State-of-the-Art Reactor 

Consequence Analyses 

Project

Volume 1:

Peach Bottom Integrated 

Analysis

Office of Nuclear Regulatory Research

NUREG/CR-7110, 

Vol. 1, Rev. 1
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State-of-the-Art Reactor 

Consequence Analyses 

Project

Volume 1:

Peach Bottom Integrated 

Analysis

Manuscript Completed: May 2013

Date Published: May 2013

Prepared by:

Nathan Bixler, Randall Gauntt, Joseph Jones, 

Mark Leonard (dycoda LLC)

Sandia National Laboratories

Albuquerque, New Mexico 87185

Operated for the U.S. Department of Energy

NRC Job Code N6306

Office of Nuclear Regulatory Research

NUREG/CR-7110,

Vol. 1, Rev. 1
Sandia National Laboratories is a multi-program laboratory managed and 

operated by Sandia Corporation, a wholly owned subsidiary of Lockheed 

Martin Corporation, for the U.S. Department of Energy’s National Nuclear 

Security Administration under contract DE-AC04-94AL85000.
iii

ABSTRACT

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

has been the subject of considerable research by the U.S. Nuclear Regulatory Commission 

(NRC) over the last several decades. 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 analysis efforts was a major motivating factor in the genesis of the State-of-the-Art 

Reactor Consequence Analysis (SOARCA) project. By applying modern analysis tools and 

techniques, the SOARCA project developed a body of knowledge regarding the realistic 

outcomes of severe reactor accidents. To accomplish this objective, the SOARCA project used

integrated modeling of accident progression and offsite consequences using 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 focused on providing a realistic evaluation 

of accident progression, source term, and offsite consequences for the Peach Bottom Nuclear 

Power Station. By using the most current emergency preparedness practices, plant capabilities, 

and best available modeling, these analyses are more detailed, integrated, and realistic than past 

analyses. These analyses also consider all mitigative measures, contributing to a more realistic 

evaluation.

PAPERWORK REDUCTION ACT STATEMENT

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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 

currently valid OMB control number.

v

TABLE OF CONTENTS

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

TABLE OF CONTENTS ............................................................................................................. v

LIST OF FIGURES ..................................................................................................................... ix

LIST OF TABLES.................................................................................................................... xvii

EXECUTIVE SUMMARY ....................................................................................................... xxi

ACKNOWLEDGEMENTS .................................................................................................... xxxi

ACRONYMS.......................................................................................................................... xxxiii

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

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

1.2 Objective ...................................................................................................................... 1-2

1.3 Outline.......................................................................................................................... 1-3

2.0 ACCIDENT SCENARIO DEVELOPMENT.............................................................. 2-1

2.1 Sequences Initiated by Internal Events........................................................................ 2-1

2.2 Sequences Initiated by External Events....................................................................... 2-2

2.3 Mitigation Measures .................................................................................................... 2-3

3.0 ACCIDENT SCENARIO DEFINITIONS................................................................... 3-1

3.1 Long-Term Station Blackout ....................................................................................... 3-1

3.1.1 Initiating Event......................................................................................................... 3-1

3.1.2 System Availabilities............................................................................................... 3-1

3.1.3 Operator Actions and Mitigation Measures............................................................. 3-2

3.1.4 Scenario Boundary Conditions................................................................................ 3-3

3.2 Short Term Station Blackout ....................................................................................... 3-8

3.2.1 Initiating Event......................................................................................................... 3-8

3.2.2 System Availabilities............................................................................................... 3-9

3.2.3 Mitigative Actions ................................................................................................... 3-9

3.2.4 Scenario Boundary Conditions.............................................................................. 3-10

3.3 Loss of Vital Alternating Current Bus E-12 .............................................................. 3-11

3.3.1 Initiating Event....................................................................................................... 3-12

3.3.2 System Availabilities............................................................................................. 3-12

3.3.3 Mitigative Actions ................................................................................................. 3-12

3.3.4 Scenario Boundary Conditions.............................................................................. 3-13

4.0 MELCOR MODEL OF THE PEACH BOTTOM PLANT....................................... 4-1

4.1 Reactor Vessel and Coolant System ............................................................................ 4-2

4.2 In-vessel Structures and Reactor Core......................................................................... 4-4

4.2.1 General Configuration of MELCOR In-vessel Nodalization .................................. 4-4

4.2.2 Treatment of Unique Design Features of a BWR Core ........................................... 4-8

4.2.3 In-vessel Structures above the Core....................................................................... 4-11

4.2.4 Leakage through In-core Instrument Tubes........................................................... 4-11

4.3 Primary Containment and Reactor Building.............................................................. 4-14

4.4 Mechanisms for Induced RPV Depressurization....................................................... 4-18
vi

4.4.1 Historical Development of BWR RPV Depressurization Mechanisms................. 4-18

4.4.2 Modeling Approach in SOARCA.......................................................................... 4-23

4.5 Behavior of Ex-Vessel Drywell Floor Debris............................................................ 4-33

4.6 Containment Failure Model ....................................................................................... 4-35

4.7 Radionuclide Inventories and Decay Heat................................................................. 4-38

4.8 Other Modeling Issues and Uncertainties.................................................................. 4-39

4.8.1 Base Case Approach on Important Phenomena..................................................... 4-40

4.8.2 Early Containment Failure Phenomena ................................................................. 4-43

5.0 INTEGRATED THERMAL HYDRAULICS, ACCIDENT PROGRESSION, AND 

RADIOLOGICAL RELEASE ANALYSIS ................................................................ 5-1

5.1 Long-Term Station Blackout – Unmitigated Response............................................... 5-1

5.1.1 Thermal Hydraulic Response................................................................................... 5-3

5.1.2 Radionuclide Release............................................................................................. 5-13

5.2 Long-Term Station Blackout with Mitigation ........................................................... 5-21

5.2.1 Thermal Hydraulic Response................................................................................. 5-22

5.2.2 Radionuclide Release............................................................................................. 5-27

5.3 Short-Term Station Blackout with RCIC Blackstart ................................................. 5-27

5.3.1 Thermal Hydraulic Response................................................................................. 5-27

5.3.2 Radionuclide Release............................................................................................. 5-35

5.4 Short-Term Station Blackout- Sensitivity Case without RCIC Blackstart ................ 5-38

5.4.1 Thermal Hydraulic Response................................................................................. 5-39

5.4.2 Radionuclide Release............................................................................................. 5-45

5.5 SBO Sensitivity Cases with Alternate SRV Failure Criteria ..................................... 5-51

5.5.1 Reasonable variations in failure criteria ................................................................ 5-52

5.5.2 Extreme variations in failure criteria ..................................................................... 5-60

5.6 Loss of Vital AC Bus E-12 – Sensitivity Cases without B.5.b Equipment ............... 5-63

5.6.1 Thermal Hydraulic Response................................................................................. 5-64

5.6.2 Radionuclide Release............................................................................................. 5-67

5.6.3 Sensitivity Analysis ............................................................................................... 5-67

5.7 Peer Review ............................................................................................................... 5-70

5.7.1 Containment Leakage before Failure..................................................................... 5-70

5.7.2 Atmospheric Mixing in the Drywell...................................................................... 5-71

5.7.3 Leakage through TIP guide tubes.......................................................................... 5-74

5.8 Barium Release Variations ........................................................................................ 5-79

6.0 EMERGENCY RESPONSE......................................................................................... 6-1

6.1 Population Attributes................................................................................................... 6-3

6.1.1 Population Distribution............................................................................................ 6-5

6.1.2 Evacuation Time Estimates...................................................................................... 6-5

6.2 WinMACCS................................................................................................................. 6-7

6.2.1 Hotspot and Normal Relocation and Habitability.................................................... 6-7

6.2.2 Shielding Factors ..................................................................................................... 6-8

6.2.3 Potassium Iodide...................................................................................................... 6-8

6.2.4 Adverse Weather...................................................................................................... 6-9

6.2.5 Modeling Using Evacuation Time Estimates .......................................................... 6-9

6.2.6 Cohort Modeling.................................................................................................... 6-11

6.3 Accident Scenarios..................................................................................................... 6-11
vii

6.3.1 Long-Term Station Blackout Unmitigated ............................................................ 6-11

6.3.2 Short-Term Station Blackout with Reactor Core Isolation Cooling Blackstart..... 6-15

6.3.3 Short-Term Station Blackout without Reactor Core Isolation Cooling Blackstart 6-18

6.4 Sensitivity Studies...................................................................................................... 6-20

6.4.1 Sensitivity 1 for the STSBO w/o RCIC Blackstart Evacuation to 16 Miles.......... 6-22

6.4.2 Sensitivity 2 for the STSBO without RCIC Blackstart Evacuation to 20 Miles ... 6-25

6.4.3 Sensitivity 3 for STSBO w/o RCIC Blackstart - Delay in Protective Actions...... 6-27

6.5 Analysis of Earthquake Impact.................................................................................. 6-29

6.5.1 Soils Review .......................................................................................................... 6-29

6.5.2 Infrastructure Analysis........................................................................................... 6-30

6.5.3 Electrical Power and Communications.................................................................. 6-33

6.5.4 Emergency Response ............................................................................................. 6-34

6.5.5 Development of WinMACCS Parameters............................................................. 6-35

6.5.6 Seismic Short-Term Station Blackout without RCIC Blackstart........................... 6-35

6.6 Accident Response and Mitigation of Source Terms ................................................ 6-38

6.6.1 External Resources................................................................................................. 6-42

6.6.2 Mitigation............................................................................................................... 6-43

6.7 Emergency Preparedness Summary and Conclusions............................................... 6-47

7.0 OFF-SITE CONSEQUENCES..................................................................................... 7-1

7.1 Introduction.................................................................................................................. 7-1

7.2 Peach Bottom Source Terms........................................................................................ 7-2

7.3 Consequence Analyses................................................................................................. 7-2

7.3.1 Unmitigated Long-Term Station Blackout Scenario ............................................... 7-3

7.3.2 Short-Term Station Blackout with Reactor Core Isolation Cooling Blackstart....... 7-8

7.3.3 Unmitigated Short-Term Station Blackout without RCIC Blackstart ................... 7-10

7.3.4 Evaluation of the Effect of Seismic Activity on Emergency Response................. 7-15

7.3.5 Evaluation of SST1 Source Term .......................................................................... 7-16

7.3.6 Comparison with the 1982 Siting Study................................................................ 7-20

7.3.7 Surface Roughness................................................................................................. 7-26

7.3.8 Importance of Chemical Classes............................................................................ 7-29

8.0 REFERENCES............................................................................................................... 8-1

APPENDIX A PEACH BOTTOM RADIONUCLIDE INVENTORY......................... A-1

APPENDIX B INPUT PARAMETERS FOR CONSEQUENCE ANALYSIS ............ B-1

ix

LIST OF FIGURES

Figure 4-1 Reactor vessel cross-section detail and MELCOR hydrodynamic 

nodalization.......................................................................................................... 4-3

Figure 4-2 Spatial nodalization of reactor pressure vessel and coolant system .................... 4-5

Figure 4-3 Spatial nodalization of the core and lower plenum.............................................. 4-6

Figure 4-4 Local relative power fraction and five ring radial boundaries of core................. 4-7

Figure 4-5 Module of Four BWR Fuel Assemblies and Associated Channel Boxes............ 4-9

Figure 4-6 BWR Lower Plenum Structures......................................................................... 4-10

Figure 4-7 Potential containment bypass transport pathway through open TIP guide 

tubes................................................................................................................... 4-12

Figure 4-8 Layout of the TIP system................................................................................... 4-13

Figure 4-9 Hydrodynamic nodalization of the primary containment .................................. 4-15

Figure 4-10 Hydrodynamic nodalization of the reactor building (a)..................................... 4-16

Figure 4-11 Hydrodynamic nodalization of the reactor building (b) .................................... 4-17

Figure 4-12 Two-stage Target Rock pilot-operated safety relief valve................................. 4-20

Figure 4-13 Crosby safety relief valve .................................................................................. 4-21

Figure 4-14 Valve stem assembly within valve body............................................................ 4-26

Figure 4-15 Arrangement of 1-dimensional safety relief valve heat-up calculation ............. 4-27

Figure 4-16 Number of safety relief valve cycles from preliminary LTSBO calculation..... 4-28

Figure 4-17 Gas velocity through safety relief valve from preliminary LTSBO 

calculation.......................................................................................................... 4-28

Figure 4-18 Gas temperature through safety relief valve from preliminary LTSBO 

calculation.......................................................................................................... 4-29

Figure 4-19 Gas composition through safety relief valve from preliminary LTSBO 

calculation.......................................................................................................... 4-29

Figure 4-20 Thermal response of safety relief valve using simple 1-dimensional 

models................................................................................................................ 4-31

Figure 4-21 Drywell floor regions for modeling molten-core/concrete interactions ............ 4-34

Figure 4-22 Drywell head flange connection details............................................................. 4-37

Figure 4-23 Drywell flange leakage model versus containment pressure............................. 4-38

Figure 5-1 LTSBO vessel pressure........................................................................................ 5-4

Figure 5-2 LTSBO reactor pressure vessel water level......................................................... 5-5

Figure 5-3 LTSBO fuel cladding temperatures at core mid-plane and in-vessel 

hydrogen generation............................................................................................. 5-7

Figure 5-4 Relocation of core material into the lower plenum during LTSBO (Note: 

The diagrams should in this figure reflect the type(s) of core material 

resident in each spatial cell of the in-vessel MELCOR nodalization as a 

snapshot in time. The order in which different materials within a cell does 

not necessarily reflect their precise spatial configuration.) ................................. 5-8

Figure 5-5 LTSBO temperature of particulate debris on inner surface of lower head ........ 5-10

Figure 5-6 LTSBO lower head temperature (IR-inner radius; OR-outer radius)............... 5-10

Figure 5-7 Debris behavior from lower head dryout through lower head failure for 

LTSBO............................................................................................................... 5-11

Figure 5-8 LTSBO containment pressure............................................................................ 5-12
x

Figure 5-9 LTSBO suppression pool temperature............................................................... 5-12

Figure 5-10 LTSBO In-vessel Fission Product Release from Fuel ....................................... 5-14

Figure 5-11 LTSBO Source Term to the Environment ......................................................... 5-16

Figure 5-12 LTSBO iodine fission product distribution........................................................ 5-17

Figure 5-13 LTSBO cesium fission product distribution ...................................................... 5-17

Figure 5-14 LTSBO tellurium fission product distribution................................................... 5-18

Figure 5-15 LTSBO cerium fission product distribution ...................................................... 5-18

Figure 5-16 LTSBO ex-vessel debris temperatures............................................................... 5-20

Figure 5-17 Reactor Building DF for unmitigated LTSBO................................................... 5-21

Figure 5-18 Mitigated LTSBO vessel pressure ..................................................................... 5-23

Figure 5-19 Mitigated LTSBO coolant level......................................................................... 5-23

Figure 5-20 Mitigated LTSBO core temperature .................................................................. 5-24

Figure 5-21 Mitigated LTSBO Suppression Pool Water Temperature ................................. 5-25

Figure 5-22 Mitigated LTSBO containment pressure ........................................................... 5-26

Figure 5-23 Mitigated LTSBO torus water level................................................................... 5-27

Figure 5-24 Reactor pressure for short term station blackout with RCIC blackstart ............ 5-30

Figure 5-25 Reactor vessel water level for STSBO with RCIC blackstart............................ 5-31

Figure 5-26 Fuel cladding temperatures at core mid-plane and in-vessel hydrogen 

generation for STSBO with RCIC blackstart..................................................... 5-31

Figure 5-27 Temperatures of core debris along inner surface of lower head for STSBO 

with RCIC blackstart.......................................................................................... 5-32

Figure 5-28 Inner and outer surface temperatures of lower head for STSBO with 

RCIC blackstart (IR-inner radius; OR-outer radius).......................................... 5-33

Figure 5-29 Containment pressure history for STSBO with RCIC blackstart ...................... 5-34

Figure 5-30 Environmental source term for STSBO with RCIC blackstart.......................... 5-35

Figure 5-31 Environmental source term for STSBO with RCIC blackstart: details for 

volatile species................................................................................................... 5-36

Figure 5-32 Iodine fission product distribution for STSBO with RCIC blackstart............... 5-37

Figure 5-33 Cesium fission product distribution for STSBO with RCIC blackstart............. 5-37

Figure 5-34 Tellurium fission product distribution for STSBO with RCIC blackstart ......... 5-38

Figure 5-35 Reactor vessel pressure: STSBO without RCIC blackstart ............................... 5-40

Figure 5-36 Reactor vessel water level: STSBO without RCIC blackstart........................... 5-41

Figure 5-37 Temperatures of fuel cladding at core mid-plane: STSBO without RCIC 

blackstart............................................................................................................ 5-41

Figure 5-38 Mass of core debris in RPV lower plenum: STSBO without RCIC 

blackstart............................................................................................................ 5-42

Figure 5-39 Temperature of debris in RPV lower plenum: STSBO without RCIC 

blackstart............................................................................................................ 5-43

Figure 5-40 Lower head temperature: STSBO without RCIC blackstart.............................. 5-43

Figure 5-41 Suppression Pool Temperature: STSBO without RCIC blackstart.................... 5-44

Figure 5-42 Containment Pressure: STSBO without RCIC blackstart.................................. 5-45

Figure 5-43 STSBO without RCIC blackstart environmental source term........................... 5-46

Figure 5-44 STSBO without RCIC blackstart environmental source term: details for 

volatile species................................................................................................... 5-47

Figure 5-45 Spatial distribution of Iodine: STSBO without RCIC blackstart....................... 5-47

Figure 5-46 Spatial distribution of Cesium: STSBO without RCIC blackstart.................... 5-48
xi

Figure 5-47 Spatial distribution of Tellurium: STSBO without RCIC blackstart................ 5-48

Figure 5-48 Reactor Building decontamination factor (DF): STSBO without RCIC 

blackstart............................................................................................................ 5-49

Figure 5-49 Water level and temperature of RPV recirculation loop-A piping: STSBO 

without RCIC blackstart .................................................................................... 5-50

Figure 5-50 CsI deposited on the surface of recirculation loop piping: STSBO without 

RCIC blackstart.................................................................................................. 5-51

Figure 5-51 Comparison of RPV Pressure Response: SRV Failure Sensitivity –

LTSBO............................................................................................................... 5-54

Figure 5-52 Comparison of RPV Level Response: SRV Failure Sensitivity - LTSBO ........ 5-54

Figure 5-53 Comparison of RPV Pressure Response: SRV Failure Sensitivity –

STSBO with RCIC blackstart ............................................................................ 5-55

Figure 5-54 Comparison of RPV Level Response: SRV Failure Sensitivity – STSBO 

with RCIC blackstart.......................................................................................... 5-55

Figure 5-55 Comparison of RPV Pressure Response: SRV Failure Sensitivity –

STSBO without RCIC blackstart....................................................................... 5-56

