides have also been useful representatives 

of radionuclides with a short half-life (iodine) and those with a long half-life (cesium). 

SOARCA analysis typically predicts iodine releases on the order of 1-2 percent for the dominant 

scenarios with the highest releases on the order of 10-15 percent for the lower frequency, more 

severe scenarios. By contrast, the SST1 source term in the 1982 Siting Study assumed an iodine 

release of 45 percent. With respect to cesium, SOARCA predicts releases of 2 percent or less. 

By contrast, the SST1 source term assumed a cesium release of 67 percent. Figure 12 and Figure 

13 provide the radionuclide release results for iodine and cesium.

Scenario

Mitigated Unmitigated

Time to 

start of

core 

damage 

(hours)

Time to 

lower 

head 

failure 

(hours)

Time to start 

of release to 

environment 

(hours)

Time to 

start of 

core 

damage 

(hours)

Time to 

lower 

head 

failure 

(hours)

Time to start 

of release to 

environment 

(hours)

Long-term 

SBO No Core Damage 16 21 45

Short-term 

SBO 3 7 66 3 7 25

Short-term 

SBO with 

thermally 

induced 

steam 

generator 

tube 

rupture

3 7.5 3.5* 3 7.5 3.5

Interfacing 

systems 

LOCA

No Core Damage 13 19 13
73

Figure 12 Iodine releases to the environment for SOARCA unmitigated scenarios and the 

1982 Siting Study SST1 case

Figure 13 Cesium releases to the environment for SOARCA unmitigated scenarios and the 

1982 Siting Study SST1 case 

Past PRAs and consequence studies showed that sequences involving large early releases were 

important risk contributors. For example, the 1982 Siting Study SST1 results were controlled by 

an internally initiated event with a large early release that was assigned a representative 
74

frequency of 1×10-5 per reactor-year. However, in the SOARCA study, no sequences resulted in 

a large early release, even considering external events and unsuccessful mitigation. This is a 

result of research conducted over the last several decades that has shown that phenomena earlier 

believed to lead to a large early release are of extremely low probability or not physically 

feasible. This research was focused on phenomena that had been previously assumed to be 

prime contributors to severe accident risk, including direct containment heating and alpha mode 

failure.

The PWR SBO with a TISGTR was historically believed to result in a large, relatively early 

release, potentially leading to higher offsite consequences. However, MELCOR analysis of 

Surry performed for SOARCA shows that the release is small, because other reactor coolant 

system piping inside containment (i.e., hot let nozzle) fails soon after the tube rupture and 

thereby retains fission products within the containment. Also, the release was somewhat 

delayed; for the STSBO where loss of injection occurred at the start of the accident, the tube 

rupture and release began about 3.5 hours into the event. Moreover, core damage, tube rupture, 

and radiological release could be delayed for many hours if auxiliary feedwater were available 

even for a relatively short time.

6.3 Offsite Radiological Consequences

The result of the accident progression and source term analysis is that releases are delayed, 

smaller, and more dispersed relative to the 1982 Siting Study SST1 case. This fact, combined 

with the realistic simulation of emergency response and the greater distances radioactive material 

is expected to disperse, led to essentially no risk of early fatalities being calculated as close-in 

populations were evacuated before or shortly after plume arrival. 

Because of the last factor, significantly more of the latent cancer fatality risk in the selected 

SOARCA scenarios comes from low doses compared to the results of the SST1 source term from 

the 1982 Siting Study. Therefore, a dose truncation significantly reduces the quantified LCF risk 

in the SOARCA scenarios, much more so than a dose truncation would have for the SST1 from 

the 1982 Siting Study. 

Latent health effects calculated using any of the dose-response models (in combination with the 

frequency of release) referenced in this study are small in comparison to the NRC Safety Goal. 

Much of the LCF risk was in fact derived from the small doses received by populations returning 

to their homes in accordance with emergency planning guidelines. Because much of the health 

risk is caused by the return of the population, it is therefore controllable. For example, for the 

Peach Bottom LTSBO, for individuals living within the EPZ, 99 percent of the LCF risk derives 

from the long-term dose received by the population returning to their homes and being exposed 

to small radiation doses. Similarly, about 70 percent of the LCF risk to individuals within 

50 miles is from returning home. The percentage is larger for the EPZ, because of its evacuation 

before the start of the release. Here, the calculation of scenario-specific LCF risk, though very 

small, is strongly influenced by the relationship between low-dose health effects modeling and 

criteria for allowing the population to return. 
75

Tables 6 and 7 show estimates of conditional (i.e., assuming the accident has occurred) 

scenario-specific probabilities of an LCF range from roughly 10-4 to 10-5

, using the LNT 

dose-response model (other dose models result in lower or much lower conditional risk). The 

tables also provide the product of this value and the scenario CDF, which is best described as the 

scenario-specific risk of LCF for an individual located within 10 miles of the plant. 

Scenario-specific risk of an LCF for an individual within 10 miles of the plant is on the order of 

10-9 to 10-11 per reactor-year. These risk estimates are millions of times lower than the general 

risk of a cancer fatality in the United States from all causes, approximately 2×10-3 per year and 

thousands of times lower than the NRC Safety Goal. 

Comparisons of SOARCA’s calculated LCF risks to the NRC Safety Goal [72] and the average 

annual U.S. cancer fatality risk from all causes [73] 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 reader should be aware. 

Relative to the safety goal comparison, the safety goal is intended to encompass all accident 

scenarios. SOARCA does not examine all scenarios typically considered in a PRA, even though 

it includes the important scenarios. SOARCA represents a mix of limited PRA models with a 

deterministic treatment of various long-term mitigating features. In fact, any analytical 

technique, including PRAs, will have inherent limitations of scope and method. As a result, 

comparison of SOARCA’s scenario-specific calculated LCF risks to the NRC Safety Goal is 

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

results in the ~ 10-10 range to approximate a summary risk from all scenarios, would yield a 

summary result that is also below the NRC Safety Goal of 2x10-6 or two in one million.

Relative to the U.S. average 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.
76

Table 6 Peach Bottom Results for Scenarios Assuming LNT Dose-Response Model

Scenario

Core damage 

frequency 

(CDF)

(per 

reactor-year)*

Mitigated Unmitigated

Conditional 

scenariospecific 

probability of 

latent cancer 

fatality for 

an individual 

located 

within 

10 miles

Scenariospecific risk

(CDF x 

Conditional)

of latent cancer 

fatality for an 

individual 

located within 

10 miles 

(per reactoryear)

Conditional 

scenariospecific 

probability of 

latent cancer 

fatality for an 

individual 

located within 

10 miles

Scenario-specific 

risk

(CDF x 

Conditional)

of latent cancer 

fatality for an 

individual 

located within 

10 miles 

(per reactoryear)

Long-term 

SBO 3×10-6 No Core Damage 9×10-5 ~ 3×10-10 ****

Short-term 

SBO with 

RCIC 

Blackstart**

3×10-7

No Core Damage *** 7×10-5 ~ 2×10-11  

Short-term 

SBO without 

RCIC 

Blackstart

Not Applicable ** * 2×10-4 ~ 6×10-11 ****

* The CDF assumes that 10 CFR 50.54(hh) equipment and procedures were not used.