Figure 5-56 Comparison of RPV Level Response: SRV Failure Sensitivity – STSBO 

without RCIC blackstart .................................................................................... 5-56

Figure 5-57 Loss of vital AC bus E-12 reactor vessel pressure ............................................ 5-65

Figure 5-58 Loss of vital AC bus E-12 reactor water level................................................... 5-66

Figure 5-59 Sensitivity of station battery duration: Reactor water level- loss of vital 

AC bus E-12....................................................................................................... 5-69

Figure 5-60 Sensitivity of station battery duration: Peak clad temperature- loss of vital 

AC bus E-12....................................................................................................... 5-69

Figure 5-61 Effect of increased containment leakage on the release of iodine to 

the environment ................................................................................................. 5-71

Figure 5-62 Drywell atmospheric temperatures in the baseline LTSBO calculation............ 5-73

Figure 5-63 Drywell atmospheric temperatures with imposed drywell circulation .............. 5-73

Figure 5-64 Effect of modeling circulation flow within the drywell on iodine and 

cesium release to the environment..................................................................... 5-74

Figure 5-65 Total Hydrogen Mass in the Reactor Building (TIP Sensitivity calculation 

versus unmitigated LTSBO baseline)................................................................ 5-76

Figure 5-66 Integral Mass of Hydrogen Released to the Reactor Building (TIP 

Sensitivity calculation versus unmitigated LTSBO baseline) ........................... 5-77

Figure 5-67 Fractional Release of Key Fission Products to the Environment (TIP 

Sensitivity calculation versus unmitigated LTSBO baseline) ........................... 5-78

Figure 5-68 Iodine and Cesium in the Reactor Building (TIP Sensitivity calculation 

versus unmitigated LTSBO baseline)................................................................ 5-78

Figure 5-69 Early Leakage of Key Fission Products to the Environment (TIP 

Sensitivity calculation versus unmitigated LTSBO baseline) ........................... 5-79

Figure 5-70 Barium release trends in STSBO compared to LTSBO..................................... 5-80

Figure 5-71 Vapor pressure of Ba metal and BaO (oxide vapor pressure is estimated) ....... 5-81

Figure 5-72 Relative core damage states for STSBO versus LTSBO. (The STSBO 

damage is slightly more extensive with portions of the core forming 

molten pools.) .................................................................................................... 5-82
xii

Figure 5-73 Maximum core temperature for STSBO and LTSBO analyses. (At the 

time of lower head failure, core temperatures are somewhat higher in the 

STSBO case.)..................................................................................................... 5-83

Figure 5-74 Zr content in-vessel............................................................................................ 5-83

Figure 6-1 Peach Bottom 10 and 20 mile analysis areas....................................................... 6-2

Figure 6-2 Unmitigated LTSBO emergency response timeline .......................................... 6-12

Figure 6-3 Duration of protective actions for unmitigated LTSBO .................................... 6-13

Figure 6-4 STSBO with RCIC blackstart emergency response timeline ............................ 6-16

Figure 6-5 Protective actions for STSBO with RCIC blackstart......................................... 6-16

Figure 6-6 STSBO without RCIC blackstart emergency response timeline ....................... 6-18

Figure 6-7 Duration of protective actions for STSBO without RCIC blackstart ................ 6-19

Figure 6-8 Evacuation timeline from Peach Bottom for the 10 to 20 mile region .............. 6-22

Figure 6-9 Sensitivity 1 STSBO without RCIC blackstart - evacuation to 16 miles .......... 6-23

Figure 6-10 Duration of protective actions for Sensitivity 1 STSBO w/o RCIC 

blackstart - evacuation to 16 miles..................................................................... 6-23

Figure 6-11 Sensitivity 2 STSBO without RCIC blackstart - evacuation to 20 miles .......... 6-25

Figure 6-12 Duration of protective actions for sensitivity 2 STSBO w/o RCIC 

blackstart - evacuation to 20 miles..................................................................... 6-26

Figure 6-13 Sensitivity 3 STSBO w/o RCIC blackstart - delay in protective actions........... 6-28

Figure 6-14 Protective action durations for sensitivity 3 STSBO w/o RCIC blackstart -

delay in protective actions ................................................................................ 6-28

Figure 6-15 Roadway network identifying potentially affected roadways and bridges........ 6-31

Figure 6-16 Bridge along Robert Fulton Highway................................................................ 6-33

Figure 6-17 STSBO w/o RCIC blackstart emergency response timeline (seismic 

analysis) ............................................................................................................. 6-36

Figure 6-18 Protective action durations - STSBO w/o RCIC blackstart (seismic 

analysis) ............................................................................................................. 6-36

Figure 6-19 Reactor Building ................................................................................................ 6-45

Figure 6-20 Volume Needed to Fill the Peach Bottom Reactor Building............................. 6-46

Figure 6-21 Peach Bottom Pumping Capacity and Time ...................................................... 6-46

Figure 7-1 Mean, individual, LCF risk per Reactor-Year (1/yr) from the Peach 

Bottom, unmitigated, LTSBO scenario for residents within a circular area 

of specified radius from the plant for three values of dose-truncation level ....... 7-5

Figure 7-2 Mean, individual, LNT, LCF risk per Reactor-Year (1/yr) from the Peach 

Bottom, unmitigated, LTSBO scenario for residents within a circular area 

of specified radius from the plant for the emergency and long-term phases....... 7-6

Figure 7-3 Mean, individual, LCF risk per Reactor-Year (1/yr) from the Peach 

Bottom, STSBO scenario with RCIC blackstart for residents within a 

circular area of specified radius from the plant for three values of dosetruncation level..................................................................................................... 7-9

Figure 7-4 Mean, individual, LNT, LCF risk per Reactor-Year (1/yr) from the Peach 

Bottom, STSBO scenario with RCIC blackstart for residents within a 

specified radius from the plant for emergency and long-term phases............... 7-10

Figure 7-5 Mean, individual, LCF risk per Reactor-Year (1/yr) from the Peach 

Bottom, unmitigated, STSBO scenario without RCIC blackstart for 
xiii

residents within a specified radius from the plant for three values of dosetruncation level................................................................................................... 7-12

Figure 7-6 Mean, individual, LNT, LCF risk per Reactor-Year (1/yr) from the Peach 

Bottom, unmitigated, STSBO scenario without RCIC blackstart for 

residents within a specified radius for the emergency and long-term phases.... 7-13

Figure 7-7 Mean, individual, LNT, LCF risk per Reactor-Year (1/yr) from the Peach 

Bottom, unmitigated, STSBO scenario without RCIC blackstart for 

residents within a circular area of specified radius from the plant showing 

the effect of the size of the evacuation zone ...................................................... 7-15

Figure 7-8 Mean, individual, LNT, LCF Risk per Reactor-Year (1/yr) from the SST1 

source term for residents within a circular area of specified radius from the 

Peach Bottom plant using the SOARCA STSBO ER timing and showing 

the risks from the emergency and long-term phases.......................................... 7-19

Figure 7-9 Percentage contribution to total, emergency-phase, and long-term-phase, 

mean, individual risk for the population within 10 miles by chemical class 

for the Peach Bottom unmitigated LTSBO based on the LNT hypothesis........ 7-31

Figure 7-10 Percentage contribution to total, mean, individual risk for the population 

within 10 miles by chemical class for the Peach Bottom unmitigated 

LTSBO based on US BGR dose truncation....................................................... 7-32

Figure 7-11 Percentage contribution to total, mean, individual risk for the population 

within 10 miles by chemical class for the Peach Bottom unmitigated 

LTSBO based on a truncation level reflecting the HPS Position for 

quantifying health effects................................................................................... 7-32

Figure 7-12 Percentage contribution to total, emergency-phase, and long-term-phase, 

mean, individual risk for the population within 20 miles by chemical class 

for the Peach Bottom unmitigated LTSBO based on the LNT hypothesis........ 7-34

Figure 7-13 Percentage contribution to total, mean, individual risk for the population 

within 20 miles by chemical class for the Peach Bottom unmitigated 

LTSBO based on US BGR dose truncation....................................................... 7-34

Figure 7-14 Percentage contribution to total, mean, individual risk for the population 

within 20 miles by chemical class for the Peach Bottom unmitigated 

LTSBO based on a truncation level reflecting the HPS Position for

quantifying health effects................................................................................... 7-35

Figure 7-15 Percentage contribution to total, emergency-phase, and long-term-phase, 

mean, individual risk for the population within 50 miles by chemical class 

for the Peach Bottom unmitigated LTSBO based on the LNT hypothesis........ 7-35

Figure 7-16 Percentage contribution to total, mean, individual risk for the population 

within 50 miles by chemical class for the Peach Bottom unmitigated 

LTSBO based on US BGR dose truncation....................................................... 7-36

Figure 7-17 Percentage contribution to total, mean, individual risk for the population 

within 50 miles by chemical class for the Peach Bottom unmitigated 

LTSBO based on a truncation level reflecting the HPS Position for 

quantifying health effects................................................................................... 7-36

Figure 7-18 Percentage contribution to total, emergency-phase, and long-term-phase, 

mean, individual risk for the population within 10 miles by chemical class 
xiv

for the Peach Bottom unmitigated STSBO with RCIC blackstart based on 

the LNT hypothesis............................................................................................ 7-37

Figure 7-19 Percentage contribution to total, mean, individual risk for the population 

within 10 miles by chemical class for the Peach Bottom unmitigated 

STSBO with RCIC blackstart based on US BGR dose truncation .................... 7-37

Figure 7-20 Percentage contribution to total, mean, individual risk for the population 

within 10 miles by chemical class for the Peach Bottom unmitigated 

STSBO with RCIC blackstart based on a truncation level reflecting the 

HPS Position for quantifying health effects....................................................... 7-38

Figure 7-21 Percentage contribution to total, emergency-phase, and long-term-phase, 

mean, individual risk for the population within 20 miles by chemical class 

for the Peach Bottom unmitigated STSBO with RCIC blackstart based on 

the LNT hypothesis............................................................................................ 7-38

Figure 7-22 Percentage contribution to total, mean, individual risk for the population 

within 20 miles by chemical class for the Peach Bottom unmitigated 

STSBO with RCIC blackstart based on US BGR dose truncation .................... 7-39

Figure 7-23 Percentage contribution to total, mean, individual risk for the population 

within 20 miles by chemical class for the Peach Bottom unmitigated 

STSBO with RCIC blackstart based on a truncation level reflecting the 

HPS Position for quantifying health effects....................................................... 7-39

Figure 7-24 Percentage contribution to total, emergency-phase, and long-term-phase, 

mean, individual risk for the population within 50 miles by chemical class 

for the Peach Bottom unmitigated STSBO with RCIC blackstart based on 

the LNT hypothesis............................................................................................ 7-40

Figure 7-25 Percentage contribution to total, mean, individual risk for the population 

within 50 miles by chemical class for the Peach Bottom unmitigated 

STSBO with RCIC blackstart based on US BGR dose truncation .................... 7-40

Figure 7-26 Percentage contribution to total, mean, individual risk for the population 

within 50 miles by chemical class for the Peach Bottom unmitigated 

STSBO with RCIC blackstart based on a truncation level reflecting the 

HPS Position for quantifying health effects....................................................... 7-41

Figure 7-27 Percentage contribution to total, emergency-phase, and long-term-phase, 

mean, individual risk for the population within 10 miles by chemical class 

for the Peach Bottom unmitigated STSBO without RCIC blackstart based 

on the LNT hypothesis....................................................................................... 7-42

Figure 7-28 Percentage contribution to total, mean, individual risk for the population 

within 10 miles by chemical class for the Peach Bottom unmitigated 

STSBO without RCIC blackstart based on US BGR dose truncation............... 7-43

Figure 7-29 Percentage contribution to total, mean, individual risk for the population 

within 10 miles by chemical class for the Peach Bottom unmitigated 

STSBO without RCIC blackstart based on a truncation level reflecting the 

HPS Position for quantifying health effects....................................................... 7-43

Figure 7-30 Percentage contribution to total, emergency-phase, and long-term-phase, 

mean, individual risk for the population within 20 miles by chemical class 

for the Peach Bottom unmitigated STSBO without RCIC blackstart based 

on the LNT hypothesis....................................................................................... 7-44
xv

Figure 7-31 Percentage contribution to total, mean, individual risk for the population 

within 20 miles by chemical class for the Peach Bottom unmitigated 

STSBO without RCIC blackstart based on US BGR dose truncation............... 7-44

Figure 7-32 Percentage contribution to total, mean, individual risk for the population 

within 20 miles by chemical class for the Peach Bottom unmitigated 

STSBO without RCIC blackstart based on a truncation level reflecting the 

HPS Position for quantifying health effects....................................................... 7-45

Figure 7-33 Percentage contribution to total, emergency-phase, and long-term-phase, 

mean, individual risk for population within 50 miles by chemical class for 

unmitigated STSBO w/o RCIC blackstart based on LNT hypothesis............... 7-45

Figure 7-34 Percentage contribution to total, mean, individual risk for the population 

within 50 miles by chemical class for the Peach Bottom unmitigated 

STSBO without RCIC blackstart based on US BGR dose truncation............... 7-46

Figure 7-35 Percentage contribution to total, mean, individual risk for the population 

within 50 miles by chemical class for the Peach Bottom unmitigated 

STSBO without RCIC blackstart based on a truncation level reflecting the 

HPS Position for quantifying health effects....................................................... 7-46

xvii

LIST OF TABLES

Table 3-1 Accident scenarios and frequencies..................................................................... 3-1

Table 4-1 Important Design Parameters for Peach Bottom ................................................. 4-2

Table 4-2 Percent change in strength of stainless steel from value at 600oF (589K) ........ 4-32

Table 4-3 Concrete Composition ....................................................................................... 4-35

Table 4-4 Decay Power in Peach Bottom MELCOR Model ............................................. 4-39

Table 5-1 Timing of Key Events for LTSBO....................................................................... 5-2

Table 5-2 Timing of Key Events for Mitigated LTSBO.................................................... 5-22

Table 5-3 Timing of Key Events for the short term station blackout with successful 

RCIC blackstart.................................................................................................. 5-29

Table 5-4 Timing of Key Events for the STSBO without RCIC blackstart....................... 5-39

Table 5-5 Key Metrics of the Impact of Reasonable Variations in SRV Failure 

Criteria ............................................................................................................... 5-59

Table 5-6 The Effects of Extreme Variations in SRV Failure Criteria on Key Severe 

Accident Metrics................................................................................................ 5-62

Table 5-7 Timing of Key Events for Loss of Vital AC Bus E-12...................................... 5-64

Table 5-8 Sensitivities for Loss of Vital AC Bus E-12...................................................... 5-70

Table 6-1 Scenarios Assessed for Emergency Response ..................................................... 6-3

Table 6-2 Peach Bottom Cohort Population Values ............................................................ 6-5

Table 6-3 Peach Bottom Shielding Factors.......................................................................... 6-8

Table 6-4 Unmitigated LTSBO Cohort Timing................................................................. 6-15

Table 6-5 STSBO with RCIC blackstart Cohort Timing. .................................................. 6-17

Table 6-6 STSBO without RCIC blackstart Cohort Timing. ............................................. 6-20

Table 6-7 STSBO without RCIC blackstart, Sensitivity 1................................................. 6-25

Table 6-8 STSBO without RCIC blackstart, Sensitivity 2................................................. 6-27

Table 6-9 Cohort Timing for Sensitivity 3......................................................................... 6-29

Table 6-10 Description of the Potential Evacuation Failure Locations ............................... 6-30

Table 6-11 Cohort Timing for STSBO without RCIC blackstart ........................................ 6-38

Table 7-1 Brief Source-Term Description for Unmitigated Peach Bottom Accident 

Scenarios and the SST1 Source Term from the 1982 Siting Study..................... 7-2

Table 7-2 Mean, Individual, LCF Risk per Event (Dimensionless) for Residents 

within the Specified Radii of the Peach Bottom Site for the Unmitigated 

LTSBO Scenario, which has a Mean, Core Damage Frequency (CDF) of 

3x10-6 pry............................................................................................................. 7-4

Table 7-3 Mean, Individual, LCF Risk per Reactor-Year (1/yr) for Residents within 

the Specified Radii of the Peach Bottom Site for the Unmitigated LTSBO 

Scenario, which has a Mean, CDF of 3x10-6 pry................................................. 7-4

Table 7-4 Mean, Individual, LCF Risk per Event (Dimensionless) for Residents 

within the Specified Radii of the Peach Bottom Site for the STSBO 

Scenario with RCIC blackstart, which Has a Mean CDF of 3x10-7 pry.............. 7-8

Table 7-5 Mean, Individual, LCF Risk per Reactor-Year (1/yr) for Residents within 

the Specified Radii of the Peach Bottom Site for the Unmitigated STSBO 

Scenario with RCIC blackstart, which has a Mean CDF of 3x10-7 pry............... 7-9

Table 7-6 Mean, Individual, LCF Risk per Event (Dimensionless) for Residents 

within the Specified Radii of the Peach Bottom Site for the Unmitigated 
xviii

STSBO Scenario without RCIC blackstart, which Has a Mean CDF of 

3x10-7 pry ........................................................................................................... 7-11

Table 7-7 Mean, Individual, LCF Risk per Reactor-Year (1/yr) for Residents within 

the Specified Radii of the Peach Bottom Site for the Unmitigated STSBO 

Scenario without RCIC blackstart, which Has a Mean CDF of 3x10-7 pry....... 7-11

Table 7-8 Effect of Size of Evacuation Zone on Mean, Individual, LNT, Latent 

Cancer Fatality Risk per Reactor-Year (1/yr) for Residents within the 

Specified Radii of the Site for the Unmitigated STSBO Scenario w/o 

RCIC blackstart.................................................................................................. 7-14

Table 7-9 Mean, Individual, LNT, LCF Risk per Event (Dimensionless) for 

Residents within the Specified Radii of the Peach Bottom Site for the 

Unmitigated, STSBO Scenario without RCIC blackstart Comparing the 

Unmodified Emergency Response and Emergency Response Adjusted for 

the Effect of Seismic Activity on Evacuation Routes and Human Response.... 7-16

Table 7-10 Mean, Individual, LCF Risk per Event (Dimensionless) for Residents 

within the Specified Radii of the Peach Bottom Site for the SST1 Source 

Term previously used in the 1982 Siting Study. All Parameters Other than 

for Source Term Are Taken from the Unmitigated STSBO Scenario 

without RCIC blackstart .................................................................................... 7-17

Table 7-11 Mean, Individual, LNT, LCF Risk per Event (dimensionless) for Residents 

within the Specified Radii of the Peach Bottom Site for the SST1 Source 

Term previously used in the 1982 Siting Study Using Emergency 

Response Parameters from the STSBO Scenario. The Final Column of the 

Table Shows the SOARCA Results for the Unmitigated STSBO without 

RCIC blackstart.................................................................................................. 7-18