** Blackstart of the reactor core isolation cooling (RCIC) system refers to starting RCIC without any ac or dc 

control power. Blackrun of RCIC refers to the long-term operation of RCIC without electricity, once it has 

been started. This typically involves using a portable generator to supply power to indications such as reactor 

pressure vessel (RPV) level to allow the operator to manually adjust RCIC flow to prevent RPV overfill and 

flooding of the RCIC turbine. STSBO RCIC blackstart and limited blackrun is credited as an unmitigated case 

for SOARCA purposes because the licensee has included its use in procedures. Past NRC severe accident 

analyses of STSBO scenarios did not credit blackstart of RCIC. A sensitivity calculation without blackstart was 

therefore performed to provide a basis for comparison to past analyses.

*** A scenario with 10 CFR 50.54(hh) mitigation, but without RCIC blackstart was not analyzed.

**** Estimated risks below 1 × 10-7 per reactor year 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.
77

Table 7 Surry Results for Scenarios Assuming LNT Dose-Response Model

Scenario

Core 

damage 

frequency 

[CDF]

(per 

reactoryear)*

Mitigated Unmitigated

Conditional 

scenariospecific 

probability of 

latent cancer 

fatality for an 

individual 

located within 

10 miles

Scenario-specific 

risk

[CDF x 

Conditional]

of latent cancer 

fatality for an 

individual 

located within 

10 miles 

(per reactoryear)

Conditional 

scenariospecific 

probability of 

latent cancer 

fatality for an 

individual 

located within 

10 miles

Scenario-specific 

risk

[CDF x 

Conditional]

of latent cancer 

fatality for an 

individual 

located within 

10 miles 

(per reactoryear)

Long-term 

SBO 2×10-5 No Core Damage 5×10-5 ~ 7×10-10 ****

Short-term 

SBO 2×10-6 No Containment Failure ** 9×10-5 ~ 1×10-10  

Short-term 

SBO with 

TISGTR

4×10-7 3×10-4 ** * ~ 1×10-10 **** 3×10-4 ~ 1×10-10  

Interfacing 

systems LOCA 3×10-8 No Core Damage 3×10-4 ~ 9×10-12 ** **

* The CDF assumes that 10 CFR 50.54(hh) equipment and procedures were not used.

** Accident progression calculations showed that source terms in the mitigated case are smaller than in the 

unmitigated case. Offsite consequence calculations were not run, since the containment fails at about 

66 hours. 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 ample time for mitigation through measures transported from offsite. 

*** Containment failure is delayed by about 46 hours in the mitigated case relative to the unmitigated case. 

Rounding to one significant figure shows conditional LCF probabilities of 3×10-4 for both mitigated and 

unmitigated cases, however the original values were 2.8×10-4 for the mitigated case and 3.2×10-4 for the 

unmitigated case.

**** Estimated risks below 1 × 10-7 per reactor year 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.

To provide perspective on alternative low-dose health effect modeling, the SOARCA project has 

also developed LCF risk estimates assuming non-LNT models, which are based on the premise 

that below a certain dose, cancer risks cannot be reliably quantified, or are nonexistent. Dose 

truncation values used for SOARCA included 620 mrem/year (representative background 

radiation including average annual medical exposures), and 5 rem/year with a 10-rem lifetime 

cap (based on the Health Physics Society’s position that there is a dose below which, because of
78

uncertainties, a quantified risk should not be assigned). Table 8 and Table 9 show the results of 

sensitivity calculations for dose truncation values compared with LNT results. Using these 

truncation values makes the already small scenario-specific LCF risk calculations even smaller, 

in some cases, by orders of magnitude. 

For Surry scenarios except ISLOCA, the background results in Table 9 differ from the HPS 

results, because the background truncation value clearly falls below the plant-specific population 

return criterion of 4 rem over 5 years, which is intended to represent EPA’s (adopted in Virginia) 

criterion of 2 rem in the first year and 0.5 rem/year in subsequent years; however, the HPS 

truncation value does not. The ISLOCA results are the same to one significant digit within a 

radius of 10 miles for both truncation values, because most of the emergency phase doses exceed 

both of these criteria, while, on the other hand, long-term doses make an insignificant 

contribution to the overall doses. The results in Table 8 and Table 9 assume that the probability 

of 10 CFR 50.54(hh) mitigation is zero.

SOARCA analyses included calculations of individual scenario-specific LCF risk for several

distance intervals, including 0 to 10 miles and 0 to 50 miles. This chapter presents results for 

selected distance intervals however Chapter 7 of NUREG/CR-7110, Volume 1 and 2, contains 

the results for all distance intervals in the Peach Bottom and Surry analyses. The analysis 

indicated that individual LCF risk estimates generally decrease with increasing distance, in large 

part because of plume dispersion and fission product deposition closer to the site. This trend is 

seen for all unmitigated scenarios modeled in SOARCA except for the Peach Bottom STSBO 

without RCIC blackstart and the Surry ISLOCA. More details regarding these two scenarios are 

discussed in Section 6.5 of this report and in Chapter 7 of NUREG/CR-7110, Volume 1 and 2. 

Table 8 Peach Bottom Results for Scenarios without Successful Mitigation 

for LNT and Alternative Dose-Response Models

Scenario

Scenario-specific risk of latent cancer fatality for an 

individual located within 10 miles (per reactor-year)

Linear NoThreshold BackgroundHealth Physics 

Society

Long-term SBO 3×10-10 2×10-12 1×10-12

Short-term SBO with 

RCIC Blackstart 2×10-11 2×10-13 9×10-14

Short-term SBO without 

RCIC Blackstart 6×10-11 4×10-12 4×10-12
79

Table 9 Surry Results for Scenarios without Successful Mitigation for LNT and 

Alternative Dose-Response Models

Scenario

Scenario-specific risk of latent cancer fatality for an 

individual located within 10 miles (per reactor-year)

Linear NoThreshold BackgroundHealth Physics 

Society

Long-term SBO 7×10-10 6×10-12 2×10-14

Short-term SBO 1×10-10 5×10-12 2×10-14

Short-term SBO with 

thermally induced steam 

generator tube rupture

1x10-10 3×10-11 5×10-12

Interfacing systems 

LOCA 9×10-12 2×10-12 1×10-12

Because the SBO scenarios were seismically induced, the study added analyses to evaluate the 

potential impact of the seismic event on the evacuation. Although road network infrastructure 

may be damaged during an earthquake, resulting in reduced evacuation speeds, other effects such 

as wider deployment of emergency responders and a larger shadow evacuation may improve 

evacuation timing. The analyses for both Surry and Peach Bottom indicated changes to the 

evacuation resulting from the earthquake would change the LCF risk by less than 10 percent and 

may actually cause the consequences from radionuclide release to decrease as in the case of the 

Peach Bottom plant, because the population is on alert after the earthquake.

6.4

The SOARCA offsite early fatality risk calculations are dramatically smaller than reported in 

NUREG/CR-2239

Comparison to NUREG/CR-2239 (the 1982 Siting Study SST1 case)

[1]. This Siting Study predicted 92 early fatalities for Peach Bottom and 

45 early fatalities for Surry for the SST1 source term. In contrast, SOARCA predicted that the 

early fatality risk was essentially zero for both sites. 