Table 7-12 Mean, Individual, Prompt-Fatality Risk per Event (Dimensionless) for 

Residents within the Specified Radii of the Peach Bottom Site for the 

SST1 Source Term previously used in the 1982 Siting Study Using 

Emergency Response Parameters from the SOARCA STSBO without 

RCIC blackstart Scenario................................................................................... 7-20

Table 7-13 Total Release Fractions by Chemical Group Comparison between the 

SST1 and the SOARCA Unmitigated STSBO without RCIC blackstart 

Peach Bottom Scenarios .................................................................................... 7-21

Table 7-14 Comparison of Modeling Choices and Parameters Used to Reconstruct 

1982 Siting Study Results with the Peach Bottom Unmitigated STSBO 

without RCIC blackstart from SOARCA .......................................................... 7-22

Table 7-15 Mean, LNT, LCF Risk per Event (dimensionless) for Residents within the 

Specified Radii of the Peach Bottom Site for the Recreation of the Siting 

Study Using the SST1 Source Term and for the Unmitigated STSBO 

without RCIC blackstart Calculated for SOARCA. CDFs Were Estimated 

to Be 10-5/yr and 3x10-7/yr for the SST1 and STSBO without RCIC 

blackstart Source Terms, Respectively.............................................................. 7-25

Table 7-16 Mean, LCF Risk per Event (dimensionless) Using US Background (620 

mrem/yr) Dose-Truncation for Residents within the Specified Radii of the 

Peach Bottom Site for the Recreation of the 1982 Siting Study Using the 

SST1 Source Term and for the Unmitigated STSBO without RCIC 
xix

blackstart Scenario Calculated for SOARCA. CDFs Were Estimated to Be 

10-5/yr and 3x10-7/yr for the SST1 and STSBO without RCIC blackstart 

Source Terms, Respectively............................................................................... 7-25

Table 7-17 Mean, LCF Risk per Event (dimensionless) for Three Levels of Dose 

Truncation for Residents within the Specified Radii of the Peach Bottom 

Site for the Recreation of the 1982 Siting Study Using the SST1 Source

Term. The CDF for This Sequence of the 1982 Siting Study is 10-5

/yr ............ 7-26

Table 7-18 Surface roughness for various land-use categories for the area surrounding 

the Peach Bottom site......................................................................................... 7-27

Table 7-19 Deposition Velocities Used for Base Case Calcs and for Surface 

Roughness Sensitivity Study for Each of the Ten Aerosol Bins in the 

MELCOR Model ............................................................................................... 7-28

Table 7-20 Mean, Individual, LCF Risk per Event (Dimensionless) for Residents 

within the Specified Radii of the Peach Bottom Site for the Unmitigated 

STSBO Scenario without RCIC blackstart, which has a Mean CDF of 

3x10-7 pry. Risks are shown for the base case (10 cm) and the sensitivity 

case (60 cm)....................................................................................................... 7-28

xxi

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,”

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

and WASH-1400, “Reactor Safety Study: An Assessment of Accident Risks in U.S. Commercial 

Nuclear Power Plants.” 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. 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 

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

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. This scenarioselection process captured the more important internally and externally initiated core damage 

scenarios.
xxiii

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 

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 
xxiv

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
xxv

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 

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 
xxvi

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 1 and 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.
xxvii

Table ES-1 Offsite Consequence Results for Peach Bottom Scenarios Assuming Linear 

No- Threshold (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

Scenario-specific 

risk

(CDF x 

Conditional)

of latent cancer 

fatality for an 

individual located 

within 10 miles 

(per reactor year)

Conditional 

scenario-specific 

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 reactor year)

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.
xxviii

Table ES-2 Offsite Consequence Results for Surry 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 reactor-year)

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 reactor-year)

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 was 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 on site within 24 hours and connected and functioning within 48 hours. Therefore 66 

hours would allow ample time for mitigation via measures brought to the site 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 x 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 

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. 
xxix

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 and to an extrapolation of the 1982 Siting 

Study SST11 results. 

Figure ES-3 Comparison of 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 

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

risk.

 

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.
xxx

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

annual US cancer fatality risk from all causes are provided to give context that may help the 

reader to understand the contribution to cancer risks from these nuclear power plant accident 

scenarios. However, such comparisons have limitations for which the 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 most 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.
xxxi

ACKNOWLEDGEMENTS

The contributions from the following individuals in preparing this document are gratefully 

acknowledged. 

Jon Ake U.S. Nuclear Regulatory Commission

Terry Brock U.S. Nuclear Regulatory Commission

Richard Chang U.S. Nuclear Regulatory Commission

Ata Istar U.S. Nuclear Regulatory Commission

Robert Prato U.S. Nuclear Regulatory Commission

Jocelyn Mitchell U.S. Nuclear Regulatory Commission

Mark Orr U.S. Nuclear Regulatory Commission

Jason Schaperow U.S. Nuclear Regulatory Commission

Abdul Sheikh U.S. Nuclear Regulatory Commission

Richard Sherry U.S. Nuclear Regulatory Commission

Randolph Sullivan U.S. Nuclear Regulatory Commission

Charles G. Tinkler U.S. Nuclear Regulatory Commission

Nathan E. Bixler Sandia National Laboratories

Shawn P. Burns Sandia National Laboratories

Randall O. Gauntt Sandia National Laboratories

Joseph A. Jones Sandia National Laboratories

Raymond J. Jun Sandia National Laboratories

Douglas M. Osborn Sandia National Laboratories

Jesse Phillips Sandia National Laboratories

Mark T. Leonard dycoda, LLC

Kenneth C. Wagner dycoda, LLC

xxxiii

ACRONYMS

AC

ATWS

BWR

Alternating Current

Anticipated Transient Without Scram

Boiler Water Reactor

CD Core Damage

CDF Core Damage Frequency

CRDHS Control Rod Drive Hydraulic System

CRGT Control Rod Guide Tube

CST Condensate Storage Tank

DC

DOE

Direct Current

Department of Energy

EAL Emergency Action Level

EAS Emergency Alert System

EOF Emergency Operations Facility

EOP Emergency Operating Procedure

EP Emergency Preparedness

EPA Environmental Protection Agency

EPDM Ethylene Propylene Diene Methylene

EPZ Emergency Planning Zone

ERO Emergency Response Organization

ETE Evacuation Time Estimate

GE General Emergency

GNF Global Nuclear Fuel

HCTL

HPCI

Heat Capacity Temperature Limit

High Pressure Coolant Injection

HPS Health Physics Society

IPE Individual Plant Examination

IPEEE Individual Plant Examination of External Event

LNT Linear, No-Threshold

LOCA Loss of Coolant Accident

LPCI Low Pressure Coolant Injection

LPI Low Pressure Injection

LTSBO

MACCS2

Long-Term Station Blackout

MELCOR Accident Consequence Code System, Version 2 

MCCI Molten Corium-Concrete Interactions

MSIV

NG

Main Steam Isolation Valve

Noble Gases

NRC Nuclear Regulatory Commission

NRCS Natural Resources Conservation Service

NRF National Response Framework

ORO Offsite Response Organization

PEMA Pennsylvania Emergency Management Agency

PRA Probabilistic Risk Assessment

QHO Quantitative Health Objective
xxxiv

RAMCAP Risk Analysis and Management for Critical Asset Protection

RCIC Reactor Core Isolation Cooling

RHR Residual Heat Removal

RPF Relative Power Fraction

RPV Reactor Pressure Vessel

SAE Site Area Emergency

SAMG Severe Accident Management Guidelines

SECPOP SECtor POPulation and Economic Estimator

SOARCA State-of-the-Art Reactor Consequence Analysis

SPAR Standardized Plant Analysis Risk

SRV Safety Relief Valve

STCP Source Term Code Package

STSBO Short-Term Station Blackout

TAF Top of Active Fuel

TSC Technical Support Center

UE Unusual Event
1-1

1.0 INTRODUCTION

This document describes the detailed severe accident analyses (i.e., MELCOR and the MELCOR 

Accident Consequence Code System, Version 2 (MACCS2) code calculations) performed for the 

Peach Bottom Atomic Power Station as part of the NRC’s State-of-the-Art Reactor Consequence 

Analyses (SOARCA) project. A separate volume of this report describes severe accident 

analyses for the Surry Power Station. A summary report describing the formal Peer Review 

Committee activities, comments, and resolutions was published as a separate document entitled 

“Summary Report: Peer Review of the State-of-the Art Reactor Consequence Analyses 

(SOARCA) Project,” [1].

1.1 Background

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

been the subject of considerable research by NRC, the nuclear power industry, and the 

international nuclear energy research community. 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. 

An insight gained from these security assessments was that updated analyses of severe reactor 

accidents were needed to reflect realistic estimates of the more likely outcomes considering the 

current state of plant design and operation and the advances in our understanding of severe 

accident behavior. The SOARCA project evaluates plant improvements and changes (either of 

which can alter safety margins) not reflected in earlier assessments. These include 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. SOARCA’s more realistic modeling updates the more conservative quantifications 

of offsite consequences found in earlier NRC publications such as NUREG/CR-2239, “Technical 

Guidance for Siting Criteria Development” referred to in this report as the Siting Study.

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 transition from event-based to symptom-based Emergency Operating Procedures

(EOPs) for the boiling-water and pressurized-water reactor 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.
1-2

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

implementation of the Severe Accident Management Guidelines (SAMGs).

• Use of additional safety enhancements, described in Title 10, Section 50.54(hh) of the 

Code of Federal Regulations (10CFR50.54(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 and/or delay or prevent the release.

• Improved phenomenological understanding of influential processes such as:

o in-vessel steam explosions,

o Mark I containment drywell shell attack,

o dominant chemical forms for fission products,

o direct containment heating,

o hot leg creep rupture,

o reactor pressure vessel failure, and

o molten core concrete interactions.

1.2 Objective

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

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

Corresponding and supporting objectives are as follows:

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

assessments including system improvements, training and emergency procedures, offsite 

emergency response, and recent security-related enhancements described in Title 10, 

Section 50.54(hh) of the Code of Federal Regulations (10CFR50.54(hh)) as well as plant 

updates in the form of power uprates and higher core burnup.

• Incorporate state-of-the-art integrated modeling of severe accident behavior that includes 

the insights of some 25 years 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.

• Enable 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.”
1-3

1.3 Outline

Section 2 briefly summarizes the method used to select the specific accident scenarios subjected 

to detailed computational analysis. Additional details of this method can be found in summary 

report (NUREG-1935) in this series. Section 3 describes the results of the mitigation measures 

assessment process when it was applied to Peach Bottom. Section 4 describes the key features of 

the MELCOR model of the Peach Bottom Atomic Power Station. Section 5 describes for each 

case the results of MELCOR calculations of the thermal hydraulics and, when core damage was 

predicted, the accident progression and radionuclide release to the environment. Section 6 

describes the way in which plant-specific emergency response actions were represented in the 

MACCS2 code calculations of offsite consequences, and Section 7 describes the MACCS2 

calculations of offsite consequences for each accident scenario. Section 7 also describes 

analyses of offsite consequences that compare SOARCA results to consequence results from 

earlier studies. References cited in this report are listed in Section 8.

2-1

2.0 ACCIDENT SCENARIO DEVELOPMENT

The SOARCA project considered accident sequences that have an estimated frequency greater 

than 1x10-6 per reactor-year (pry) of reactor operation as candidate sequences for further 

deterministic evaluations. It also considered sequences with frequency as low as 1x10-7 pry if 

they were judged to proceed rapidly enough to have the potential for generating significant early 

releases of radionuclides to the environment or involve a radiological transport pathway from the 

reactor to the environment that bypasses the containment pressure boundary (i.e., so-called 

‘bypass sequences’). Section 2.1 and Section 2.2 summarize the methods used to identify these 

sequences and the screening process for retaining candidate sequences. 

Once candidate accident sequences were identified, the analysts evaluated realistic opportunities 

for plant personnel to respond to the observed failures of control and safety systems. The 

manner in which mitigation measures were evaluated for each accident sequence is described in 

Section 2.3.

The end result of this process was a list of accident scenarios (i.e., event sequence plus options 

for mitigation) that were subjected to detailed analysis of plant response including, as 

appropriate, radionuclide release to the environment (described in Sections 4 and 5) and offsite 

radiological consequence (Sections 6 and 7).

2.1 Sequences Initiated by Internal Events

The following scenario selection process was used to determine the scenarios for further 

analysis:

1. Identified candidate accident sequences in analyses using plant-specific Standardized

Plant Analysis Risk (SPAR) models (Version 3.31).

a. Initial Screening: Core damage sequences with low frequencies (less than 

1x10-8 pry) were eliminated from consideration. This step affected only 4 percent 

of the overall core damage frequency (CDF) for Peach Bottom.

b. Sequence Evaluation: Dominant cutsets for the remaining sequences were 

reviewed to characterize system and equipment availabilities and accident 

sequence timing.

c. Sequence Grouping: Sequence cutsets with similar equipment availabilities and 

estimated time for the onset of core damage were aggregated into a single 

‘sequence group’ or ‘scenario.’

2. The availability of containment systems was evaluated for each sequence, using system

dependency tables. These tables delineate the support systems required for containment 

systems to function. The status of containment systems was then appended to the 

accident sequence description.

3. Core damage sequences from the licensee’s probabilistic risk assessment (PRA) model 

were compared with the scenarios determined by using the SPAR models. Differences 

were resolved during meetings with licensee staff.

4. The screening criteria described above were applied to eliminate extremely low 

frequency sequences from further analyses.
2-2

The initial pass through this process identified only one sequence at Peach Bottom that satisfied 

the 1.x10-6 pry frequency threshold criteria. The sequence is initiated by the failure of vital 

alternating current (AC) bus E-12, which disables several (but not all) trains of safety equipment. 

The estimated frequency of this sequence was initially found to be above the 1x10-6 pry

threshold. As a result, the sequence was forwarded for an assessment of mitigative measures 

(see Section 2.3) and a deterministic analysis of accident progression and radiological release. 

However, the SPAR model was later found to incorrectly represent certain features of this 

sequence, and the sequence’s frequency was reduced to below the screening criterion. Further, 

the MELCOR analysis performed for this sequence determined that it would not, in fact, result in 

core damage. Despite both of these late conclusions about the characteristics of this sequence, 

the analysis results provide unique insights into the effectiveness of small capacity non-safety 

related equipment in the plant to mitigate certain accident sequences. As a result, the calculation 

of event progression for this sequence was retained in this report (refer to Section 5.6).

This process provided the basic characteristics of each scenario. However, it is necessary when 

calculating a consistent integrated response to have more detailed information about a scenario 

than is provided in a PRA model. To capture the additional sequence details, the project 

conducted further analysis of system descriptions and a review of the normal and emergency 

operating procedures.

2.2 Sequences Initiated by External Events

External events include internal flooding and fire; seismic events; extreme wind-, tornado-, and 

hurricane-related events; and other similar events that may be applicable 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, individual plant examination for external event 

submittals, and other relevant generic information. 

Seismic-initiated sequences were found to be the most restrictive in terms of the timing of 

equipment failure and the ability to successfully implement onsite mitigative measures and 

offsite protective actions. In addition, the seismic-initiated sequences (as a group) are important 

contributors to the external event core damage and release frequencies. As a result, plant and 

offsite response to external event sequences was assumed to be represented by an earthquake of 

sufficiently large magnitude to result in widespread damage to important plant support systems, 

such as electric power sources. The seismic events considered in SOARCA result in loss of 

offsite and onsite AC power (i.e., long-term station blackout (LTSBO)) and, for the more severe 

seismic events, loss of DC power (i.e., short-term station blackout (STSBO)). The sequence 

selection process identified the LTSBO, which has an estimated frequency of 1x10-6 to 5x10-6

pry. Even though the STSBO, which has an estimated frequency of 1x10-7 to 5x10-7 pry, did not 

meet the screening criterion, we elected to also analyze it to address the impact of the timing of 

the offsite release on early and latent cancer fatality risks.

The magnitude of the seismic initiating event reflected by these frequencies corresponds to 0.3 to 

0.5 peak ground acceleration (pga) for the LTSBO, and 0.5 to 1.0 pga for the STSBO. As noted 

earlier, the initiating event for all external event sequences was assumed to be represented by a 

seismic event because it was judged to limit equipment available to prevent, minimize or delay
2-3

radiological releases to the environment, and affect offsite response. Seismic PRAs for several 

BWRs were reviewed to assess the availability of installed systems. 

2.3 Mitigation Measures

Actions that would be taken by plant personnel in response to system failures caused by the 

postulated seismic initiating events were reviewed and incorporated into the development of the 

accident scenarios evaluated with deterministic calculations. These actions are guided by 

plant-specific EOPs, severe accident management guidelines (SAMGs) and mitigation measures 

developed specifically in response to security concerns that arose from the events of September 

11, 2001, as codified in 10CFR50.54(hh). Examples of the latter type of measures include 

portable equipment, such as generators or other power supplies to open or close valves or 

energize key instrumentation, diesel driven pumps, and air bottles to open air operated valves. 

Applicable procedures have been written to align and operate these mitigative measures under 

severe accident conditions. 

The SOARCA analysis team developed a timeline for implementing the mitigation measures 

directed in plant-specific procedures and mobilizing support organizations after discussing each 

scenario with licensee personnel that have experience in operations, engineering and facility 

management. 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. The 

resulting sequence of events and estimated timelines for each scenario are described in the next 

section.

The seismic events considered in SOARCA result in loss of offsite and onsite AC power 

(i.e., LTSBO) and, for the more severe seismic events, loss of DC power (i.e., STSBO). Under 

these conditions, the use of the turbine driven reactor core isolation cooling (RCIC) system is an 

important mitigation measure. Diverse procedures have been developed for boiling water 

reactors (BWRs), including a procedure to start and operate the RCIC system without DC control 

power, which facilitates a managed response to station blackout conditions. These procedures 

were discussed during site visits. This is known as RCIC blackstart. Under 

10CFR50.54(hh), mitigation measures also include long-term operation of the RCIC system 

without electricity (RCIC blackrun), using a portable generator to supply power for 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. For a LTSBO, RCIC can be used to 

cool the core until battery exhaustion. After battery exhaustion, RCIC black run can be used to 

continue to cool the core. 