For LCF results, the exact basis for NUREG/CR-2239 estimates could not be recovered, but 

literature searches and sensitivity analyses with MACCS2 suggested that these estimates are for 

the population within 500 miles of the site. Moreover, an attempt to reproduce the results of 

NUREG/CR-2239 led to agreement within about a factor of 2. Given the uncertainty in the basis 

for these results, the SOARCA study performed an additional set of calculations to enable the 

current, state-of-the-art results to be compared with the 1982 Siting Study. For this set of 

calculations, the most severe source terms predicted by the SOARCA analyses (see Figure 12 

and Figure 13) were replaced by the largest source term from the Siting Study—the SST1 source 

term. No other modeling or parameter changes were made, including the timing of public 

evacuation. Thus, this comparison does not attempt to replicate the Siting Study; it simply 

evaluates the largest source term, SST1, from that study and compares the results with those 

from the current work.
80

Table 10 and Table 11 summarize the comparison to the Siting Study source term results for the 

Peach Bottom and Surry sites, respectively, assuming an LNT dose-response function. Although 

the SST1 source term is identical in both comparisons, the scenario-specific LCF probabilities 

associated with this source term shown in the tables are different because of the difference in 

evacuation modeling and other offsite consequence parameters for the two sites.

Table 10 Conditional (i.e., assuming accident occurs), Mean, LNT, Scenario-Specific 

Probabilities of LCF for People within the Specified Radii of the Peach Bottom 

Site

Radius of

Circular Area (mi)

1982 Siting 

Study SST1

SOARCA

Unmitigated

STSBO

10 3.3×10-3 2.1×10-4

20 1.8×10-3 5.7×10-4

50 4.6×10-4 1.9×10-4

Table 11 Conditional, Mean, LNT, Scenario-Specific Probabilities of LCF for People 

within the Specified Radii of the Surry Site 

Radius of

Circular Area (mi)

1982 Siting 

Study SST1

SOARCA

Unmitigated 

ISLOCA

SOARCA 

Unmitigated 

STSBO with 

TISGTR

10 1.0×10-2 3.0×10-4 3.2×10-4

20 5.1×10-3 3.4×10-4 1.9×10-4

50 1.5×10-3 1.6×10-4 6.5×10-5

For the 0-10 mile radius, the area associated with the NRC Safety Goal for latent cancers, the 

scenario-specific probabilities of latent cancer fatality calculated for SOARCA are substantially 

smaller than predicted in the 1982 Siting Study for SST1. Considering both the Peach Bottom 

and Surry comparisons, this difference diminishes with increasing radius, falling from a factor of 

33 within 10 miles to a factor of about 2.4 within a 50-mile radius and beyond. The 

quantification of latent cancer fatality risk is more uncertain at low doses, such as at far 

distances, and the diminishing difference in conditional LCF risk can be explained as an 

influence of competing factors. For instance, while the SOARCA radionuclide releases are 

significantly smaller than the 1982 Siting Study SST1 case, they are also calculated to persist to 

longer distances because the effective deposition velocity is lower than assumed in the 1982 

Siting Study. A farther spread of radioactive material is more likely to create lightly 

contaminated, habitable areas, in which people are assumed to receive very small doses over an 

extended period of time, as well as very low LCF risk as calculated using the LNT dose-response 

model. Previous studies, with larger releases and plumes that deposited over shorter distances, 

were more likely to predict either uninhabitable or unaffected areas, both of which are areas 
81

where individuals do not receive elevated LCF risk. In addition, the cancer risk factors from 

radiation are estimated to be higher than previously thought [56] and the inventory of fission 

products has been updated to account for greater burnup and power levels. This combination of 

modeling differences between SOARCA and the 1982 Siting Study accounts for the fact that the 

calculated risks are similar for populations farther from the site.

Figure 14 provides additional comparisons of SOARCA results for both mitigated and 

unmitigated scenarios to 1982 Siting Study SST1 results for people within 10 miles of the plant.

Figure 14 Comparison of average individual LCF risk results for SOARCA mitigated and 

unmitigated scenarios to the NRC Safety Goal and to extrapolations of the 1982 

Siting Study SST1 (plotted on logarithmic scale)
82

6.5 Sensitivity Analyses on the Size of the Evacuation Zone

As discussed earlier in Section 5.3.2, SOARCA included sensitivities to analyze the effects of 

increased evacuation sizes on scenario-specific latent cancer fatality risks for people within 

different distances of each plant. These sensitivities were performed for distances of 16 miles 

and 20 miles from the plant and compared to the base case of a 10 mile radius evacuation. 

Selected results are presented below in Table 12 and Figures 15 and 16; however more details 

are provided in NUREG/CR-7110, Volumes 1 and 2. The Peach Bottom unmitigated STSBO 

scenario without RCIC blackstart and the Surry unmitigated ISLOCA scenario were chosen for 

these sensitivities because they result in the largest releases of radioactive materials of all 

scenarios analyzed for each plant. The Peach Bottom unmitigated STSBO scenario without 

RCIC blackstart releases about 12% of the core inventory of I-131 and 2% of the Cs-137 to the 

environment and begins at 8 hours. The Surry unmitigated ISLOCA scenario releases about 

16% of the core inventory of I-131 and 2% of the Cs-137 and begins at about 13 hours. 

Table 12 Effect of Size of Evacuation Zone on Mean, Individual, LNT, Scenario-Specific 

LCF Risk for People within the Specified Radii of the Plant for the Peach Bottom 

Unmitigated STSBO without RCIC Blackstart and the Surry Unmitigated 

ISLOCA

Radius of 

Circular Area 

(mi)

Peach Bottom

Unmitigated STSBO

without RCIC Blackstart

Surry

Unmitigated ISLOCA

Base Case

10-mile 

Evacuation

Sensitivity

20-mile 

Evacuation

Base Case

10-mile 

Evacuation

Sensitivity

20-mile 

Evacuation

10 6×10-11 2×10-10 9×10-12 1×10-11

20 2×10-10 7×10-11 1×10-11 8×10-12

30 1×10-10 8×10-11 8×10-12 7×10-12

40 7×10-11 6×10-11 6×10-12 5×10-12

50 6×10-11 5×10-11 5×10-12 4×10-12
83

Figure 15 Mean, individual, LNT, scenario-specific LCF risk for the Peach Bottom 

unmitigated STSBO scenario without RCIC blackstart for people within a 

circular area of specified radius from the plant.

Figure 16 Mean, individual, LNT, scenario-specific LCF risk for the Surry unmitigated 

ISLOCA scenario for people within a circular area of specified radius from the 

plant.
84

Analysis of the evacuation size sensitivities for both Peach Bottom and Surry shows that for the 

base case, LCF risk is slightly higher for people within a 20 mile radius of the plant compared to 

people within a 10 mile radius of the plant. This result is likely due to SOARCA’s modeling 

assumptions rather than physical reality. SOARCA models a shadow evacuation, which is the 

voluntary (self-initiated) evacuation of members of the public from areas that are not under 

official evacuation orders, yet it does not model an evacuation that (for which there is no 

preplanning) would be ordered by officials for certain areas outside the EPZ. Therefore, the 

slight increase in LCF risk in the base case for people within a 20 mile radius of the plant relative 

to people within a 10 mile radius is not considered a meaningful insight.

Analysis of Figure 15 and 16 also shows that expanding the evacuation size from a 10 mile 

radius to a 20 mile radius results in increased LCF risk for people in the 0-10 mile area. 

SOARCA analyses show that an evacuation beyond the area closest to the plant will delay those 

most at risk, i.e., closest to the plant. The increased risk to the population within 10 miles of the 

plant is due to slower evacuation speeds because of additional traffic congestion and delays that 

result from evacuation of a larger population. 

Further, SOARCA’s evacuation size sensitivities show that an evacuation out to 20 miles from 

the plant results in decreased risk for the population relative to the base case 10 mile evacuation. 

For the Peach Bottom unmitigated STSBO without RCIC blackstart, the scenario-specific LCF 

risk falls from 2 x10-10 per reactor-year to about 7 x10-11, a factor of about 3, while the risk 

reduction is much smaller for the Surry unmitigated ISLOCA scenario. The decrease in average 

LCF risk out to 20 miles is seen as within the bounds of modeling assumptions and not indicative 

of a measurable benefit. For the 20 mile evacuation, SOARCA did not model a shadow 

evacuation; however if this was included, it would likely delay the evacuation of the people 

within 20 miles and increase the scenario-specific LCF risk. This likely leads to an 

underestimation in scenario-specific LCF risk for the people within 20 miles during the 20 mile 

evacuation. 

Overall, the increases and reductions to scenario-specific LCF risk shown in Table 12 and Figure 

15 and 16 are extremely small on an absolute scale. The LCF risks calculated for SOARCA’s 

base case and 20-mile evacuation sensitivity are all millions of times smaller than the average 

annual risk of cancer death for an individual in the United States.

6.6 Conclusions

The results of the SOARCA project represent a major change in the staff’s perception of severe 

reactor accidents and their consequences. Specific conclusions of the project are as follows:

The SOARCA results demonstrate the potential benefits of employing 10 CFR 50.54(hh) 

mitigation enhancements for the scenarios analyzed. When successful mitigation is assumed, the 

MELCOR results indicate no core damage for all scenarios except the Surry STSBO and its 

TISGTR variant. For the Surry STSBO with mitigation, the core is damaged; however,

containment failure is delayed by an additional 41 hours compared to the unmitigated case. The 
85

mitigation measures (i.e., containment sprays) are effective in knocking down the airborne 

aerosols. For the Surry STSBO with TISGTR with mitigation, the core is damaged and 

containment failure is delayed by an additional 46 hours compared to the unmitigated case. This 

is a bypass scenario, and therefore the release to the environment begins at the same time as in 

the unmitigated case. For both the mitigated and unmitigated cases, the individual scenariospecific LCF risk for the EPZ was small, approximately 1×10-10 per reactor-year, assuming an 

LNT dose-response model. 

When the selected SOARCA scenarios were assumed to proceed unmitigated (i.e., neither

10 CFR 50.54(hh) implementation nor other key operator actions that would prevent core 

damage), MELCOR analyses indicated that the accidents progress more slowly and with smaller 

releases than the 1982 Siting Study SST1. Whereas the 1982 Siting Study SST1 case results in a 

large early release at 1.5 hours, the SOARCA analyses show no large early releases for the 

scenarios analyzed.

The individual early fatality risk from SOARCA scenarios is essentially zero. Individual LCF 

risk from the selected specific, important scenarios is thousands of times lower than the NRC 

Safety Goal and millions of times lower than the general cancer fatality risk in the United States 

from all causes, even assuming the LNT dose-response model. Using a dose-response model 

that truncates annual doses below normal background levels (including medical exposures)

results in a further reduction to the LCF risk (by a factor of 100 for smaller releases and a factor 

of 3 for larger releases). LCF risk calculations are generally dominated by long-term exposure to 

small annual doses (about 500 mrem per year) corresponding to evacuees returning to their 

homes after the accident and being exposed to residual radiation over a long period of time.

SOARCA results indicate that bypass events (e.g., Surry ISLOCA) do not pose a higher 

scenario-specific latent cancer fatality risk than non-bypass events (e.g., Surry SBO). While 

consequences are greater when the bypass scenario happens, this is offset by the scenario being 

less likely to happen. SOARCA reinforces the importance of external events relative to internal 

events and the need to continue ongoing work related to external events risk assessment.

The SOARCA analyses show that emergency response programs, implemented as planned and 

practiced, reduce the scenario-specific risk of health consequences among the public during a 

severe reactor accident. Sensitivity analyses of seismic impacts on site-specific emergency 

response (e.g., loss of bridges, traffic signals, and delayed notification) at Peach Bottom and 

Surry do not significantly affect LCF risk. 

SOARCA results, while specific to Peach Bottom and Surry, may be generally applicable for 

plants with similar designs. However, additional work is needed to confirm this, since 

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

 

87

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A-1

APPENDIX A

SOARCA AND THE FUKUSHIMA DAIICHI ACCIDENT

Objective

The State-of-the-Art Reactor Consequence Analyses (SOARCA) study was nearly at the end of 

its peer review when the Fukushima Daiichi accident occurred on March 11, 2011. Following 

the accident, the U.S. Department of Energy (DOE) and the U.S. Nuclear Regulatory 

Commission (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 has many similarities 

and differences with some of the Peach Bottom sequences analyzed in SOARCA. The objective 

of this appendix is to compare and contrast 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 (SFP) risk. It must be emphasized that there are 

significant gaps in information and uncertainties about what actually occurred in the Fukushima 

reactors. These uncertainties do not allow firm conclusions on comparisons with SOARCA 

results. It is expected to take a number of years for the Japanese organizations involved to be 

able to access the containments and fully evaluate the conditions of the nuclear fuel and other 

equipment to allow a more complete understanding of the events. 

Background

The Great East Japan Earthquake, which rated a magnitude 9.0 on the moment magnitude scale

(Mw), occurred northeast of Tokyo off the east coast of Honshu Island. This earthquake resulted 

in the automatic shutdown of the Fukushima Daiichi reactors. The earthquake precipitated a 

tsunami that exceeded 14 meters (45 feet) in height at the Fukushima Daiichi site. The 

earthquake and subsequent tsunami produced widespread devastation across northeastern Japan, 

resulting in approximately 25,000 people dead or missing, displacing many tens of thousands of 

people, and significantly affecting the infrastructure and industry in the northeastern coastal 

areas of Japan.

On March 11, 2011, Fukushima Daiichi Units 1, 2, and 3 were in operation, and Units 4, 5, and 6 

were shut down for routine refueling and maintenance activities; the Unit 4 reactor fuel was 

offloaded to the Unit 4 SFP. The description of events below is based on our current 

understanding of the accident, which, as previously stated, is subject to significant uncertainty.

As a result of the earthquake, all of the operating units appeared to experience a normal reactor 

trip within the capability of the safety design of the plants. The three operating units (Units 1, 2, 

and 3) automatically shut down, inserting all control rods into the respective reactors. Also, as a 

result of the earthquake, offsite power was lost to the entire facility. The emergency diesel 
A-2

generators started at all six units providing alternating current (ac) electrical power to critical 

systems, and the facility response to the seismic event appears to have been normal.