The seismic initiating event for the station blackout accident scenarios might rupture the 

condensate storage tank (CST), which is the primary water reservoir for RCIC. However, the 

CST is surrounded by a reinforced concrete wall or moat, which could retain water drained from 

the CST. Therefore, suction from the CST would not necessarily be interrupted by a loss of CST

integrity. Plant specific procedures provide plant personnel with instructions for refilling the 

CST from a variety of possible resources using equipment that is not dependent on availability of

AC or DC power (e.g., portable generators, pumps, etc.). Therefore, sustained availability of the 
2-4

CST is assumed in the current analysis. This assumption is not critical to the calculation of the 

LTSBO scenario. Operators could also manually re-align RCIC suction to the torus, if 

necessary. MELCOR calculations for the LTSBO showed that several hours would be available 

before torus temperature and pressure conditions would reach RCIC isolation setpoints. 
3-1

3.0 ACCIDENT SCENARIO DEFINITIONS

As discussed in Section 2.0, three scenarios were chosen for analysis. Table 3-1 summarizes the 

estimated frequency for each of these scenarios. Sections 3.1 and 3.2 provide detailed 

descriptions of the LTSBO and STSBO, respectively. Section 3.3 describes the loss of vital AC 

bus E-12 scenario, which was determined to not result in core damage.

Table 3-1 Accident scenarios and frequencies

Scenario Description

Frequency 

(per reactor year)

LTSBO 1x10-6 to 5x10-6

STSBO 1x10-7 to 5x10-7

Loss of vital AC bus E-12 5x10-7

3.1 Long-Term Station Blackout

Section 3.1.1 describes the status of the plant immediately following the seismic event. 

Section 3.1.2 then discusses the response of plant systems available to the initiating event. 

Actions that can be taken by plant personnel to mitigate the effects of failed plant safety 

functions are described in Section 3.1.3. Section 3.1.4 describes two scenarios that differ in the 

assumed success (or failure) of the mitigative actions. Mitigated scenarios are defined as those 

in which the mitigative actions are successful. Unmitigated scenarios are defined as those in 

which certain key mitigation measures are not successfully implemented.

3.1.1 Initiating Event

The LTSBO scenario is a composite of several similar sequences that differ only by their 

initiating event. Initiators can be a seismic event, or an internal fire, or flood. The seismic event 

is the largest contributor to the composite frequency of this scenario, and is used as the basis for 

defining consequential events and conditions at the plant. Damage caused by the earthquake is 

assumed to result in a total loss of offsite power. In addition, onsite AC power is unavailable, 

due to failures of diesel generators to start or run as needed. The diesel generators at Peach 

Bottom have a shared configuration between the two units, which causes the loss of offsite and 

onsite AC power failure to affect both units. However, the deterministic analysis of subsequent 

accident progression described in Section 5.0 considers only the response of one unit.

3.1.2 System Availabilities

Reactor scram, reactor isolation and containment isolation immediately follow the initiating 

event. Neither active AC nor DC power is necessary for these safety functions to occur. The 

station blackout electric power line from the hydroelectric station downstream of the plant site is 

also assumed to be unavailable because of structural damage to the dam and electric station 

components. The station batteries are assumed to provide DC power for 4 hours following loss 

of AC power, allowing DC controlled components and systems to operate as required for this 

period. Instrumentation would also be available using station batteries. This duration of 

DC power assumes that the batteries are at their end of life and that operators successfully follow 

procedural actions to shed nonessential loads from the emergency DC bus. As a result, the 

steam-driven High-Pressure Coolant Injection (HPCI) system and RCIC would be available with 
3-2

automatic activation for at least the first 4 hours of the scenario. Only RCIC operation is 

considered in the current analysis because the larger and functionally redundant HPCI system is 

not needed to respond if RCIC successfully operates. Additionally, remote manual control of the 

safety relief valves (SRVs) would be available. 

3.1.3 Operator Actions and Mitigation Measures

An unmitigated MELCOR calculation was performed for the LTSBO scenario assuming that 

manual actions to mitigate the loss of vital safety systems are limited to those currently 

implemented in EOPs2

. The effects of additional mitigative actions and equipment at the plant 

(i.e., 10CFR50.54(hh) measures) were then examined in a separate ‘mitigated’ calculation. 

Results of the unmitigated calculation are described in Section 5.1, and results of the mitigated 

accident scenario are described in Section 5.2.

Two operator actions were credited in the unmitigated long-term station blackout calculation3. 

First, operators are assumed to open one SRV to begin a controlled depressurization of the 

reactor vessel approximately 1 hour after the initiating event. This action is prescribed in station 

emergency procedures to prevent excessive cycles on the SRV. The target reactor vessel 

pressure is at or above 125 pounds per square inch (psi), which would permit continued 

operation of RCIC or HPCI, if necessary. Five SRVs associated with the automatic 

depressurization system would be available for this operation. These SRVs are provided with 

accumulators that provide a back-up pneumatic supply for operation of the valves upon loss of 

the Instrument Nitrogen System. Second, operators are assumed to take manual control of RCIC

approximately 2 hours after the initiating event. This involves remote (i.e., from the control 

room) manipulation of the position of the steam throttle valve at the inlet to the RCIC turbine to 

reduce and control turbine speed. This action reduces and stabilizes coolant flow from the RCIC 

pump to maintain the reactor vessel level within a prescribed range.

The mitigated LTSBO calculation credits four additional manual actions. First, two portable 

AC power supplies (e.g., 10CFR50.54(hh) equipment) are assumed to be connected to restore 

power to the DC buses delivering power to at least one SRV and (separately) to essential 

instrumentation (e.g., level indication). The precise time this action is completed is not 

 

2 Current procedures for using B.5.b equipment (portable nitrogen bottles, in particular) to open containment 

vent path isolation valves only address the drywell ventilation system pathway. This path is not desirable 

under accident conditions in which the vented gas would be comprised of steam and/or hydrogen and/or

high levels of radioactivity. The potential for adverse effects to the reactor building have been 

demonstrated in past NRC research programs concerning containment venting at Peach Bottom 

(NUREG/CR-4696, for example) and plant-specific emergency procedures caution against using this vent 

path if an alternative is available, such as the 16-in hardened vent or the 6-in integrated leak rate test 

(ILRT) line. B5.b equipment could be used to open the isolation valves for either of these paths, but plantspecific procedures have not yet been developed.

3

 The action times used in this analysis were based on ‘table-top’ exercises among NRC staff and licensee 

personnel, in which the anticipated accident sequence timeline was reviewed to characterize a reasonable 

time at which action would be taken. If the two actions credited here are completed sooner than the times 

assumed in these calculations (which is possible), the net effect on the overall hydraulic response of the 

reactor and containment would be inconsequential.
3-3

important, provided it occurs before station battery power is exhausted (i.e., at least 4 hours after 

the initiating event). This ensures continuous control of RPV pressure and water level, by 

holding open a single SRV and facilitating operation of the RCIC system after the loss of the 

onsite emergency DC power supply4.

The second action involves staging and operation of portable pumping equipment to refill the 

CST, which is located outside the reactor building. Various equipment and water resources 

could be used for this purpose. The installed fire protection system is assumed to be disabled as a 

consequence of the initiating event (i.e., seismic damage). However, the diesel-driven portable 

pump (i.e., 10CFR50.54(hh) equipment) could be used and/or other mobile equipment. 

Calculations described in Section 5.2 indicate a pumping capacity of less than 200 gpm would 

ensure the CST inventory is never depleted for the LTSBO sequence considered here. It is 

assumed that pumps would take suction from the cooling tower basin or the Susquehanna River.

The third manual action involves opening a containment vent pathway to relieve pressure and 

prevent structural damage to the containment pressure boundary. Instructions for this action are 

available in the form of detailed plant-specific procedures, which are outlined in Section 3.1.4.2. 

These procedures include guidance for selecting an appropriate containment vent path and the 

actuation criteria for venting. 

Procedures also address manual actions to prevent automatic isolation of RCIC by manually 

defeating trip signals that might be received as plant conditions change in time. For example, the 

lack of active containment heat removal in the LTSBO scenario, combined with steam discharge 

from the RPV to the suppression pool, leads to an increase in suppression pool temperature. This

in turn, would eventually generate signals for low steam inlet pressure for the RCIC turbine and 

(possibly) high RCIC room temperature. Among the observations made from the events that 

occurred at the Fukushima Daiichi Nuclear Power Station in Japan in March 2011 is that during 

a station blackout accident sequence, RCIC can operate for a considerable period of time (i.e., 

>24 hrs) beyond the point at which these isolation signals would occur. As a result, the fourth

manual action assumed in the MELCOR analysis of the mitigated LTSBO is that operators

defeat RCIC trip signals. If these actions are not successful, and the RCIC pump were to trip off 

coolant makeup is assumed to be provided through low-pressure injection lines by means of realigning the portable diesel-driven pump for direct RPV injection.

3.1.4 Scenario Boundary Conditions

Section 3.1.4.1 lists the sequence of events prescribed in the unmitigated LTSBO calculation. 

Section 3.1.4.2 summarizes the sequence of events in the mitigated LTSBO calculation.

3.1.4.1 Unmitigated Event Sequence

The unmitigated case credits automatic system responses and manual actions that would be 

directed by plant EOPs, such as operator reactor vessel depressurization and intervention to 

 

4 Attention also needs to be paid to managing pneumatic supply to the SRV accumulators. The limited gas 

supply in the accumulators can be replenished from multiple sources including alignment to the CAD tank 

via the instrument nitrogen header or placing the ADS nitrogen bottles in service. An external tanker truck 

connection can also be made to maintain a long-term pneumatic supply.
3-4

control RCIC injection flow (after its automatic actuation) to stabilize and maintain the level 

within a target range. The unmitigated case did not credit operator actions that are beyond the 

scope of EOPs – primarily, the mitigation measures installed in response to 10CFR50.54(hh). 

The effects of such actions were examined in the mitigated scenario, which is described in 

Section 3.1.4.2. 

The following is the timeline of events and operator actions that were credited in the unmitigated 

case.

Unmitigated Event Timeline

Event Initiation and Initial Plant Response

• Seismic event results in a loss of offsite power.

• All diesel generators assumed to fail to start.

• DC power (station batteries) and associated emergency buses are available.

• Reactor trips (successful scram).

• Reactor pressure vessel isolates (i.e., main steam line isolation valves (MSIVs) close) and 

containment isolation valves close.

• The control rod drive hydraulic system (CRDHS), low pressure coolant injection (LPCI)

mode of the residual heat removal (RHR) system, standby liquid control, condensate, 

containment cooling and containment spray systems are not available.

• Control room receives indication that plant is in a station blackout condition requiring the

operator to enter Special Event Procedure SE-11, “Station Blackout Procedure.” RCIC

automatically starts when level drops to low-level setpoint with suction aligned to the 

CST.

10-15 minutes

• Plant operations personnel complete an initial assessment of plant status.

• RCIC auto-starts to make up for coolant losses and maintain RPV level.

• HPCI might auto-start in response to initial transient, but is secured by operations 

personnel.

• In accordance with SE-11, plant operations personnel initiate the following mitigation 

measures:

o Attempt to line up the Conowingo hydroelectric dam (i.e., station blackout line) 

as an alternative offsite power source, but the line is not available.

o Attempt manual start of emergency diesel generators, but none is available.

o Begin to shed non-essential loads from the emergency DC bus.
3-5

50 minutes

• Emergency Operation Facility (EOF) is manned. The EOF is located in the Philadelphia 

area, far away from the plant. Therefore, the timing should not be affected by the seismic 

event.

1 hour

• Actions to shed non-essential loads from DC bus is complete, battery life extended to an 

estimated 4 hours. 

• Initiate RPV depressurization by opening one SRV. The target reactor coolant system 

(RCS) pressure is 125 psi.

2 hours

• Operator assumes remote manual control of RCIC flow.

2.25 hours

• Technical Support Center (TSC) is assumed operational.

4 hours

• DC power from station batteries is exhausted. The consequences of a loss of DC power 

are:

o Open SRV closes.

o Remote control of RCIC flow terminates. The system is assumed to continue 

operate at the conditions it experienced immediately prior to battery exhaustion. 

This effectively assumes the RCIC pump continues to operate at a constant rate, 

ultimately flooding the main steam line causing delayed termination of RCIC.

3.1.4.2 Mitigated Event Sequence

The mitigated case credits the same actions assumed in the unmitigated event sequence, but also 

credits implementation of the mitigation measures installed in response to the requirements 

outlined in 10CFR50.54(hh). As noted in Section 3.1.3, this includes the staging and alignment 

of a variety of portable equipment, such as electric power supplies, low-pressure coolant 

injection pump and gas bottles for air-operated valve control. 

It also credits manual opening of a containment vent path when containment pressure reaches 

unacceptably high levels. In the current analysis, a 16-in. (hard-pipe) vent path is assumed to be 

opened when containment pressure exceeds 45 psig. This vent path and opening pressure were

selected based on a review of plant-specific containment vent procedures. Selection of the vent 

pathway took two factors into account. One factor is the availability of electric power or other 

equipment needed to open vent path isolation valves. The loss of all AC power supplies in the 
3-6

LTSBO sequence demands use of portable 10CFR50.54(hh) equipment for this purpose.5 A 

second factor is caution to avoid creating a hazardous environment in the reactor building by 

using vent pathways involving containment ventilation system ductwork. These factors led to 

the assumption that the preferred vent path would either be the 2-in hard pipe vent to the standby 

gas treatment system (SGTS), or the 16-in hard pipe vent pathway would be used. The larger 

vent path was used in this analysis because it was judged necessary to prevent further increases 

in containment pressure.

The procedure for containment pressure control recommends opening a vent path if pressure 

exceeds the ‘peak containment pressure limit’ (PCPL) of 60 psig. However, a high turbine 

exhaust pressure isolation signal for RCIC would be received at a pressure of 50 psig. Since 

RCIC is the only operating coolant injection system available in this scenario,

6 this analysis 

assumed operators would open the containment vent path at a slightly lower pressure (45 psig), 

thereby averting termination of RCIC flow.

The following is a timeline for these and other actions after the initiating event.

Mitigated Event Timeline

Event Initiation and Initial Plant Response

• Seismic event results in a loss of offsite power.

• All diesel generators fail to start.

• DC power (station batteries) and associated emergency buses are available.

• Reactor trips (successful scram).

• Reactor pressure vessel isolates (i.e., MSIVs close) and containment isolation valves 

close.

• The CRDHS, LPCI mode of the RHR system, standby liquid control, condensate, 

containment cooling and containment spray systems are not available.

 

5 Current procedures for using B.5.b equipment (portable nitrogen bottles, in particular) to open containment 

vent path isolation valves only address the drywell ventilation system pathway. This path is not desirable 

under accident conditions in which the vented gas would be comprised of steam and/or hydrogen and/or

high levels of radioactivity. The potential for adverse effects to the reactor building have been 

demonstrated in past NRC research programs concerning containment venting at Peach Bottom 

(NUREG/CR-4696, for example) and plant-specific emergency procedures caution against using this vent 

path if an alternative is available, such as the 16-in hardened vent or the 6-in integrated leak rate test 

(ILRT) line. B5.b equipment could be used to open the isolation valves for either of these paths, but plantspecific procedures have not yet been developed. 6 The portable coolant injection pump is considered a viable backup to RCIC, but continued RCIC operation 

would be the preferred method for coolant injection.
3-7

• Control room receives indication that plant is in a station blackout condition requiring the

operator to enter Special Event Procedure SE-11, “Station Blackout Procedure.”

• RCIC automatically starts when level drops to low-level setpoint with suction aligned to 

the CST.

• Cooling tower basin is assumed to be undamaged, containing about 3.55 million gallons 

of water and is a source for refilling the CST.

10-15 minutes

• Plant operations personnel complete an initial assessment of plant status.

• RCIC auto-starts to make up for coolant losses and maintain RPV level.

• HPCI might auto-start in response to initial transient, but is secured by operations 

personnel.

• In accordance with SE-11, plant operations personnel initiate the following mitigation 

measures:

o Attempt to line up the Conowingo hydroelectric dam (i.e., station blackout line) 

as an alternative offsite power source, but the line is not available.

o Attempt manual start of emergency diesel generators, but none is available.

o Begin to shed non-essential loads from the emergency DC bus.

50 minutes

• EOF is manned. The EOF is located in the Philadelphia area, far away from the plant. 

Therefore, the timing should not be affected by the seismic event.

1 hour

• Actions to shed non-essential loads from DC bus is complete, battery life extended to an 

estimated 4 hours. 

• RPV depressurization is initiated using one SRV. The target RCS pressure is 125 psi.

1.5 hours 

• The EOF is operational. The EOF reviews actions taken by Operations and determines 

the availability of the remotely located trailer-mounted portable diesel-driven pump 

stored outside of the protected area. Actions recommended by the EOF include the 

following: 

o Use portable power supply for operating SRVs and for RPV level indication.

o Perform RCIC blackstart.

o Use portable diesel driven pump (250 psi, 500 gpm) to provide makeup to RCS, 

Hotwell, CST, and other locations.

o Use portable air supply to manually operate containment vent valves.

o Use pumper truck in place of portable diesel-driven pump.
3-8

1.75 hours 

• Operators assess and concur with EOF recommendations. Operators prioritize 

recommendations based on plant conditions and begin implementation.

2 hours

• Operator assumes remote manual control of RCIC flow.

• TSC is manned. Because of the magnitude of the event, loss of causeway, other potential 

infrastructure failures, and multiple emergency responders located on both sides of the 

river, a 1-hour delay in minimum manning of the TSC was assumed.

2.25 hours

• TSC is assumed operational.

3.5 hours 

• Portable DC power supply (i.e., AC generator operating through an converter) is 

connected to continue operating the SRV to depressurize the RPV.

• RCIC system is manually controlled to limit the use of site batteries and to continue 

providing makeup to the RCS.

Before 10 hours 

• Portable air supply to manually operate containment vent valves is in place and ready for 

operation. 

• Portable diesel driven pump is staged and available for service.

3.2 Short Term Station Blackout

Section 3.2.1 describes the initial status of the plant following the seismic event. Section 3.2.2

then discusses the availability of plant systems to respond to the initiating event. Actions that 

can be taken by plant personnel to mitigate the effects of failed plant safety functions are 

described in 3.2.3. Section 3.2.4 describes two scenarios that differ in the assumed success (or 

failure) of the mitigative actions. Mitigated scenarios are defined as those in which 

the mitigative actions are successful. Unmitigated scenarios are defined as those in which 

certain key mitigation measures are not successfully implemented. 

3.2.1 Initiating Event

The STSBO is initiated by the same spectrum of events that lead to the LTSBO, but is more 

extensive in the amount of consequential damage to plant systems. The most frequent initiators

are large seismic events or internal fires or floods. The seismic event is a major contributor to 

the composite frequency of this sequence and is conservatively used as the basis for defining 

consequential events and conditions at the plant. Damage caused by the earthquake is assumed 

to result in a total loss of offsite power. In addition, all diesel generators fail to start or run as 
3-9

needed, rendering all onsite AC power unavailable. The diesel generators have a shared 

configuration between the two units, which causes power failure to affect both units. However, 

this analysis considers only the response to failures at one of the units. Additionally, the 

earthquake results in failure of all onsite DC power.