Approximately 40 minutes after the earthquake and shutdown of the operating units, a large 

tsunami wave inundated the site, followed by multiple additional waves. The estimated height of 

the tsunami exceeded the height for which site protection features against tsunamis were 

designed by approximately 8 meters (27 feet). The tsunami resulted in extensive damage to site 

facilities and a complete loss of ac electrical power at Units 1 through 5 (i.e., a station blackout 

(SBO)). Unit 6 retained the function of one of the diesel generators.

Without ac power, the plants relied on batteries and turbine-driven and diesel-driven pumps for 

reactor core cooling (it should be noted that immediately after the tsunami, Units 1 and 2 were 

without 125 volt dc power too). The operators took actions to maintain core cooling functions 

well beyond the normal capacity of the station batteries. However, without sufficient offsite 

assistance, which appears to have been hampered by the devastation in the area, among other 

factors, Units 1 through 3 eventually lost the ability to further extend cooling of the reactor cores. 

This ultimately resulted in significant damage to the reactor cores in these units, the extent of 

which is still the subject of evaluation.

At varying points in time after the tsunami, Units 1, 3, and 4 experienced explosions, further 

damaging the facilities and containment and reactor buildings. The Unit 1 and 3 explosions were

apparently caused by the buildup of hydrogen gas within containment produced during fuel 

damage in the reactor and subsequent movement of that hydrogen gas from the drywell into the 

reactor building. The explosion that occurred in Unit 4 may have involved hydrogen that was 

transported through a ventilation system connected to Unit 3. 

As information about the damage to plant safety functions was gathered over the weeks and 

months following these events, many similarities became apparent between the calculated 

damage progression in the boiling-water reactor (BWR) SBO accident scenarios in the SOARCA 

analyses 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 

onset of core damage and fission product release from fuel, 

• challenges to containment integrity that accompanied the loss of decay heat removal and 

the accumulation of hydrogen generated during in-vessel damage to reactor fuel, and

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

As noted in the discussion of hydrogen combustion below, the SOARCA analyses and the 

Fukushima events appear to have released hydrogen to the reactor building by different 

mechanisms. But in both cases, the end result was structural failure of the building and 

radionuclide release to the environment caused by energetic combustion. Similarities were also 

observed in characteristics of radionuclide release to the environment in the SOARCA 

calculations and early measurements of activity in the areas surrounding the Fukushima site.
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Some notable differences in the events that unfolded at Fukushima and the BWR long-term SBO

(LTSBO) scenario studied in the SOARCA project were also readily apparent. These differences 

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

analyses. The NRC’s SOARCA team qualitatively compared the results from the SOARCA 

analyses to the preliminary events and information available at this early stage in the evaluation 

of the Fukushima Daiichi accident, and the results are discussed below.

The information used to evaluate these topics was gleaned from a variety of sources. Most 

important among these are the following:

• “Report of the Japanese Government to the IAEA [International Atomic Energy Agency]

Ministerial Conference on Nuclear Safety,” June 2011 [[1]]

• “Special Report on the Nuclear Accident at the Fukushima Daiichi Nuclear Power 

Station,” Institute of Nuclear Power Operations (INPO), November 2011 [7]

Shortly after the accident, the NRC established a task force to conduct a methodical and 

systematic review of the agency’s processes and regulations to determine whether it should make 

additional improvements to its regulatory system. The task force report, “Recommendations for 

Enhancing Reactor Safety in the 21st Century” [[2]], found, among other things, that prolonged 

SBO and multiunit events present challenges to emergency response. The task force report 

presented a number of recommendations that address physical, administrative, and regulatory 

enhancements to further reduce the risk of similar challenges occurring among the U.S. fleet of 

nuclear power plants. 

Operation of the Reactor Core Isolation Cooling System

According to the “Report of Japanese Government to IAEA Ministerial Conference on Nuclear 

Safety” [[1]], and the Institute of Nuclear Power Operations (INPO) “Special Report on the 

Nuclear Accident at the Fukushima Daiichi Nuclear Power Station” [[7]], the RCIC system was 

used to maintain coolant injection to the reactor pressure vessel (RPV) for approximately 

70 hours in Unit 2 and for 21 hours in Unit 31

 1 When the RCIC system tripped in Unit 3, a separate steam-driven high-pressure coolant injection system (HPCI) 

automatically started and ran for approximately 2 hours. Unit 1 has a different system (isolation condenser) and does 

not have the RCIC system.

. The SOARCA study performed MELCOR 

analyses for the Peach Bottom station blackout scenarios shown in the following table.
A-4

Start of RCIC 

operation (hours)

End of RCIC 

operation (hours)

Duration of RCIC 

operation (hours)

Fukushima Unit 2 0 70 70

Fukushima Unit 3 0 21 21

Peach Bottom unmitigated LTSBO 0 5 5

Peach Bottom unmitigated STSBO 

with RCIC blackstart 1 3 2

Peach Bottom unmitigated STSBO 

without RCIC blackstart Not modeled Not modeled N/A

Note: The MELCOR analysis was truncated at 48 hours, as discussed below.

The operators at Fukushima Units 2 and 3 were able to successfully operate their RCIC systems

to maintain water inventory within the core for a period of time that greatly exceeded the 

operating period assumed in the SOARCA calculations of the unmitigated LTSBO scenario. 

In the SOARCA analysis of the unmitigated LTSBO scenario, the RCIC system operates for 

4 hours while direct current (dc) power is available and an additional 1.2 hours after station 

batteries are exhausted2 (i.e., RCIC blackrun). Measurement of the reactor water level would be 

lost at 4 hours, when station batteries that provide dc power to critical plant instrumentation are 

exhausted. This condition led to the assumption that manual operation of RCIC would maintain 

a constant (throttled) flow rate with the goal of maintaining the reactor coolant water level. 

However, as discussed in the detailed description of the MELCOR analysis of the unmitigated 

LTSBO scenario, sustained (constant flow) operation of RCIC after the loss of dc power would 

lead to an increase in the RPV water level and steamline flooding in approximately 1.2 hours. 

Flooding the main steamline would result in flooding the RCIC turbine, disabling the system. 

Overfilling the RPV is a consequence of the imbalance between the termination of coolant losses 

through safety relief valves (which reclose on a loss of dc power) and continued coolant addition 

by the RCIC system.

Flooding of the RCIC turbine by reactor vessel overfill does not appear to have occurred at 

Fukushima, and the operators successfully ran the system for an extended period of time in 

Units 2 and 3. One reason for the extended length of RCIC operation at Fukushima, in 

comparison to the unmitigated LTSBO timeline in the SOARCA analysis, is the station batteries 

at Fukushima were designed to provide dc power for a longer period of time than the batteries at 

Peach Bottom (8 hours for Fukushima versus 2 hours). At both plants, the actual duration of dc 

power would be longer than the design basis because of margins incorporated into the system 

design, as well as manual actions that can be taken to shed nonessential loads on the dc 

emergency bus. The maximum length of time that dc power was available at Fukushima appears 

to have been considerably longer than the maximum battery duration considered in the SOARCA 

analysis for Peach Bottom, even when load shedding is taken into account. 