3.2.2 System Availabilities

Similar to the LTSBO scenario, reactor scram, reactor isolation and containment isolation 

immediately follow the initiating event. Neither active AC nor DC power is necessary for these 

safety functions to occur. The station blackout electric power line from the hydroelectric station 

downstream of the plant site is also assumed to be unavailable because of structural damage to 

the dam and electric station components. 

The principal difference between this scenario and the LTSBO is that DC power from station 

batteries is also not available. Thus, a total loss of all onsite and offsite electrical power occurs 

immediately following the initiating event rather than several hours later, thereby disabling all 

plant equipment dependent on control or motive power for start-up and operation. 

Loss of all DC power disables electronic start-up and control of steam-driven emergency coolant 

makeup systems (RCIC and HPCI), as well as control and motive power to reactor pressure relief 

valves, which were available for a few hours in the LTSBO. As described in Section 2.3, 

however, plant operations personnel would attempt to blackstart RCIC in this situation. The 

specific actions necessary to accomplish local, manual start-up and operation of RCIC are 

delineated in plant procedures, and the actions are reviewed as part of routine operator training. 

Therefore, successful RCIC blackstart is assumed to occur in the baseline calculation for the 

STSBO. However, a sensitivity calculation was also performed to investigate the ways in which 

failure of RCIC blackstart (i.e., loss of all coolant injection) alters the chronology of severe 

accident progression and the resulting source term. Results of this sensitivity case are also useful 

for comparison to past STSBO analyses.

3.2.3 Mitigative Actions

Manual operation of RCIC under blackstart and blackrun conditions would delay (blackstart) or 

prevent (blackrun) core damage during an STSBO. A calculation assuming RCIC blackstart (but 

not blackrun) is described in Section 5.3. This calculation shows RCIC blackstart delays the 

onset of core damage by more than 5 hours, which is sufficient time to mobilize and align 

equipment added under 10CFR50.54(hh), i.e., the independent diesel-driven pump. The 

independent diesel-driven pump would then be used as a means of direct injection into the RPV 

if RCIC operation could not be sustained via blackrun. Alternatively, long term operation of 

RCIC could be maintained under blackrun conditions if the portable electric generator energized 

the instrumentation that measures and indicates RPV water level. Based on the calculation 

described in Section 5.3, this action would need to be completed within 3.4 hours of the initiating 

event to prevent failure of RCIC due to RPV overfill. The independent diesel-driven pump could 

then be aligned to replenish the CST, thereby maintaining RCIC suction from a source that is not 

adversely affected by the absence of suppression pool cooling and resulting increases 

suppression pool water temperature. Full mitigation (i.e., prevent core damage and long term 

containment heat removal) would result if portable equipment necessary to manually open and 

close a containment vent path is available. This equipment was described in Section 3.1.3 for the 
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LTSBO accident sequence. The resulting plant response would be very similar to the mitigated 

LTSBO described in Section 5.2. 

Procedures and equipment added under 10CFR50.54(hh) (i.e., the independent diesel driven 

pump and a portable electric generator) were not explicitly modeled for the unmitigated STSBO

(i.e., no blackstart or blackrun) because it was judged that insufficient time is available to 

mobilize and align the portable equipment prior to the onset of core damage.

3.2.4 Scenario Boundary Conditions

Two variations of the STSBO scenario were considered. The only difference in the two cases is 

success or failure of actions to manually actuate (blackstart) the steam-driven RCIC system. The 

base case assumes successful blackstart; a sensitivity case examines the effects of failure to 

blackstart RCIC.

3.2.4.1 STSBO with RCIC Blackstart

Blackstart of RCIC during a STSBO requires several manual actions by plant operations 

personnel. These actions include local, manual opening of normally closed valves to admit 

steam from the main steam lines into the RCIC turbine and pump discharge valves to direct 

water into the reactor vessel. 

The baseline STSBO calculation assumes operators successfully complete these actions within 

one hour after the initiating event, at which time coolant flow to the reactor vessel begins. 

Manual actions to regulate steam flow into the RCIC turbine (thereby controlling pump

discharge rate) after blackstart is accomplished are not credited in this scenario because electric 

power to instrumentation needed to monitor reactor coolant level would not be available. As a 

result, the system effectively operates at a constant flow rate equivalent to the rated capacity of 

the system (i.e., 600 gpm). Because this flow rate is greater than the rate required to make up for 

evaporative losses, the reactor water level rises above nominal and eventually overfills the 

reactor vessel. In this context, ‘overfill’ means that the reactor water level increases to the 

elevation of the main steam line nozzles, allowing water to spill into the steam lines and causing 

them to flood with water. The steam supply line for the RCIC turbine connects to the main 

steam line at a low elevation (adjacent to the inboard MSIVs). Therefore, water spilling over 

into the main steam lines blocks or flows toward the RCIC turbine, causing the system to cease 

functioning. RCIC blackstart without effective RPV level control is therefore a temporary 

measure for managing reactor coolant inventory, and core damage occurs approximately 6 hrs

later than would be observed in a case without RCIC blackstart.

Results of the STSBO with RCIC blackstart are described in Section 5.3.

The following is the timeline of events and operator actions that were credited in the STSBO 

scenario with successful RCIC blackstart.

Event Timeline for STSBO with Successful RCIC Blackstart

Event Initiation and Initial Plant Response

• Large seismic event results in a loss of offsite power.
3-11

• All diesel generators fail to start.

• DC power (station batteries) and/or associated emergency buses are not available.

• Reactor trips (successful scram).

• Reactor pressure vessel isolates (i.e., MSIVs close) and containment isolation valves 

close.

• The CRDHS, LPCI mode of the RHR system, standby liquid control, condensate, 

containment cooling and containment spray systems are not available.

• Control room receives indication that plant is in a station blackout condition requiring the

operator to enter Special Event Procedure SE-11, “Station Blackout Procedure.”

15 minutes

• Plant operations personnel complete an initial assessment of plant status.

• Plant operations personnel begin to implement actions to blackstart RCIC.

• In accordance with SE-11, plant operations personnel initiate the following mitigation 

measures:

o Attempt to line up the Conowingo hydroelectric dam (i.e., station blackout line) 

as an alternative offsite power source, but the line is not available.

o Attempt manual start of emergency diesel generators, but none is available.

1 hour

• RCIC is successfully started and begins to inject water from the CST at rated capacity

(600 gpm).

3.2.4.2 STSBO without RCIC Blackstart

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. Section 5.4 describes results of the sensitivity calculation.

3.3 Loss of Vital Alternating Current Bus E-12

The scenario is initiated by the loss of vital AC bus E-12. It was initially estimated to have a 

frequency above the SOARCA screening criterion of 1x10-6 pry. However, after further review 

of the SPAR model and comparison with the licensee’s PRA, the scenario was determined to 

have a CDF below the screening criteria. Because the MELCOR analysis provided unique 

insights into the response of the plant to an internal event sequence, the MELCOR analysis was 

retained.
3-12

Section 3.3.1 describes the initial status of the plant following the initiating event. The key 

system availabilities during the course of the accident are summarized in Section 3.3.2. The 

pertinent mitigative measures available to address the accident progression are described in 

Section 3.3.3. Section 3.3.4 describes various scenarios based on the success of the mitigative 

actions. Mitigated scenarios are defined as those in which the mitigative actions are successful. 

Unmitigated scenarios are defined as those in which certain key mitigate measures are not 

successfully implemented. 

3.3.1 Initiating Event

The initiating event for this scenario is a loss of Division I vital AC bus E-12. Loss of power 

through this bus eliminates power that sustains power to the Division I DC bus (through the 

battery charger). The Division I DC bus would continue to remain energized for the lifetime of 

the batteries, which is expected to be a minimum of 2 hours.

3.3.2 System Availabilities

Loss of one vital AC bus disables motive or control power to some plant equipment, but not all. 

For example, power to the instrument and control air system would be lost, and the converters 

that charge the station batteries would not function. However, other AC buses would direct 

motive power to the RHR and core spray pumps, permitting use of low-pressure coolant 

injection. One of the two CRDHS pumps would also remain available.

Steam-driven injection systems (HPCI and RCIC)7 operate as long as station batteries 

deliver DC power to control system components. Station batteries also facilitate manual control 

of SRVs. When battery power is depleted, HPCI, RCIC, and SRV controls are assumed to be 

lost. 

The shut-down cooling mode of RHR would not be available because of loss of AC power 

disables valves needed to align the system for that configuration. However, the system can be 

aligned to operate suppression pool cooling and drywell sprays.

The duration of DC power is treated as an uncertain parameter in this scenario. The licensee 

PRA uses a value of 2 hours, which is the minimum value and represents the worst possible 

condition. The licensee’s engineering judgment is that batteries can last 4 hours with effective 

DC load shedding. As described in Section 5.6.3, a precise value is not particularly important, 

provided battery duration is greater than 3 hours.

3.3.3 Mitigative Actions

This event was shown to be satisfactorily mitigated without crediting any of the security

related mitigative actions mentioned in Section 3.1.3. Therefore, although scenario boundary 

conditions were determined for a mitigated case (see Section 3.3.4.2), a thermal-hydraulic 

accident progression analysis was not performed for the mitigated scenario. Further, the base 

 

7 Although RCIC is available in all the standard plant analysis risk cut sets for this sequence, HPCI is 

disabled due to independent failures in some of the sequences. Availability of HPCI is not important in this 

sequence and is neglected.
3-13

case among the unmitigated cases, as well as many of the sensitivity calculations described 

below, did not result in core damage as described later in Section 5.6.

It should also be noted that the licensee ran the Loss of E12 Bus scenario on their plant simulator 

in November 2011. The licensee offered the following description of the observed response:

The Training Instructor that ran the simulation determined that the immediate impact of 

the bus loss did not result in a plant scram, only a one-half logic primary containment 

isolation (1⁄2 PCIS isolation); however, loss of the bus does cause all condenser hotwell 

indication to fail and condenser hotwell makeup valves to go full open. This eventually 

results in a loss of vacuum due to the high hotwell level covering the condenser tubes.

Attempts to control level manually are not an option because you do not have indication 

of hotwell level. Additionally, vacuum will eventually be lost due to the ‘A” Steam Jet 

Air Ejector (SJAE) pressure controller failing to 0 (maybe a simulator issue but could be 

a plant issue also) requiring a manual scram before receipt of the automatic scram (per 

procedure). During the simulation, attempts were made to swap to the "B" SJAE 

controller but the "B" SJAE train could not be placed in service apparently due to the loss 

of power was not allowing the valve interlock logic to work.

3.3.4 Scenario Boundary Conditions

Section 3.3.4.1 lists the sequence of events to be prescribed in the unmitigated accident scenario 

for loss of AC bus E-12 accident scenario. Section 3.3.4.2 summarizes the sequence of events in 

the mitigated case.

3.3.4.1 Unmitigated Cases

Several unmitigated cases were considered, which differ only in terms of the assumed duration 

of station batteries. Unmitigated cases, as noted above, generally did not result in core damage. 

Variations in station battery duration affected the length of time RCIC was available for 

controlled coolant makeup and an SRV could be held in the open position to maintain lower

RPV pressure. As noted later, all cases in which station batteries continued to provide control 

power to RCIC for at least 3 hours were found to avoid core damage. The unmitigated cases did 

not credit mitigation measures developed under the requirements of 10CFR50.54(hh), such as the 

staging and alignment of a back-up power supply and portable coolant injection pump. 

The timeline following the initiating event is summarized below.

Unmitigated Event Timeline

Event Initiation and Initial Plant Response

• Loss of Division I AC bus E-12.

• Loss of all AC powered coolant injection except one CRDHS pump.

• Successful reactor trip, reactor and containment isolation.
3-14

• DC power (station batteries) functional.

• RCIC auto-initiates when reactor level drops to low-level setpoint. Suction is initially 

aligned to the CST.

• When level rises to operating range, operator takes manual control of RCIC to maintain 

RPV level.

15 minutes

• Initial Operations assessment of plant status is complete.

• RCIC is operating, maintaining RCS level.

• In accordance with SE-11, actions to shed non-essential loads from the emergency DC

bus are initiated.

1 hour

• DC load shedding is complete, extending battery life from 2 to 4 hours. 

• Also available is access to the CRD throttle valve to increase flow from 110 gpm to 

140 gpm without reactor depressurization. 

1.25 hours

• TSC is fully operational.

• EOF is fully operational.

1.5 hours

• Operators initiate RPV depressurization by opening one SRV. Target RCS pressure is 

approximately 125 psi.

4 hours

• DC power from station batteries is exhausted.

• Open SRV re-closes.

• RCIC stops operating due to turbine overspeed trip.

Parameters Varied in Sensitivity Calculations

Section 5.6 describes results of several calculations of plant response to the unmitigated scenario. 

Several calculations were performed to quantitatively assess the effects of uncertainties in some 

of the boundary conditions for this scenario. Among the boundary conditions are:

• RPV depressurization: The rate of coolant flow delivered by the single, operating 

CRDHS pump is affected by RPV pressure. The nominal case assumes operators follow 

EOP guidance to depressurize the RPV. This assumption increases the delivered CRDHS 

flow. Sensitivity calculations were performed assuming actions to depressurize the RPV 

were not successful, resulting in lower CRDHS flow.
3-15

• Managing CRDHS flow: plant-specific procedures instruct operators to maximize 

CRDHS flow if the system serves as an alternate means of RPV injection. In this case, 

CRDHS is the only system available for coolant injection after station battery exhaustion. 

Sensitivity calculations were therefore performed to examine the effects of not increasing 

CRDHS flow from nominal values to maximum flow conditions.

• Duration of DC power: Battery lifetimes of 2, 3, 4, and 6 hours are analyzed to 

determine the extent to which sustained operation of RCIC is necessary to prevent core 

damage.

3.3.4.2 Mitigated Case

Implementation of the mitigation measures developed in response to 10CFR50.54(hh) provides 

diverse and redundant means of maintaining coolant injection beyond the systems that continue 

to operate in the unmitigated cases. The following is a timeline for implementing these and 

other actions after the initiating event.

Mitigated Event Timeline

Event Initiation

• Loss of Division I AC bus E-12.

• Loss of all AC powered coolant injection except one CRDHS pump.

• Successful reactor trip, reactor and containment isolation.

• DC power (station batteries) functional.

• RCIC auto-initiates when reactor level drops to low-level setpoint. Suction is initially 

aligned to the CST.

• When level rises to operating range, operator takes manual control of RCIC to maintain 

RPV level.

• Standby liquid control system is available but neglected because its cooling injection 

flow of 50 gpm is not necessary.

• Drywell spray is available, but neglected because it is not necessary.

15 minutes

• Initial Operations assessment of plant status is complete.

• RCIC is operating, maintaining RCS level.
3-16

• In accordance with SE-11, actions to shed non-essential loads from the emergency DC

bus are initiated.

50 minutes

• TSC staffing is underway. Their primary function would be to review initiating event, 

plant status, and operator action to provide guidance on alternative mitigative measures.

• EOF staffing is underway. Their primary function would be to review initiating event, 

plant status, and operator action to provide guidance on alternative mitigative measures. 

The primary users of SAMGs and Extreme Damage Mitigation Guidelines (EDMGs) are 

the Tech Manager, Operations Manager, and Emergency Director who are trained on 

SAMGs and EDMGs.

1 hour

• DC load shedding is complete, extending battery life from 2 to 4 hours. 

• Also available is access to the CRD throttle valve to increase flow from 110 gpm to 

140 gpm without reactor depressurization. 

1.25 hours

• TSC is fully operational.

• EOF is fully operational.

1.5 hours

• Operators initiate RPV depressurization by opening one SRV. Target RCS pressure is 

approximately 125 psi.

• TSC and/or EOF review actions taken by Operations and determines the availability of

the remotely located equipment. Recommend the following actions: 

o Portable power supply connected through an converter to the emergency DC bus 

to ensure power necessary to hold SRV open .

o Portable diesel driven pump (500 gpm at 250 psi) is available for makeup to the 

RCS, Hotwell, or CST.

o Portable air supply is available to manually operate containment vent isolation 

valves, as required.

o Portable diesel driven pump is available to inject water into the drywell via RHR 

or the RCS.

1.75 hours 

• Operations staff assesses and concurs with TSC’s and/or EOF’s recommendations. 

• Operations staff prioritizes recommendations based on plant conditions and begins

implementation.

4 hours 

• Manual operation of the RCIC to sustain RCS level after battery depletion.
4-1

4.0 MELCOR MODEL OF THE PEACH BOTTOM PLANT

This section summarizes the MELCOR model of the Peach Bottom Atomic Power Station. A 

comprehensive description of the model is available in separate documentation.

The MELCOR Peach Bottom model was originally generated for code assessment applications 

with MELCOR 1.8.0 at Brookhaven National Laboratories. The model was subsequently 

adopted by J. Carbajo at Oak Ridge National Laboratories to study differences between fission

product source terms predicted by MELCOR 1.8.1 and those generated for use in NUREG-1150, 

“Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants,” issued December 

1990, using the Source Term Code Package [1]. In 2001, Sandia National Laboratories refined

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 core 

(in terms of fuel and structural material and hydrodynamic volumes) used to calculate in-vessel 

melt progression. However, the overall scope of the model also expanded to permit a wider 

spectrum of accident scenarios to be examined, some of which involved operation or delayed 

failures of plant safety systems.

These developments culminated in a model that was applied in the reassessment of radiological 

source terms for high burnup and MOX core designs, and a comparison of their release 

characteristics [5] to the regulatory prescription outlined in NUREG-1465, “Accident Source 

Term for Light-Water Nuclear Power Plants,” issued February 1995 [6]. These calculations 

addressed a wide spectrum of postulated accident sequences, which required the following new 

models to represent diverse plant design features:

• Modifications of modeling features needed to achieve steady-state reactor conditions 

(e.g., recirculation loops, jet pumps, steam separators, steam dryers, feedwater flow, 

CRDHS, main steam lines, turbine/hotwell, core power profile), 

• New models and control logic to represent coolant injection systems ( e.g., RCIC, HPCI, 

RHR, LPCS) and supporting water resources (e.g., CST with switchover), and

• New models to simulate reactor vessel pressure management (e.g., SRV, safety valves, 

ADS, and logic for manual actions to affect a controlled depressurization if torus water 

temperatures exceed the heat capacity temperature limit).