A second reason for the difference in RCIC operating time is that manual actions taken at 

Fukushima to manage RCIC operation after the loss of dc power appear to have differed from 

 2 RCIC is a steam-driven coolant injection system that does not require ac power for it to start or operate as a coolant 

injection system. Steam flow to the RCIC turbine can be remotely controlled with dc power from the station batteries. 

However, manual operation using valve handwheels (i.e., blackstart, blackrun) is also possible.
A-5

those assumed in the unmitigated LTSBO scenario. The precise actions taken by TEPCO 

operations personnel to control RCIC flow in Units 2 and 3 are not known. However, it is clear 

that RCIC flow at Fukushima Units 2 and 3 was regulated at values that prevented overfill of the 

RPV and flooding of the main steamlines. The INPO report [7] indicates portable electric 

generators were installed and power was restored to critical plant instrumentation in the control 

room, including RPV water-level measurement. This action facilitated control of RCIC flow to 

maintain the RPV level at desired values. The SOARCA analysis of the unmitigated LTSBO 

sequence did not credit staging and alignment of portable electric generators (i.e., equipment 

required in Title 10 of the Code of Federal Regulations (10 CFR) 50.54(hh)). The 

10 CFR 50.54(hh) equipment at Peach Bottom also includes portable coolant pumps, which, 

according to the SOARCA analyses, would satisfactorily maintain core cooling and avert core 

damage if aligned and operated successfully. Therefore, differences in operator actions to 

manage RCIC flow after the loss of dc power versus those assumed for SOARCA, as well as 

differences in the availability of portable mitigation equipment, contributed to differences in the 

timeline of events at Fukushima versus the timeline calculated for the unmitigated LTSBO 

scenario.

Finally, the RCIC system in Fukushima Units 2 and 3 appears to have run for many hours under 

conditions that exceed established operating limits for the turbine-driven pumps. The reason that 

the RCIC pumps eventually stopped running is not known. However, preliminary DOE/NRC 

forensic analysis of event progression at Units 2 and 3 indicate the torus water temperature in 

both units exceeded values that would have challenged pump operation because of loss of 

adequate net positive suction head, vibration and mechanical damage from pump cavitation, or 

overheating of pump bearings caused by inadequate cooling. The SOARCA models did not 

anticipate nor incorporate sustained endurance of the RCIC system (well beyond design limits).

These differences in the factors contributing to the duration of RCIC operation at Fukushima and 

the SOARCA analyses result in two differences in the observed chronology of events that follow 

the eventual loss of coolant injection. First is the difference in the times at which core damage 

and fission product release to the environment begin. These events were predicted to begin at 

20 hours in the SOARCA unmitigated LTSBO analysis, but they began at Fukushima Units 2 

and 3 on the third and second day of the accident, respectively. Second, sustained operation of 

the RCIC system at Fukushima resulted in a larger cumulative transport of heat from fission 

product decay (in the form of steam) from the RPV to the suppression pool in the containment 

(torus). Suppression pool temperatures at the time the RCIC pumps ceased operating in 

Fukushima Units 2 and 3 were, therefore, much higher than the calculated pool temperature in 

the SOARCA analysis of the unmitigated LTSBO scenario. Increases in suppression pool 

temperature result in additional evaporation of water from the pool to the containment 

atmosphere; this in turn results in an increase in containment pressure. Therefore, containment 

pressure in the Fukushima reactors at the time core damage began was higher than the pressure 

calculated in the Peach Bottom LTSBO scenario. When hydrogen, generated by oxidation of 

Zircaloy cladding in the core, was released to the containment atmosphere, containment pressure 

increased further. The combination of a high base pressure from long-term evaporation of steam 

and accumulation of noncondensible hydrogen gas in the Fukushima containments likely 

resulted in pressures that were sufficiently high to induce leakage through the drywell head 

flange while in-vessel core damage was underway. Release of hydrogen to the reactor building 
A-6

through the drywell head flange likely led to the destruction of the Fukushima reactor buildings 

by hydrogen combustion. In contrast, the shorter duration of RCIC operation in the unmitigated 

LTSBO scenario resulted in less heating of the suppression pool, less evaporation of water to the 

containment atmosphere, and a lower base pressure in containment at the time core damage and 

hydrogen generation began. 

The extended period of core cooling by sustained RCIC operation at Fukushima affected more 

than the timeline for core damage and containment pressure at the time core damage began. The 

mechanisms for hydrogen (and fission product) leakage out of containment into the reactor 

building were affected by differences in containment thermodynamic conditions, which were 

influenced by the operation of the steam-driven RCIC system.

Hydrogen Release and Combustion

The physical damage to Fukushima reactor buildings will perhaps be the most enduring visible 

image of plant damage initiated by the earthquake and tsunami in Japan in March 2011. The 

apparent cause was combustion of hydrogen that was generated by high-temperature oxidation of 

fuel cladding. Extensive cladding oxidation and core material melting is believed to have 

occurred in Fukushima Units 1, 2, and 3, although the timelines for core damage differed in each 

unit because of differences in equipment and operator response. 

At Fukushima Units 1 and 3, hydrogen generated from the oxidation of fuel cladding in the core 

was likely transported from the RPV to the containment (drywell and wetwell) through an open 

or cycling safety relief valve. Hydrogen is predicted to be released to the containment by the 

same pathway in the unmitigated SOARCA scenarios. The precise pathway by which hydrogen 

was released from the containment to the reactor building at Fukushima is uncertain. The 

Japanese Report to the IAEA [[1]] suggests the pathway was leakage through the drywell head 

flange, which is normally sealed by a pair of O-ring seals. High internal pressures developed 

within the drywell as a consequence of the failure of engineered systems for containment heat 

removal from loss of ac power from various causes and the accumulation of noncondensible 

hydrogen. The resulting mechanical loads transmitted to the drywell head and closure bolts are 

believed to have resulted in leakage past the closure seals, through the head flange to the upper 

portion of the reactor building. The Japanese government report [[1]] suggests this leakage 

pathway developed in all three units in which core damage occurred (Units 1, 2, and 3).

Leakage across the drywell head flange is modeled in the SOARCA scenarios. Opening criteria 

for this leak pathway are based on NRC calculations of the internal pressure required to cancel 

the compressive force on the closure head flange created by the torque applied to the head bolts. 

Only one of the SOARCA calculations (unmitigated LTSBO) resulted in a drywell pressure 

sufficiently high to open this release pathway before drywell liner melt-through. In this case, the 

liner melt-through occurs shortly after head flange leakage begins. As a result, in the SOARCA 

scenarios, significant hydrogen release to the reactor building occurs only after mechanical 

failure of the containment pressure boundary, which in the SOARCA calculations results from 

molten debris failing the drywell liner after RPV lower head failure (i.e., drywell liner 

melt-through). Containment failure by this mechanism is not believed to have occurred in any of 

the units at Fukushima. However, the Japanese government report suggests the possibility of 
A-7

some amount of molten debris being released from the RPV lower head to the drywell floor 

within the reactor pedestal in Units 2 and 3. Additionally, TEPCO has announced that a recent 

analysis of Unit 1 (accident simulations performed with the MAAP computer code) suggest the 

bulk of the fuel was released into the drywell through RPV lower head failure [1]. TEPCO 

concluded from this analysis that fuel released to the drywell floor eroded approximately 

70 centimeters (2.3 feet) of concrete on the drywell floor within the reactor pedestal, but the fuel 

did not move laterally across the drywell floor. 

The BWR MELCOR model used in the SOARCA calculations ignites and burns hydrogen in 

regions of the reactor building where local concentrations satisfy assumed flammability criteria

(see NUREG/CR-7110, Volume 1). The pressure generated by hydrogen combustion within the 

reactor building results in opening the blowout panels in the walls of the refueling floor. In the 

station blackout scenarios examined in the SOARCA calculations, the combustion pressure is 

sufficiently high to also fail (open) many doorways within the building (e.g., into and out of the 

stairwells) and the large railroad access doorways to the environment at grade level. Structural 

failure of the steel roof of the reactor building can also occur, if these pathways are insufficient 

to relieve the internal pressure generated by hydrogen combustion. Roof failure was calculated 

to have occurred in the SOARCA short-term SBO (STSBO) scenario but not in the LTSBO

scenario.

Generation of hydrogen from oxidation of fuel stored in the SFP of Fukushima Unit 4, which 

was shut down for maintenance at the time of the accident, is not believed to have occurred in 

significant quantities. However, the Unit 4 reactor building was severely damaged, apparently 

by the combustion of hydrogen that leaked into the building. It has been proposed that the 

source of hydrogen in the Unit 4 reactor building was hydrogen that flowed through piping that 

connects the standby gas treatment system (SGTS) in Unit 3 to a parallel system in Unit 4. 

Hydrogen may have entered the SGTS system in Unit 3 when containment venting was 

performed, because the containment vent system at Fukushima allows the operator to direct 

gases through the SGTS for filtration before being released to the stack3

. Although venting 

through a similar pathway is possible at Peach Bottom (e.g., opening the containment ventilation 

system, which is connected to the SGTS), a different release pathway for containment venting is 

simulated in the SOARCA calculations for Peach Bottom4

 3 The precise configuration of the containment vent pathway used at Fukushima is not clear in terms of its discharge 

location relative to the SGTS filters. The INPO report describes a configuration that bypasses SGTS. However, as 

noted earlier, SGTS ductwork is believed to be the path by which hydrogen flowed from Unit 3 into Unit 4.

. Virtually all BWR Mark I 

containments in the United States have a hardened vent that bypasses the normal containment 

ventilation system and associated SGTS and vents the containment atmosphere directly to the 

environment. This pathway bypasses the SGTS filters and is comprised of rigid piping rather 

than the thin metallic ductwork common to building ventilation systems. As a result, hydrogen 

gas would be discharged to the environment, rather than leaking to the reactor building through 

leaks or ruptures in ventilation duct work, which would not likely survive the internal pressure 

4 Selection of a containment venting pathway is a proceduralized action at Peach Bottom and includes an assessment of 

potential adverse characteristics of each pathway. For example, venting through the drywell (or wetwell) ventilation 

system to SGTS could adversely affect the environment in the reactor building if relatively weak ventilation ductwork 

were to fail because of high internal pressure, releasing hydrogen and radioactivity into the reactor building. This could 

cause accessibility issues for other operator actions.
A-8

anticipated if containment venting were to become necessary. Therefore, hydrogen leakage to 

the reactor building would not occur as a result of containment venting if the hardened vent is 

used, as assumed in the SOARCA models. While hardened vents that allow the operator to 

bypass the SGTS filters were installed at Fukushima between 1999 and 2001, it is unclear 

whether they were used during the March 2011 events.

48-Hour Truncation of Releases in SOARCA

The 48-hour truncation time for SOARCA was based on the many resources available at the 

State, regional, and national level that would be available to mitigate a severe reactor accident. 

The staff reviewed available resources and emergency plans and determined that adequate 

mitigation measures (at minimum, the ability to flood the reactor building) could be brought 

onsite within 24 hours and connected and functioning within 48 hours. The decision to truncate 

releases at 48 hours (72 hours for the Surry LTSBO) was made well before the Fukushima 

accident. Based on the assumptions made for SOARCA, the releases that would occur within 

48 hours for the Peach Bottom unmitigated scenarios cease because of reactor building flooding. 

For Fukushima, as discussed above, the operators delayed releases beyond the SOARCA 

assumption, so substantial releases occurred beyond 48 hours. In addition, the operators at 

Fukushima were not able to flood the reactor buildings, as assumed for SOARCA. 

For mitigated cases, the SOARCA analysis assumed the effectiveness of mitigation measures 

well within 48 hours. This assumption is considered reasonable, given the vast network of 

resources available in the United States. These resources include an offsite emergency 

operations facility, which would provide access to fleetwide emergency response personnel and 

equipment, including the 10 CFR 50.54(hh) mitigation measures and equipment from sister 

plants. These assets, as well as those from neighboring utilities and State preparedness 

programs, could be brought to bear on the accident if needed. In addition, SOARCA did not 

assume a tsunami, and such an event is considered highly unlikely at Peach Bottom and Surry. If 

sites were subject to tsunamis, these events could affect the availability and effectiveness of 

mitigation measures. In response to the recommendation of the NRC’s Near Term Task Force 

report, SECY-11-0093, dated July 12, 2011, the NRC is currently evaluating whether changes to 

mitigation strategies are warranted. 

Multiunit Risk

As demonstrated by the Fukushima accident, severe accidents that affect multiple reactors 

located at a common site are possible. Such accidents may happen following an initiating event 

that simultaneously challenges all reactors (e.g., earthquakes, tsunamis, loss of the electrical 

power grid) or following an accident in a single reactor that cascades to other reactors through

interconnected electric power or cooling water systems, or inaccessibility to areas of the plant 

because of an ongoing radioactive release at one unit. An example of physical interactions for 

multiunit risk is the ad hoc installation of a temporary power cable from a mobile electric power 

supply to the standby liquid control pump in the Fukushima Unit 2 reactor building [[7]]. 

TEPCO personnel completed their work to install this equipment minutes before the explosion in 

the Unit 1 reactor building occurred. Debris generated by the explosion in Unit 1 damaged the 
A-9

temporary cables and the power supply vehicle in Unit 2, defeating earlier actions to recover the 

standby liquid control pump as a resource for high-pressure coolant injection. 

Although beyond the scope of the SOARCA project, the NRC staff previously recognized the 

potential risks of multiunit accidents and has taken steps to further analyze them. As a result of 

the SOARCA analyses, the NRC established a generic issue to further consider the implications 

of multiunit accidents. Subsequently, in SECY-11-0089 [[3]], the staff proposed a site-wide 

Level 3 probabilistic risk assessment (PRA) that included an analysis of multiunit accidents 

initiated by internal and external causes during any plant operating mode. The scope of this 

proposed analysis includes all spent fuel stored onsite either in SFPs or in dry casks in addition 

to all reactors. The proposed analysis would assess the radiological consequences from multiple 

releases that may occur at separate times. In its staff requirements memorandum dated 

September 21, 2011 [[3]], the Commission directed the staff to complete the site-wide Level 3 

PRA project within 4 years.