Subsequent work in support of other NRC research programs motivated further refinement and 

expansion of the model in two broad areas. The first area focused on the spatial representation of 

primary and secondary containment. The drywell portion of primary containment has been 

subdivided to distinguish thermodynamic conditions internal to the pedestal from those within 

the drywell itself. Refinements have also been made to the spatial representation and flow paths 

within the reactor building (i.e., secondary containment). The second area has focused on 

bringing the model up to current best practice standards for MELCOR 1.8.6. Table 4-1 provides 

a brief summary of plant design parameters that are helpful in comparing the configuration of 

Peach Bottom to other reactors of interest.
4-2

Table 4-1 Important Design Parameters for Peach Bottom

Parameter Value

(SI units)

Value

(British units)

Rated Core Power [MWth] 3514

Number of Fuel Assemblies in Core 764

Assumed Average Specific Power [MWD/MTU] 25.5

Fuel (UO2) Mass [kg / lb] 155,500 342,800

Zircaloy Mass in Fuel Cladding [kg / lb] 40,580 89,500

Zircaloy mass in Fuel Channels [kg / lb] 19,600 43,200

RPV Inner Diameter [m / ft] 6.4 21.0

RPV Height [m / ft] 22.2 72.9

RPV Wall Thickness (core mid-plane) [cm / in] 16.4 6.4

Containment Design Pressure [MPa / psig] 0.49 56

Drywell Free Volume [m3 / ft3] 4,980 176,000*

Nom. Wetwell Free Volume (airspace) [m3 / ft3] 3,570 126,000

Nom. Suppression Pool Water Volume [m3 / ft3] 3,570 126,000

Reactor Building Free Volume [m3 / ft3] 75,200 2,655,600

Condensate Storage Tank Water Volume [L / gal] 757,082 200,000

RCIC Rated Flow [tonne per hr / gpm] ** 136 600

CRDHS Flow [tonne per hr / gpm]**

 Nominal (during full-power reactor operation)

 Post-scram

 Maximum capacity per pump

14

25

32

63

110

140

* Including the volume of wetwell-drywell vent pipes

** Values with reactor at nominal pressure.

4.1 Reactor Vessel and Coolant System

Excluding the core region, the RPV is represented by seven hydrodynamic control volumes, nine 

flow paths, and 24 heat structures. Nodalization for the core region between the core top guide 

and the bottom of active fuel are described in detail in Section 4.2. Figure 4-1 illustrates the 

reactor vessel nodalization by comparing the actual vessel design (on the left) to the MELCOR 

control volume representation (on the right). In Figure 4-1, control volumes are indicated by 

‘CV’ followed by the three-digit control volume number, and flow paths are indicated by ‘FL’ 

followed by the three-digit flow path number.

Appended to the MELCOR control volumes representation of the RPV shown in Figure 4-1 are 

several additional control volumes and flow paths representing a variety of reactor support 

systems, such as:

• reactor recirculation piping,

• main feedwater and steam lines, and 

• connections to emergency coolant injection and heat removal systems.

The MELCOR representation of the entire reactor coolant system is illustrated in Figure 4-2. 

Collectively, these ancillary systems permit the model to properly calculate steady state, as well 

as a wide variety of transient conditions. To optimize numerical performance of this model, 

some parallel lines or trains of certain systems have been consolidated. For example, the four 
4-3

main steam lines have been represented by two parallel ‘lines,’ one of which represents the 

single steam line containing the lead (i.e., lowest set point) SRV, and the second, the composite 

geometry of the remaining three lines. Isolating the steam line with the lead SRV permits the 

proper geometry (e.g., internal volume, structural surface area) to be represented for fission 

product transport from the reactor to the suppression pool during accident sequences in which 

fuel damage begins while the reactor vessel is at high pressure and pressure relief is 

accomplished by SRV operation.

Figure 4-1 Reactor vessel cross-section detail and MELCOR hydrodynamic 

nodalization.

607.5 in

635. in

739. in

745. in

125. 1/2 in

749. 13/16 in

658.5. in

379 1/8 in

RV wall

251 in diam

498.5 in

Feedwater

Main Steam

steam shroud:

245 in OD

Recirculation Nozzles 211 9/16 in

Top of Jet

Pumps

318 in

Inlet 181 in.

Outlet 161.5 in.

121 9/16 in

0.0 in

CV355

CV320

CV310

CV300

FL355

FL352

FL351

FL312

FL383

Recirc-B

FL382

Recirc-A

FL387

Drive flow - B

FL386

Drive flow - A

Suction flow - B

FL311

Suction flow - A

FL310

FL312

CV345

FL345

CV360

FL356

FL357

Feedwater

FL376

Main Steam

B, C & D

FL370

Main Steam A

FL373

CV350
4-4

4.2 In-vessel Structures and Reactor Core

Structures within the RPV are described in the next three sections. First, Section 4.2.1 describes 

the general configuration of the spatial nodalization of the core and structures below the core. 

Certain aspects of this nodalization, and some characteristics of the MELCOR model of material 

degradation, are tailored to the unique design features of BWRs. These are briefly summarized 

in Section 4.2.2. Finally, Section 4.2.3 mentions the manner in which in-vessel structures above 

the core are treated in the MELCOR model.

4.2.1 General Configuration of MELCOR In-vessel Nodalization

In MELCOR, the region tracked directly by the COR package model includes a cylindrical space 

extending vertically downward along the inner surface of the core shroud from the core top guide 

to the reactor vessel lower head. It also extends radially outward from the core shroud to the 

hemispherical lower head in the region of the lower plenum below the base of the downcomer, 

preserving the curvature of the lower head from this point back to the vessel centerline. 

The core and lower plenum regions are divided into concentric radial rings and axial levels. 

Each core cell may contain one or more core components, including fuel pellets, cladding, 

canister walls, supporting structures (e.g., the lower core plate and control rod guide tubes), 

non-supporting structures (e.g., control blades, the upper tie plate, and core top guide) and once 

fuel damage begins, particulate and molten debris.

The spatial nodalization of the core is shown in Figure 4-3 and Figure 4-4. The entire core and

lower plenum regions are divided into six radial rings. As indicated in Figure 4-4, rings one, 

two, three, four, and five represent 112, 160, 200, 168, and 124 fuel assemblies, respectively. 

The radial distance between each of the five rings is not uniform. The radius of each ring was 

defined so as to preserve the radial power distribution in the Unit 2 core, based on plant 

operating data from four recent and consecutive operating cycles. Radial ring 6 represents the 

region in the lower plenum outside of the core shroud and below the downcomer. Ring 6 exists 

only at the lowest axial levels in the core model. 
4-5

Figure 4-2 Spatial nodalization of reactor pressure vessel and coolant system FL387 LOOP B FL383 FL385 LOOP A FL384 FL382 FL614 CST CV360 Steam Dome CV320 Lower Plenum FL304 CV310 Downcomer CV310 Downcomer

CV383

CV385

CV384

CV382

FL386

FL312 FL312

CV300

CV300

FL310 FL311

FL356

FL357

FL370

CV391

Turbine / Hotwell

(time independent)

FL371

Inboard

MSIVs

FW

Pump

CV392

Feedwater source

(time independent)

CRD

Pump

RCIC

Core Spray

(Loop 1)

CV600

FL372

FL702

FL706

FL606

CV370 (Steam Lines B, C & D)

FL399

(RV Breach)

CV350

 Steam Separators

SVs

(2 valves)

FL910

(Wetwell Vent)

to drywell

atmosphere

Drywell

CV220

CV901

Environment

FL360 - FL362,

FL364 - FL369

FL358

FL653

RHR-2 RHR-1

HPSW

FL648

FL649

FL652

SPRSR01

CV652

CV651

CV650

CV700

Reactor

Building RWCU return

(not modeled)

CV374 CV372

FL611

FL376 FL374

CV220

CV653

FL656 (LPCI)

FL655(supp.poolcooling)

FL657(LPCI)

FL654(supp.poolcooling)

CV394

(time

independent

keep-full at

steady state)

S/RV (typical)

9 parallel valves

Turbine

Building

FL600

FL613

CV611

FL351

FL699

FL651

FL650

A C B D

CV355

Steam Dryers

FL355 FL352

FL704

SPRSR02

CV631

HPCI

FL633

FL631

FL626

FL634

CV345

 Shroud Dome

FL345

CV373 (Steam Line A) CV375 (Steam Line A)

Inboard

MSIV FL377 FL375 FL373

FL359,

FL363

S/RV (typical)

2 parallel valves
4-6

Figure 4-3 Spatial nodalization of the core and lower plenumCV514

CV524

CV534CV544CV554

CV345

Shroud Dome

CV320

Lower 

Plenum

CVH Nodalization accompanying

COR Nodalization

CV111CV121

CV131

CV141CV

CV

151

511

CV521

CV531

CV541

CV551

CV512

CV522

CV532

CV542CV552

CV112

CV122

CV132

CV142CV152

CV513

CV523

CV533

CV543CV553

CV113

CV123

CV133

CV143CV153

CV114

CV124

CV134

CV144CV154

CV515CV525

CV535

CV545

CV555

CV115CV125

CV135

CV145CV155

BWR/4 Core Nodalization

101 201 301 401 501

Top Guide &

Upper Tie Plate

Lower Tie Plate 

& Nose Pieces

105 205 305 405 505

Height of Fuel Assemblies

Heated Fuel Length

Lower Core Plate & 

"Elephant's Foot"

Control Rod 

Guides Tubes

106 206 306 406 506

117 217 317 417 517

103 203 303 403 503

102 202 302 402 502 108 208 308 408

508

107

207

307

407507

109

209

309

409509

110

210

310

410510

111

211

311

411511

112

212

312

412512

113

213

313

413513

114

214

314

414514

115

215

315

415515

116

216

316

416516

Penetration - R1

Control Rod 

Stub Tubes

Penetration - R2

Penetration - R3

Penetration - R4

Penetration - R5

Lower head

TAF

BAF

0.0 m

1.295 m

1.892 m

2.490 m

3.088 m

5.267 m

5.494 m

5.875 m

6.256 m

6.637 m

7.018 m

7.399 m

7.780 m

8.161 m

8.542 m

8.923 m

9.304 m

9.630 m

104 204 304 404 504

603

604

602
4-7

Figure 4-4 Local relative power fraction and five ring radial boundaries of core

The core and lower plenum are divided into 17 axially stacked levels. The height of a given 

level varies but generally corresponds to the vertical distance between major changes in the flow 

area, structural materials, or other physical features of the core (and below core) structures. 

Axial levels 1 through 5 represent the open space and structures within the lower plenum. 

Initially, this region has no fuel and no internal heat source but contains a considerable mass of 

steel associated with the control rod guide and in-core instrument tubes. During the core 

degradation process, the fuel, cladding, and other core components displace the free volume 

within the lower plenum as they relocate downward in the form of particulate or molten debris.

Axial level 6 represents the steel associated with fuel assembly lower tie plates, fuel nose pieces, 

and the lower core plate and its associated support structures. Particulate and molten debris

formed by failed fuel, canister, and control blades above the lower core plate will be supported at 

this level until the lower core plate yields. Axial levels 7 through 16 represent the active fuel

region. All fuel is initially in this region and generates the fission and decay power. Axial level 

17 represents the nonfuel region above the core, including the top of the canisters, the upper tie 

plate, and the core top guide.

radius > 0 0.153 0.305 0.458 0.610 0.763 0.915 1.068 1.220 1.373 1.525 1.678 1.830 1.983 2.135 2.288 (m)

v 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30

0 center

0.153 16

0.928 1.306 1.088 1.258 1.045 1.413 1.094 1.344 1.069 1.440 1.090 1.162 0.727 0.528 0.270

0.305 17 1.305 1.100 1.373 1.065 1.395 1.207 1.468 1.209 1.463 1.303 1.421 1.225 0.818 0.536 0.272

0.458 18

1.084 1.367 1.156 1.389 1.182 1.471 1.246 1.487 1.208 1.496 1.236 1.273 0.791 0.531 0.268

0.610 19 1.253 1.044 1.365 1.064 1.341 1.134 1.482 1.215 1.454 1.268 1.403 1.156 0.756 0.510 0.257

0.763 20 1.040 1.368 1.159 1.335 1.120 1.463 1.242 1.448 1.203 1.465 1.098 1.110 0.704 0.484 0.241

0.915 21 1.410 1.199 1.460 1.128 1.459 1.246 1.500 1.163 1.480 1.323 1.337 1.121 0.700 0.453 0.215

1.068 22 1.092 1.462 1.240 1.473 1.233 1.489 1.235 1.455 1.250 1.378 0.995 0.980 0.620 0.336 0.168

1.220 23 1.343 1.205 1.479 1.206 1.431 1.129 1.417 1.145 1.258 1.139 1.014 0.724 0.450 0.246

1.373 24 1.068 1.459 1.201 1.443 1.186 1.432 1.203 1.245 0.954 0.945 0.720 0.522 0.295

1.525 25 1.439 1.299 1.489 1.257 1.447 1.297 1.352 1.128 0.942 0.775 0.498 0.306 0.172

1.678 26 1.089 1.417 1.231 1.394 1.087 1.320 0.981 1.007 0.716 0.497 0.314 # assemblies

1.830 27 1.161 1.222 1.268 1.152 1.104 1.112 0.970 0.707 0.517 0.304 112 1.223

1.983 28 0.727 0.813 0.788 0.754 0.701 0.695 0.615 0.445 0.291 0.171 160 1.313

2.135 29 0.528 0.532 0.530 0.508 0.482 0.450 0.333 0.244 200 1.275

2.288 30

0.270 0.271 0.268 0.256 0.240 0.214 0.166 168

0.746

(m) 124 0.295

764

RING 4

RING 5

3-cycle

average

RPF

RING 1

RING 2

RING 3
4-8

4.2.2 Treatment of Unique Design Features of a BWR Core

Several design features of the BWR core, and associated structures located below the core, merit 

special attention in modeling severe accidents. These are not discussed exhaustively here, but a 

few are mentioned to illustrate the way in which the modeling approach to this type of reactor 

differed from the approach used to model a Pressurized Water Reactor (PWR), or other designs.

Fuel Channels and Control Blades

Each BWR fuel assembly in the core is shrouded by a solid rectangular channel box, which 

confines coolant flow to that single assembly. Therefore, cross-flow between adjacent 

assemblies is not possible in a BWR unless a pair of adjacent channel boxes fails. Figure 4-5

illustrates this configuration for a module of four fuel assemblies, which surrounds a single 

control blade. Unlike control rod clusters in a typical PWR, which are inserted directly into 

certain fuel assemblies, the control blade in a BWR is inserted into the interstitial space between 

(and outside) adjacent fuel channels.

The MELCOR hydrodynamic model for this configuration recognizes the vertical constraint 

placed on flow through the core unless or until structural damage to fuel channels occurs. The 

in-core heat transfer model also accounts for lateral (or radial) differences in the materials 

adjacent to ‘bladed’ sides of a channel (i.e., sides neighboring a control blade) versus ‘unbladed’

sides, which communicate only with a neighboring fuel channel. As fuel temperature rise in the 

core during an accident simulation, axial and radial heat transfer calculations account for 

oxidation of Zircaloy fuel cladding and the Zircaloy channels. Separate failure criteria are used 

to ascertain when highly oxidized fuel rods and highly oxidized channels can no longer maintain 

their normal upright configuration and collapse into particulate debris. Failure criteria for fuel 

rods and channels are discussed briefly in Section 4.8.1.

Radial cross-flow from one ring of the core nodalization to a neighboring ring is possible if the 

channels in both rings melt or collapse to open a flow path in the radial direction. As indicated 

by in the nodalization diagram on the right-hand side of Figure 4-3, radial flow out of the interior 

region of the channels within a particular must first flow into the neighboring core bypass area 

(i.e., the space between channels normally occupied by control blades.) If the stainless steel clad 

control blades and Zircaloy channel boxes in the neighboring fuel channel also melt or collapse, 

fluid could continue to flow radially out of the bypass region into the neighboring channel, and 

so on. Circular, natural circulation flow patterns within the core region can, therefore, only 

occur after a sufficient number of channel boxes (properly distributed) melt, or otherwise 

relocate downward, creating open space in the radial direction. Logical control functions in the 

MELCOR model track the structural status of channel boxes at each cell (node) in the core to 

determine whether flow paths in the radial direction should be open or closed.
4-9

Notable BWR Core Design Features

1. Core top guide

3. Fuel assembly upper tie plate

6. Channel box

7. Cruciform control blade

8. Fuel rod

9. Spacer

10. Core plate

Figure 4-5 Module of Four BWR Fuel Assemblies and Associated Channel Boxes8

The axial nodalization of the core is designed, in part, to account changes in material 

composition and mass along the axial length of a typical fuel assembly. For example, Figure 4-5

does not explicitly show the fact that some BWR fuel assembly designs (modern 10x10 

assemblies, for example) incorporate fuel rods of different length within a single assembly. As a 

result, the amount of UO2, and other constituents can differ at the top of an assembly from the 

bottom. 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. 

Lower Plenum Structures 

The COR Package in MELCOR, which models the oxidation, melting, and downward relocation 

of overheated core materials, extends downward below the lower core plate into the RPV lower 

plenum. Molten core debris that flows, or particulate debris that falls, under gravity into the 

lower plenum collects as a mixed debris bed on the inner surface of the lower head. In a BWR, 

this region is filled with a forest of vertical cylinders that house the drive mechanisms for core 

control blades as well as in-core instrumentation. Figure 4-6 depicts a typical configuration of a 

control rod guide tube (CRGT) this region of the RPV.

 

8 Illustration courtesy of General Electric.
4-10

The MELCOR model for Peach Bottom accounts for the structural mass and surface area of the 

CRGTs, which participate in heat transfer within the debris bed that accumulates around them. 

The physical space occupied by intact CRGTs displaces volume available for debris to occupy in 

the lower plenum.

9

 Therefore, debris accumulates to an elevation higher than the value that 

would be calculated by neglecting the volume of the CRGTs. Debris heat transfer, and 

continued oxidation of metallic components within the lower head, accounts for the heat capacity 

and stainless steel composition of the CRGTs. The CRTGs are modeled as vertical columns, 

which fail (buckle) when load carried by the CRGTs exceeds their residual strength, which 

decreases as heat is transferred from core debris. The calculation of CRGT integrity is 

performed on a radial ring-by-ring basis, with 5 rings in the Peach Bottom model as indicated in

the diagram on the left-hand side of Figure 4-3.

Figure 4-6 BWR Lower Plenum Structures

Collapse of the CRGTs has two important effects on the composition and mass of debris in the 

lower plenum prior to lower head failure. The first, and most important, is that all core rested on 

or positioned above the lower core plate within that ring collapses as particulate debris into the 

lower plenum. This fragmented core debris is not uniformly mixed with the debris bed already 

in the lower plenum, but is added to the top of the debris bed according to the spatial 

nodalization shown in Figure 4-3. Therefore, the material composition and porosity of debris 

within the debris bed can vary considerably within the 2-dimensional spatial nodalization of the 

lower plenum. Second, the material composition of the CRGTs themselves is added to the debris 

bed at the location where the intact components were originally positioned. This has the 

tendency to increase the surface area of metallic components in the lower plenum because the 

 

9 Debris relocating to the lower plenum is assumed to accumulate in the volume between intact CRGTs, and 

not within them, based on observations from melt relocation experiments for BWR geometries (e.g., see 

NUREG/CR-6527).
4-11

surface area of particulate debris is generally higher than the surface area of the intact CRGT 

columns.