Spent Fuel Pool Risk

The SOARCA analyses did not include impacts on the spent fuel pools (SFPs) for either Peach 

Bottom or Surry. However, various recent risk studies, most recently NUREG-1738, “Technical 

Study of Spent Fuel Pool Accident Risk at Decommissioning Nuclear Power Plants,” issued 

February 2001 [[4]], have shown that storage of fuel in a high-density configuration in SFPs is 

safe and that the risk of a significant release of radioactive materials as a result of loss of SFP 

cooling is expected to be less than reported in previous studies. More advanced analyses of SFPs 

have been conducted as part of the NRC’s post-9/11 security assessments. However, these are 

not publicly available because of their sensitive nature. The agency has since restated its views 

that spent fuel is stored safely in high-density configurations in a response to SECY-08-0036,

“Denial of Two Petitions for Rulemaking Concerning the Environmental Impacts of HighDensity Storage of Spent Nuclear Fuel in Spent Fuel Pools (PRM [Petition for 

Rulemaking]-51-10 and PRM-51-12),” dated June 19, 2008 [[5]], as well as the revision to 

NUREG-1437, “Generic Environmental Impact Statement for License Renewal of Nuclear 

Plants—Draft Report for Comment,” issued July 2009 [[6]]. Partly because of changes in the 

path forward of the planned Yucca Mountain geologic repository and the Fukushima accident, 

interest in the safety of spent fuel storage has recently increased. Therefore, the NRC has 

commissioned an SFP scoping study, which started in 2011, aimed at updating the estimation of 

radiological consequences of a severe accident on SFPs with both high-density and low-density 

storage configurations. The scenario considered in the study is a beyond-design-basis seismic 

event in the range of 0.5 to 1 g peak ground acceleration. The study involves seismic and 

structural analysis of the earthquake and its effects on the SFP; thermal-hydraulic and severe 

accident progression modeling with the MELCOR computer code; emergency preparedness and 

response; and, finally, offsite consequence analyses with the MACCS2 code. The plan is to 

document the results of the study in a publicly available report within the next year.

In the analyses presented in this report, hydrogen produced by oxidation of Zircaloy, whether 

produced in-vessel during core degradation, or ex-vessel by core-concrete interactions, is 

predicted to be burned in compartments of the reactor building as released via the failure of the 

drywell liner by melt-attack. These burns occur as flammability conditions are attained, mainly 
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as hydrogen concentrations increase to the point that combustion can occur. The burns produce 

sufficient building over pressure to open the refueling bay blowout panels and blow open doors 

in the building. The explosions that were observed in the accidents at Fukushima were 

significantly larger than predicted in the SOARCA analyses, perhaps involving detonations 

where ignition might have taken place at higher concentrations than predicted in SOARCA. The 

damage to the Fukushima Unit 3 refueling bay was especially significant with building debris 

(steel and concrete) falling into the spent fuel pool, also located in the refueling bay. The debris 

observed in the Unit 3 spent fuel pool may well have mechanically damaged some of the fuel 

assemblies stored there; however, isotopic analysis of the pool water performed by TEPCO for 

radioactive contamination does not suggest any significant releases from the fuel rods. Moreover, 

the water of the spent fuel pool provides massive scrubbing capability for any released fission 

products such that this potential source for environmental release becomes vanishingly small. 

The structural damage could, on the other hand, present engineering challenges for maintaining 

long term cooling of the fuel stored in the pool in the days and weeks following the accident. For 

these reasons, SOARCA did not consider source terms from ancillary damage to the spent fuel 

pool from hydrogen deflagrations.
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References

[1] Nuclear Emergency Repose Headquarters - Government of Japan, "Report of the 

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

Accident at TEPCO's Fukushima Nulcear Power Stations," IAEA Ministerial Conference 

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

[2] Miller, C., et al., "Recommendations for Enhancing Reactor Safety in the 21st Century -

The Near-Term Task Force Review of Insights from the Fukushima Daiichi Accident," 

SECY-11-0093, Nuclear Regulatory Commission, Washington DC, July 12, 2011. 

[3] U.S. Nuclear Regulatory Commission (NRC), "Options for Proceeding with the Future 

Level 3 Probabilistic Risk Assessment Activities," SECY-11-0089, Nuclear Regulatory 

Commission, Washington DC, September 21, 2011.

[4] U.S. Nuclear Regulatory Commission, "Technical Study of Spent Fuel Pool Accident 

Risk at Decommissioning Nuclear Power Plants," NUREG-1738, U.S. Nuclear 

Regulatory Commission, Washington DC, Feburary, 2001.

[5] U.S. Nuclear Regulatory Commission, "Denial of Two Petitions for Rulemaking 

Concerning the Environmental Impacts of High-Density Storage of Spent Nuclear Fuel in 

Spent Fuel Pools (PRM-51-10 and PRM-51-12)," SECY-08-0036, Nuclear Regulatory 

Commission, Washington DC, June 19, 2008.

[6] U.S. Nuclear Regulatory Commission, "Generic Environmental Impact Statement for 

License Renewal of Nuclear Plants - Draft Report for Comment," NUREG-1437 

Volume 1, Revision, Nuclear Regulatory Commission, Washington DC, July, 2009.

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

Fukushima Daiichi Nuclear Power Station,” INPO-11-005, Institute of Nuclear Power 

Operations, Atlanta, GA, November 2011.

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APPENDIX B

SOARCA PEER REVIEW COMMITTEE FINAL LETTERS

AND

SOARCA PEER REVIEW COMMENT RESOLUTION REPORT

The SOARCA team assembled a panel of independent, external technical experts from industry, 

consulting, academia, and research laboratories to review the SOARCA analyses and assure their 

technical accuracy. The 11 members of the committee possess technical expertise in the fields of 

severe accident phenomenology and modeling; plant design, operation, and maintenance; 

mitigation measures; offsite emergency planning, preparedness, and response; radiological health 

consequences; seismic and structural analysis; and probabilistic risk assessment applications.

In addition to ensuring technical accuracy, the committee also assessed whether the underlying 

technical work supported the project’s conclusions.

The SOARCA team provided draft reports of NUREG-1935, “State-of-the-Art Reactor 

Consequence Analyses (SOARCA) Report,” and NUREG/CR-7110, Volume 1, “SOARCA 

Peach Bottom Integrated Analysis,” and Volume 2, “SOARCA Surry Integrated Analysis,” to 

the peer review committee at various points and held meetings with the members of the 

committee in July 2009, September 2009, March 2010, October 2010, and December 2011. 

During some of these meetings, the NRC staff explained how peer reviewer comments were 

considered and addressed. 

The final letters from the individual members of the peer review committee are provided in this 

appendix. Individual letters, rather than a consensus report, were provided so that each 

member’s points of view could be fully expressed. The entirety of the final peer review 

committee report is publicly available in the NRC’s Agencywide Documents Access and 

Management System (ADAMS) at Accession No. ML120610005. 

This appendix also includes the NRC letter providing resolutions of open peer review comments 

from the March 2010 and October 2010 meetings, which is also publicly available at ADAMS 

Accession No. ML11118A056.

In addition, the full documentation of all interactions with the peer review committee is publicly 

available at ADAMS Accession No. ML121250030.

Contents

Peer Review Committee Final Letters, January 2012...........................................B-2

NRC Comment Resolution Report, July 2011...................................................B-89
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