Thermal interactions between molten core debris and penetrations through the lower head are not 

explicitly modeled in the Peach Bottom SOARCA calculations. A simple, lumped parameter 

model for bulk heat transfer to a penetration assembly is available in MELCOR, but the model is 

not sufficiently refined to calculate multi-dimensional heat transfer and phase change that would 

occur in the neighborhood of a CRGT or instrument penetration; or to calculate molten material 

drainage into an open penetration, as could occur in a BWR RPV drain line. This limitation in 

MELCOR, combined with observations from several large-scale experimental programs that 

examined failure mechanisms for failure of lower head with penetrations [16], led to a modeling 

approach that focuses on creep rupture of the hemispherical lower head as the dominant 

mechanism for lower head failure. This approach is supported by the observation that none of 

the MELCOR calculations predict elevated RPV pressure at the time debris temperatures are 

sufficiently high to challenge lower head integrity. That is, detailed analysis of lower head 

penetration failure [17] suggest internal pressures greater than those obtained in the SOARCA 

calculations are necessary to eject a penetration assembly from the lower head.

4.2.3 In-vessel Structures above the Core

The COR Package in MELCOR does not explicitly model the mechanical response 

(i.e., potential material melting or collapse) of structures above the core top guide. This is an 

inherent limitation in the architecture of MELCOR. As indicated in Figure 4-1, several large 

steel structures are positioned above the core, which can absorb a substantial amount of energy 

carried away from the core during the early periods of core damage. They also represent a 

significant surface area for deposition of fission product aerosols released from the core. These 

structures most notably include the upper shroud dome, steam dryers and separators.

The thermal response and fission product deposition properties of these structures are modeled 

using 1-dimensional heat structures. This modeling approach, combined with the nodalization 

and connectivity of control volumes above the core, provides sufficient spatial resolution to 

calculate the time-dependent temperature response of each structure. A limitation of this 

modeling approach is that the mechanical response of these structures, material melting or 

collapse and the potential for incorporation of steel into core debris, is not modeled. Changes in 

the flow area through this region of the RPV that might be caused by changes in structure 

geometry are also not modeled.

4.2.4 Leakage through In-core Instrument Tubes

Among the comments offered by the independent peer review panel that reviewed the models 

and calculations described in this report was one that questioned whether failure of in-core 

instrument tubes (i.e., due to melting of neighboring fuel assemblies) could create a pathway for 

fission product transport outside the containment pressure boundary that should be considered in 

the SOARCA analyses. The proposed fission product leakage pathway involves leakage through 

traverse in-core probe (TIP) guide tubes, which run from various locations in the core, through 

penetrations in the RPV lower head and the containment pressure boundary, and terminate at TIP 

drive mechanisms located in the reactor building. Leakage through TIP guide tubes would 
4-12

represent a containment bypass pathway, it the tubes are not isolated. Figure 4-7 illustrates the 

TIP guide tube pathway.

Plan view

Figure 4-7 Potential containment bypass transport pathway through open TIP guide 

tubes

TIP Guide Tube Configuration

The TIP system drives three probes of fissile material into various locations in the core to 

calibrate in-core instrumentation. The probes are normally stored in a shield chamber located 

outside the containment pressure boundary. As illustrated in Figure 4-8, each probe can be 

driven out of its shield chamber, through a guide tube, to one of several locations within the core. 

The probe is pushed through its guide tube by a steel cable. Each TIP guide tube enters an 

indexing unit located in the drywell. The indexing units guide a TIP probe into one of several 

exit tubes, each of which are connected to particular local power range monitors (LPRMs) in the 

core. Therefore, the indexing units serve as TIP tube multipliers, allowing three tubes to serve 

the calibration needs of multiple in-core instruments. 

The diameter of TIP probe is approximately 0.211 inches. The internal diameter of the TIP 

guide tube is 0.280 inches and the diameter of the drive cable is 0.258 inches. Therefore, the 

TIP guide tubes

TIP drive

mechanism

TIP room
4-13

available cross-sectional area for flow through an operating guide is small (0.009 in2 per tube is 

the probe is inserted, and 0.06 in2 if the probe is withdrawn into its shield chamber).

Figure 4-8 Layout of the TIP system

The TIP guide tubes can be isolated with either a motor-operated (AC powered) globe valve or a 

squib-actuated shear valve. The globe valve functions only when the TIP probe is fully 

withdrawn because the drive cable for the in-core probe runs through the valve body. If, for 

some reason, the probe cannot be retracted, the squib (explosive) operators on the shear valves 

can be manually actuated to shear the drive cable and seal the guide tube. The squib actuators 

require DC power to operate. The explosive operators on the squib valves could, in principle, be 

actuated locally using a portable dc power supply. However, procedures for this action are not 

formally implemented for this action.

TIP System Operation

Operating experience at Peach Bottom suggests the TIP system is used to calibrate in-core 

instrumentation once every four months. The calibration exercise takes approximately 1 hour to 

complete and involves simultaneous operation of all three TIP probes. Therefore, an individual

TIP guide tube is open (i.e., not insolated), on average, a small fraction of the time (i.e., 1 hour of 

each 2920 hour period). Over a typical operating cycle, this represents a conditional probability 

of an open TIP tube of approximately 3.4E-4. 

Indexing

Units

Ball

Valve

Shear

Valve

Probe

Shield

Chamber

Drive

Mechanism

from Drive 

Control unit

to Flux Probing

monitor

Drywell wall

Shield wall

N2 Purge 

Equipment

RVs (2)

to RPV

A.C. D.C.

A.C. D.C.

A.C. D.C.

DRYWELL REACTOR BUILDING
4-14

MELCOR Model of TIP Guide Tubes

The baseline MELCOR calculations of accident progression and source terms did not consider 

leakage of steam, hydrogen or fission products from the RPV to the reactor building through the 

TIP guide tubes. The conditional probability of the TIP system being in operation at the time of 

a severe accident initiating event was well below the truncation limit for accident sequence 

selection. 

However, a sensitivity calculation was performed to evaluate the effects of open (unisolated) TIP 

guide tubes on accident progression and the radionuclide source term for the LTSBO accident 

sequence. The MELCOR model for leakage through the TIP guide tube accounts for internal 

geometry of the tubes and flow resistance due to wall friction along the length of the tubes 

(approx. 150 ft). As shown in Figure 3, a single control volume is used to represent the total 

internal volume of the guide tube(s) and flow paths connect each end of the tube coolant space 

within the core and the reactor building. Heat transfer between flowing gas and the guide tube(s) 

is accounted for; and the tube internal wall represents an aerosol deposition surface. Results of 

the sensitivity calculation using this additional feature to the Peach Bottom MELCOR model are 

described in Section 5.7.3.

4.3 Primary Containment and Reactor Building

The primary containment of the BWR Mark I design consists of two separate regions: a drywell

and a wetwell. As shown in Figure 4-9 each region is explicitly represented in the MELCOR 

model with distinct hydrodynamic control volumes, flow paths, and heat structures to preserve 

the geometric configuration and major functional features of the Mark I design (e.g., steam 

pressure suppression, fission product scrubbing, and surface deposition). The drywell is further 

divided into four connected volumes to account for non-uniformities in the temperature and 

composition of the atmosphere during late phases of a severe accident.
4-15

Figure 4-9 Hydrodynamic nodalization of the primary containment

The internal volume, airflow flow pathways, and structures of the reactor building are modeled 

in considerable detail as illustrated in Figure 4-10 and Figure 4-11. The reactor building fully 

encloses the primary containment and participates in the release pathway of fission products 

from the containment to the environment by offering a large volume within which an airborne 

radionuclide concentration can be diluted by expansion into and mixing with the building 

atmosphere. 

CV200

CV220 FL012

FL020

CV220

FL021

FL903

(DW head 

flange failure)

CV210

CV210

CV205

FL014

Personnel

access

FL016

CRD hyd. 

pipe chase

FL015

CRD removal

FL904

(DW liner 

melt-through)

FL017

(DW nominal leakage)

FL022

(RB-WW vacuum breaker 

to NE torus corner room)

FL023

(RB-WW vacuum breaker 

to SE torus corner room)

CV201

CV202

FL202

FL200

FL910

(Wetwell Hard-Pipe vent

to atmosphere)

Drywell (DW)

Wetwell (WW)
4-16

Figure 4-10 Hydrodynamic nodalization of the reactor building (a)

CV412

(Refueling Bay)

CV411

CV409

(195' SE 

Quadrant) CV407

(195' NE

Quadrant)

CV405

(165' SE 

Quadrant)

CV404

(165' North Half)

CV403

(135' South Half)

CV402

(135' North Half)

FL407 

(open hatch)

FL423

FL902

(DW liner shear)

FL422 

(open hatch)

CV401 (Torus Rm)

FL425 (blowout panels)

CV903

(Environment)

CV401

FL414 

(open hatch)

FL017

(DW nom leakage)

FL402 (open grating) FL401 (open floor grating)

FL404 (E)

FL408 (E)

FL409 (W)

CV410

(195' SW

Quadrant)

FL416 FL417

Dryer - Separator Storage Pit 

divides the SE/SW Quadrants

CV408

(195' NW

Quadrant)

FL415

Spent Fuel Pool Volume 

included in CV412

which divides

NE/NW Quadrants of 195'

FL403 (W)

Section A-A

CV570

(Equip Access 

Airlock)

FL446 FL445

N

A

A

B

B

CV901

(Environment)

FL424 (nom. leakage)

FL421 (roof failure)

FL903

CV406

(165' SW 

Quadrant)

FL410

FL904 

Drywell Liner 

MeltThrough
4-17

Figure 4-11 Hydrodynamic nodalization of the reactor building (b)

CV412

(Refueling Bay)

CV411

CV407

(195' NE Quadrant)

CV406

(165' SW 

Quadrant)

CV404

(165' North Half)

CV403

(135' South Half)

CV402

(135' North Half)

FL423

CV401 CV401

CV410

(195' SW Quadrant)

Section B-B

FL455

CV453

FL480 FL481 FL482

FL483

FL484

FL485

FL486

FL487

FL462 FL46

1

FL465

FL466

FL463

FL460

FL470

FL471

FL474

FL476 FL475

FL425 (blowout panels)

Elevator Shaft (not modeled)

CV491

CV455 CV490 CV492 CV493

CV494 CV495

B

B

CV481 CV482 CV483 CV484
4-18

The airborne concentration of fission product aerosols within the reactor building is attenuated 

by gravitational settling and other natural deposition mechanisms. The building is also equipped 

with a ventilation system with aerosol and charcoal filters, which would greatly aid in reducing 

an airborne radioactive release. However, these systems would not be available during the 

particular accident scenarios examined in this work, because of a loss of electrical power or other 

equipment failures. Therefore, the reactor building is occasionally referred to as a secondary 

containment, although it has a negligible capacity for internal pressure. 

4.4 Mechanisms for Induced RPV Depressurization 

Mechanisms for induced depressurization of the RPV in BWR severe accident sequences were 

first introduced to the U.S. NRC MELCOR models in the calculations performed for the reactor 

security assessment in 2002-2003. The principal motivation at the time was to correct previous 

calculations of high pressure accident sequences (e.g., SBO), which allowed SRVs to continue 

cycling for several hours after the onset of core damage, when gas discharge temperatures 

exceeded values at which material damage to moving valve components would challenge normal 

valve behavior. Specific values at which this could occur are discussed later. Data were not 

available to support a deterministic valve degradation model. As a result, an intuitive approach 

was used to capture the basic idea that moving valve components would eventually break or 

seize when sufficient heat was transferred from the flow gas stream to the valve body. 

Eventually other mechanisms for RPV depressurization were added to the MELCOR model. 

Section 4.4.1 summarizes the historical development of this aspect of BWR severe accident 

modeling practices. The specific modeling approach used in the current SOARCA calculations 

is then described in Section 4.4.2.

4.4.1 Historical Development of BWR RPV Depressurization Mechanisms

Within this section a summary of SRV seizures at high temperatures is discussed in 

Section 4.4.1.1. Section 4.4.1.2 provides a discussion on the stochastic failure for a cycling 

SRV, and Section 4.4.1.3 discusses the potential for creep rupture of a BWR main steam line

pipe or RPV nozzle.

4.4.1.1 Valve Seizure at High Temperatures

An initial criterion for high-temperature valve failure was based on manufacturers’ information 

describing the strength of stainless steel, published by the Stainless Steel Information Center.

10 

Softening or loss of strength of stainless steel (300 series) was described as “about 1000 oF”

(811 K). The same reference also suggested the maximum service temperature for intermittent 

exposure of stainless steel components is 1600 oF (1100 K). It was assumed that ‘service 

temperature’ referred to the temperature of thermal environment within which steel components 

operate. In the case of a valve, this was assumed to be the internal gas temperature. Therefore, 

in earlier analyses, a cycling valve was judged to cease functioning properly when the internal 

gas temperature exceeded a value between the two referenced values (i.e., somewhere between 

811 K and 1100 K). In particular, valve seizure (failure to reclose) was assumed to occur if the 

valve was exposed to discharge gas temperatures greater than ~1000 K for several cycles. 

 

10 Reference: www.ssina.com/composition/temperature.html
4-19

The response of valve components to high temperature gas strongly depends on valve design and 

operation. For example, different thermal failure criteria were developed for the Peach Bottom 

and Grand Gulf MELCOR models because the two plants have different types of SRVs. 

Three-stage Target Rock SRVs are installed at Peach Bottom, and Dikkers SRVs are installed at 

Grand Gulf. Other valve designs in the BWR fleet include the two-stage Target Rock, which is 

modified version of the earlier three-stage valve, and another SRV is manufactured by Crosby 

(now a subsidiary of Tyco). Figure 4-12 depicts the cross-section of a typical two-stage Target 

Rock valve; Figure 4-13 shows the cross-section of a Crosby valve, which is similar in design to 

the Dikkers valves. 

Valve design and operating features are important for several reasons. First, the way in which 

the valve opens to relieve pressure differs between the Target Rock and Dikkers/Crosby designs. 

Target Rock SRVs are pilot-operated valves, which lift to a full-open position when pressure 

within the SRV exceeds a setpoint. When pressure decreases below another setpoint, the valve 

fully re-seats. Movement of the main valve disc is controlled by a pilot valve, which is distinct 

from, but integral to, the main SRV valve body. Movement of the pilot valve re-aligns gas flow 

through small ports and vent lines within the valve body (see Figure 4-12), allowing RPV 

pressure to help keep the valve fully-seated when pressure is within desired values, and to help 

lift the valve if pressure gets too high. In contrast, the Dikkers/Crosby SRV design is a springloaded valve that ‘pops’ open to relief RPV pressure, and then gradually recloses as internal 

pressure decreases. The variable valve stem position (or valve open fraction) allows RPV 

pressure to be maintained close to a target value until RPV pressure reduces below a minimum 

setpoint when the valve recloses.
4-20

Closed Position

Open Position

Figure 4-12 Two-stage Target Rock pilot-operated safety relief valve

(Taken from Target Rock Corporation drawing M-1-R-213) 10 in

6 in

Pilot Valve

Assembly

10 in

6 in

Pilot Valve

Assembly
4-21

Figure 4-13 Crosby safety relief valve

These differences in valve design and operation led to different criteria for valve seizure at high 

temperatures in the Peach Bottom and Grand Gulf MELCOR models. In both cases, seizure was 

assumed to occur when sufficient heat was transferred from discharged gas to the valve body for 

internal valve components to cease functioning. In the absence of data on valve performance at 

high temperatures, or detailed calculations of valve heat up, the following simple assumptions 

were made:

1. The number of times a Target Rock valve could cycle at temperatures at or above 1000 K 

wasn’t known. A value of 10 cycles was chosen to represent the expectation that several 

cycles would be necessary to transfer enough heat to valve internal components to 

deform or expand valve components, causing failure. It was recognized that although 

convective heat transfer from the gas would only occur when the valve was open, heat 

transfer within the valve body would continue after the valve stem reseated. Therefore, it 

was postulated that non-uniform thermal expansion could reduce clearances of valve 

components and cause the valve seize in the closed position; or material softening and 

deformation could cause the valve to fail in the open position. Uncertainties in both 

failure mechanisms (i.e., thermal expansion and material softening/deformation) were 

also reflected in the number of cycles used to characterize failure to reclose. That is, if 

the lowest set point SRV seized closed after 2 or 3 cycles with gas temperatures above 

1000 K, the next lowest set point SRV would pick up the load and begin cycling without 

significant pre-heating. A nominal valve failure (seize open) criterion was defined as: 

10 cycles with gas temperatures, prior to opening, above 1000 K. This was judged to be 
4-22

a reasonable approximation of conditions under which one of the 11 Target Rock valves 

at Peach Bottom would seize in the open position.

2. Tracking the number of valve cycles was judged to not be a meaningful method for 

characterizing heat up and seizure of a Dikkers/Crosby valve. Instead, seizure was 

assumed to occur if the valve discharged high temperature gas for a sufficiently long 

period of time. A failure criterion was developed based on the concept of a cumulative 

damage function, which tracked the amount of time high temperature gas was discharged 

through the valve and compared it to a time limit. The time limit was assumed to be 

inversely proportional to temperature, and the valve was assumed to seize in the position 

it held at the time the failure limit was reached. This was often a few percent of full 

open. The specific values used in this time-at-temperature criterion were:

a. 60 minutes at temperatures of 1000 K.

b. 30 minutes above 1500 K.

These criteria were used in the calculations of severe accident progression for Peach Bottom and 

Grand Gulf in the 2002 security assessment as well as the 2005 analyses of fission product 

source terms for high burnup fuels for both plants. The valve failure criterion which was used 

for Grand Gulf has not been benefitted from the additional work done for Peach Bottom, and 

may be examined in the future. The thermal criterion for the Target Rock SRVs was modified 

after a more detailed assessment as described in Section 4.4.2.2.

4.4.1.2 Stochastic Failure

The possibility that a cycling SRV would randomly fail to open, or to reclose after opening, was 

not represented in the security assessment calculations, nor in the high burnup fuel source term 

analyses. However, several hundred cycles were calculated for some accident sequences, such as 

station blackout, which raised concerns that random failures should not be ignored. The Peach 

Bottom Individual Plant Examination (IPE) cites a plant-specific failure rate (to reclose) for an 

SRV of 3.7x10-3

/demand .

11

It must be noted that the NRC assessment of component reliability in nuclear systems suggests a 

failure lower than the value cited in the Peach Bottom IPE. An NRC analysis of industry

average data for SRV performance in NUREG/CR-6928 [18], for example, estimated a mean 

value of approx. 8x10-4/demand with 5th and 95th percentile values of 4.3x10-5 and 

3.1x10-3/demand, respectively.

12 Therefore, the Peach Bottom value is slightly larger than the 

95th percentile of the industry-average value.

13

 

11 Reference: Philadelphia Electric Co., Individual Plant Examination, Peach Bottom Units 2 and 3, Vol., 2, §4.6.5.3.1. 12 The 5th and 95th percentiles are not reported in NUREG/CR-6928, but can be readily calculated from the Beta 

distribution parameters listed in Table 5-1. 13 Significant reductions in the observed rate of spurious valve opening were achieved by various modifications to the 

Target Rock SRV through a BWR Owners Group initiative coordinated under General Safety Issue (GSI) B-55. The 

extent to which these modifications would also affect the expected rates for failure to reclose is not known. However, 

this might contribute to the difference in the current (NRC) estimate of SRV failure rate (to reclose) and the older (circa 

1990) failure rate reflected in the Peach Bottom IPE.
4-23

Control logic was added to the Peach Bottom MELCOR model to calculate the cumulative 

probability of random failure of a cycling SRV to reclose based on the IPE failure rate. Failure 

was assumed to occur when the cumulative probability (i.e., confidence of failure) reached 90%. 

The high confidence level was used in recognition of the principal in PRA system reliability 

analysis that credit should not be taken for ‘benevolent’ failures of components. That is, failure 

of a reactor component should not be assumed if it has a potentially beneficial effect on the 

accident sequence.

14 Therefore, rather than use an ‘expected value’ condition for predicting 

failure, valve failure was assumed to occur only if the number of cycles was sufficiently large 

that the failure probability was high (0.9).

4.4.1.3 Main Steam Line Creep Rupture

The potential for creep rupture of a BWR main stream line (i.e., piping or RPV nozzle) was not 

evaluated in the MELCOR calculations performed for the security assessment, nor was it 

represented in the high burnup source term calculations. This feature was added to the Peach 

Bottom model developed for SOARCA. Initially the model was implemented for the sole 

purpose of monitoring the potential for conditions that would suggest the possibility of creep 

rupture. Flow paths to represent the rupture of main steam line piping were added relatively late 

in the SOARCA model development process, in response to comments received from the 

SOARCA peer review panel.

The creep rupture model is a direct translation of the hot leg creep rupture model used in the 

Surry MELCOR calculations. The Larson-Miller control function available in MELCOR is 

applied to a heat structure representing the RPV nozzle to main steam Line A and (separately) to 

the heat structure representing the horizontal section of the main steam line piping immediately 

adjacent to the RPV nozzle (approx. 3.5 m in length). The nozzle is assumed to have twice the 

thickness of the main steam line pipe. Creep rupture of either structure is assumed to result in a 

an opening to the drywell atmosphere equivalent to the full open area of the main steam line.

4.4.2 Modeling Approach in SOARCA

Within this section a summary of SRV stochastic failure to reclose is discussed in 

Section 4.4.2.1. Section 4.4.2.2 provides a discussion on the thermal failure mechanism for a 

cycling SRV, and Section 4.4.2.3 discusses the differential thermal expansion for materials in a 

SRV. Section 4.4.2.4, Section 4.4.2.5, and Section 4.4.2.6 discuss material deformation, SRV 

pilot valve failure, and SRV spring softening respectively.

4.4.2.1 Stochastic Failure to Reclose

The SOARCA Peer Review Panel agreed with the basic method used to represent stochastic 

failure of an SRV in the MELCOR calculations. One panel member noted, however, that a 

recent EPRI study of SRV performance, which is undocumented at the time of the review,

 

14 In Level 1 PRA, this principal primarily applies to components that would enable the operation of alternate or backup 

safety systems. An example would be failure of a cycling SRV to reclose would depressurize the RPV and enable the

use of low-pressure coolant injection systems for makeup. Refer to requirement SY-A12 in ASME/ANS RA-Sa-2009, 

“Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant 

Applications.”
4-24

suggests the failure rate cited in the plant-specific IPE and the value in the generic data base are 

probably too low, especially when valve temperatures begin to rise above the design value. This 

view was acknowledged as an opinion, and is not supported by failure data that would supplant 

the information used to calculate the IPE or generic failure rates. Therefore, the failure rate 

obtained from the plant-specific IPE is retained in the SOARCA calculations. Deterioration in 

valve performance at temperatures above design is represented by a separate, and independent, 

failure model described in the next section.

It should be noted, however, that the failure rate reflected in the generic data base and the value 

obtained from the Peach Bottom IPE are conceptually different from the situation modeled here. 

In simple terms, the rate at which an SRV fails to reclose is calculated by dividing the number of 

observed valve failures (to reclose) by the number of valve demands. This ratio, therefore, 

reflects the conditional probability that a valve would fail to reclose, given a successful demand 

to open. However, the failure events that represent the numerator of this ratio occurred after only

a few valve cycles. The precise number is difficult to determine from the raw data documented 

in NUREC/CR-6928. However, it is clear that valve failure data after numerous cycles are

extremely rare (perhaps non-existent) primarily because events involving numerous, continuous 

valve cycling are not observed. It is, therefore, debatable whether the failure rate used to 

calculate the (low) probability of failure to reclose after a few cycles should be extrapolated to 

estimate the (higher) probability of failure after a large number of cycles. Other unknown failure 

mechanisms would likely overwhelm those that lie behind the nominal failure rate. This 

qualitative observation is consistent with the opinion expressed by members of the peer review 

panel that the valve failure rates obtained from the PRA data base are too low. Also, it should be 

noted that comments provided by the licensee on this analysis indicated the early failure rate 

reported in their IPE was not based on plant-specific performance data, and they have since 

replaced this value with the industry value reported in NUREG/CR-6928.

Accident progression modeling in SOARCA reflects best estimate values for uncertain 

parameters. Therefore, the conditions at which random valve failure would occur reflect the 

expected value, or mean, number of cycles at which the valve fails to reclose. This number is the 

inverse of the failure rate (1/λ), or 1/(3.7x10-3 per demand), which is 270 consecutive cycles. 

4.4.2.2 Thermal Failure Mechanisms

Several members of the SOARCA Peer Review Panel expressed the opinion that the criteria used 

to represent conditions at which a cycling SRV would seize in the open position due to thermal 

effects (described earlier) were too severe. They agreed that seizure due to physical deformation 

of moving valve components would be expected before the temperature of valve components 

reached 1000 K. They also expressed the opinion that multiple valve cycles above 1000 K was 

not credible. 

Four thermal failure mechanisms were examined to refine the criteria for valve seizure. Each is 

described below. These mechanisms are assumed to operate independently of the mechanisms 

causing stochastic failure. 
4-25

4.4.2.3 Differential Thermal Expansion

Heat transfer from hot gases discharged through an open SRV to internal surfaces of the valve 

body would gradually increase the temperature of valve components. The internal geometry of 

the valve is complex and the temperature distribution within the valve would not be uniform, 

causing differential thermal expansion of valve components. If displacements are sufficiently 

large, the clearances required for moving components to operate could close, resulting in valve 

seizure.

The extent to which differential expansion might reduce valve stem clearances is examined very 

simplistically. First, the differential expansion of the valve stem from its surrounding sleeve and 

valve body, due solely to differences in material properties was considered. That is, the stainless 

steel stem will expand at a slightly faster rate than its carbon steel enclosure due to differences in 

the linear coefficient of expansion for stainless verses carbon steel. Second, the effects of 

temperature gradient within the valve body were considered.

Differences in material properties: The mean linear expansion coefficient for the valve body 

(ASTM A216 Gr WCB steel)

15 is 8.3 x 10-6 in/in oF; the coefficient for the valve stem and disc 

(A276 Type 304 stainless steel)

15 is 11 x10-6 in/in oF. Therefore, if the temperature of the entire 

valve increased to 1000 oF (425 oF greater than the design value), outward radial displacement of 

the valve stem (approx. 2.5-in diameter) would be greater than the surrounding valve body by 

approximately 1.4 mil:

dr = (11x10-6 – 8.3x10-6) in/in- oF * 2.5/2 in * 425oF = 1.4x10-3 inches

The valve stem is a solid cylinder, surrounded by a thermal sleeve, as illustrated in Figure 4-14. 

This is probably not sufficient to cause seizure of the valve stem. 

Effects of temperature gradient

In the absence of mechanistic, multi-dimensional calculations of the transient thermal response 

of the valve body to cyclic heating, it is difficult to judge whether temperature gradients within 

the valve would be sufficiently large for adjacent valve components to bind against each other 

due to differential thermal expansion. However, the rates at which major components within the 

valve would heat up can be estimated using some simple, but reasonable geometric models. In 

particular, the response times of major valve components were estimated by representing them as 

a geometric shape that can be modeled using a one-dimensional heat transfer model. 

The main portion of the valve body, which encloses the gas flow stream, was approximated as a 

cylinder with an internal diameter of 6 inches (i.e., the approximate valve throat diameter) and 

length to diameter ratio (L/D) of 5. The outer diameter, or wall thickness, of the cylinder was 

treated parametrically because, as indicated in Figure 4-12, the shape of the valve body is not a 

perfect cylinder. Two alternatives were considered. The first alternative examined a 1-in wall 

thickness, which represents the approximate wall thickness of the entrance and exit portion of the 

valve body. The second alternative attempts to capture the effects of the large steel mass 

 

15 Reference: Peach Bottom UFSAR, 4.4-4.
4-26

appended to one side of the main valve body, and houses the pilot valve operator. The wall 

thickness in this case was 3 in, which is the equivalent (uniform) thickness of a cylinder with 

one-half the weight of a fully-assembled SRV.

16 Calculating the thermal response of a cylinder 

with this range of material masses (or wall thicknesses) to cyclic heating by gas flowing within it 

provides an indication of the rate at which the main valve body would respond to heat transfer 

from gas flow when the SRV opens.

A specific internal component of interest is the main valve stem that must slide along a 

cylindrical sleeve that is integral to the valve body to seat and unseat the disc. Also, the upper 

portion of the valve stem is mounted to a ring piston that slides within a hollow cavity within the 

main valve body. These components are highlighted in Figure 4-14. If the valve stem and 

attached ring piston were to expand at a faster rate than the surrounding valve body, mechanical 

binding could occur, and the valve would stick or seize in the open position. Seizure in the open 

position is judged more likely than the closed position because gas flow through the valve 

(i.e., the heat source) only operates when the valve is open. Since the disc and a portion of the 

valve stem are directly immersed in the high temperature gas flow stream, one would expect 

these components to heat faster than the surrounding valve body. The thermal response of the 

valve stem was therefore estimated by modeling it as a solid 3-in diameter cylinder with an 

L/D=3. This diameter represents the combined dimension of the valve stem (~2.5-in.) plus the 

exposed portion of the surrounding valve stem sleeve. 

Figure 4-14 Valve stem assembly within valve body

(Taken from Target Rock Corporation drawing M-1-R-213)

Boundary conditions for each of the 1-dimensional conduction models (i.e., the two valve body 

models and the valve stem model) were taken from a preliminary MELCOR calculation of the 

 

16 Total weight of a fully-assembled valve is given as 1500 lbs in an EPRI SRV Testing and Maintenance Guide 

(TR-105872s, Aug 1996). 

Approx. 3-in

6 in

Valve stem sleeve

Valve disc Valve stem

Ring piston
4-27

long-term station blackout accident sequence. The calculated temperature history, composition 

of gases passing through the SRV, and the time-dependent (cyclic) gas velocity through the 

valve were tabulated and applied as a boundary condition for the simple flow model illustrated in

Figure 4-15. 

Figure 4-15 Arrangement of 1-dimensional safety relief valve heat-up calculation

The boundary conditions are illustrated in Figure 4-16 through Figure 4-19 as follows:

• Figure 4-16: Gas flow through the valve is based on a cycle period of 45 seconds with an 

open cycle duration of 5 seconds.

• Figure 4-17: Gas velocity during an open cycle is constant during a single 5-second 

cycle, but increases from 420 m/s to 500 m/s as the gas temperature (Figure 14) increases 

from 600 K to 1100 K.

• Figure 4-18: Gas temperature as a function of time. Approximately 100 K temperature 

changes between the beginning and end of a cycle were neglected. The mid-point values 

were used as shown in the table adjacent to the figure.

• Figure 4-19: The Steam and hydrogen mole fractions were specified as a function time 

as indicated in the figure and adjacent table.

System pressure was assumed to be constant at 1150 psia.

CV100 CV001 CV200

Valve gas velocity defined 

as func(time):

* Periodic cycles 45 sec 

apart, 5 sec duration

Cylindrical HS with 

variable wall thickness

CV100 CV200

Valve inlet (gas) boundary 

conditions as func(time):

* Temp

* Composition (steam/H2)

Valve Body Model

Valve Stem Model

CV001

Solid cylinder HS
4-28

Figure 4-16 Number of safety relief valve cycles from preliminary LTSBO calculation

Figure 4-17 Gas velocity through safety relief valve from preliminary LTSBO calculation

420

430

440

450

460

470

480

490

500

660 670 680 690 700 710 720

time [min]

Nr. Of Cycles

Average: Cycle every 45 seconds

 Each cycle approx 4-6 sec duration

0

100

200

300

400

500

600

660 670 680 690 700 710 720

time [min]

Velocity thru SRV [m/s]

Cycles with Gas Temperatures over ...

1000K

900K 800K 700K

0

100

200

300

400

500

600

0 10 20 30 40 50 60 70 80 90 100

time [sec]

Velocity thru SRV [m/s]

Detail beginning at 680 min
4-29

Time 

(sec)

Ave 

Temp 

(K)

36000. 600.

38400. 625.

39600. 650.

39900. 700.

40080. 800.

40800. 940.

42000. 1100.

Figure 4-18 Gas temperature through safety relief valve from preliminary LTSBO

calculation

Time (sec) Steam 

v/o

36000. 1.0

37200. 0.98

37800. 0.94

38400. 0.85

39600. 0.72

39900. 0.67

40500. 0.77

40800. 0.74

41640. 0.83

41760. 0.75

42000. 0.77

Figure 4-19 Gas composition through safety relief valve from preliminary LTSBO

calculation

400

500

600

700

800

900

1000

1100

1200

1300

1400

600 620 640 660 680 700 720

time [min]

Main Steam Line Gas Temp [K]

SRV seizes open 

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

1

600 620 640 660 680 700 720

time [min]

Mole Fraction

Steam

Hydrogen

SRV Heat-up BC
4-30

The calculated thermal response of the 1-dimensional conduction models are compared in 

Figure 4-20. As expected, the relatively small thermal mass of the 3-in diameter shaft responds 

more quickly to increasing gas temperature than the heavier valve body. The centerline 

temperature of the 3-in diameter cylinder, representing the valve stem, lags the gas temperature 

by approximately 4 minutes. In comparison, the 1-in thick cylinder lags by approximately 

6 minutes and the 750 lb cylinder, each represent the valve body, by almost 20 minutes, when the 

internal gas temperature reaches 850 K. 

From the perspective of thermal expansion, the principal result of these calculations is the 

temperature difference between the valve body and the valve stem when gas temperatures exceed 

800 K. This difference depends on the geometry and mass of steel attributed to the valve body. 

However, based on the two simple options examined here (i.e., a 1-in versus 3-in wall thickness), 

the valve body temperature is 20 to 140 K (11 to 78 oF) less than the valve stem temperature. A 

bounding estimate of the resulting differential expansion of the valve stem within the cooler 

valve body is obtained by assuming the maximum of this range. The resulting differential radial 

expansion is 1.1 mil:

11 x 10-6 in/in-oF * 2.5/2 in * 78 oF = 1.1x10-3 inches

If this value is added to the difference associated with material properties (estimated earlier), the 

combined radial expansion is less than 3 mil. This is within the range of possible values for the 

gap between sliding metal components, but is not sufficiently large to confidently conclude,

without more detailed analysis, that thermal expansion alone would cause valve seizure with gas 

temperatures up to approximately 1000 K. 
4-31

Figure 4-20 Thermal response of safety relief valve using simple 1-dimensional models

4.4.2.4 Material Deformation

Mechanical properties of stainless steel at very high temperatures described by a trade 

association for the specialty steam industry17 were used to estimate the temperature at which 

SRV failure would occur in early (preliminary) MELCOR calculations for the SOARCA 

Program. However, two aspects of this evaluation merit re-consideration. First, the ‘failure 

temperature’ was applied to the calculated temperature of gases flowing through the valve. As 

indicated in Figure 4-20, the temperature of at least some valve components lags the gas 

temperature by a relatively small amount (e.g., perhaps as little as 20 K). Therefore, the ‘error’ 

introduced by this simplification is not significant in comparison to the temperature at which 

failure is likely to occur. 

The second aspect is more important; namely, the temperature at which material deformation 

occurs is lower than the value used in the early calculations (i.e., 1000 K). As noted earlier, this 

value was selected primarily to reflect the ‘service temperature’ for stainless steel components, 

as reported by the steel industry trade association. However, a review of vendor literature on 

material properties of 304 stainless steel clearly indicates the maximum service temperature of 

approximately 1600 o

F (~1100 K) is based on the scaling properties (or resistance to corrosion) 

 

17 Reference: High Temperature Properties of Stainless Steel, published by the Specialty Steel Industry of 

North America. Available at: www.ssina.com/composition/temperature.html

600

650

700

750

800

850

900

950

1000

1050

1100

600 620 640 660 680 700

time [min]

Temperature [K]

Gas Temp

Valve Stem: Centerline of 3-in diam SS cylinder

Valve Body: Outer surface of 1-in thick CS cylinder

Valve Body: Mid-point of 750 lb CS cylinder

~140K

~20K
4-32

of 300 series stainless steel, rather than its mechanical properties. Spec sheets for 304/304L 

stainless steel from several vendors suggest a more appropriate temperature at which mechanical 

strength begins to deteriorate is 1000 o

F (811 K). 

Table 4-2 compares the tensile and yield strength of 304 stainless steel at elevated temperature to 

values at the design temperature of the reactor coolant system (~600 oF). The reduction in 

ultimate tensile strength doubles for every 200 oF increase in temperature. This provides a 

quantitative measure of the qualitative statement given in a steel industry Spec Sheet, which 

states that stainless steel experiences a loss of strength at temperatures above 1000 oF.

Engineering judgment is needed to apply the information in Table 4-2 for the purposes of 

estimating when moving valve components would cease to function due to material deformation. 

If one assumes valve components continue to function up to a temperature 1000 oF, where 

material softening begins, the valve stem and other internal components would have already 

endured a reduction in tensile strength of approximately 18%, and a reduction in yield strength of 

28%. If material temperature increases another 200 oF, however, the ultimate tensile strength 

reduces by another 22% (below its already reduced value at 1000 oF). At a temperature above 

1000 oF but below 1200 oF (i.e., approximately 900 K), the ultimate tensile strength and yield 

strength are 30% lower than their values at the SRV design temperature. It is unreasonable to 

expect moving valve components to continue functioning with this deterioration in mechanical 

properties. A maximum material temperature of 900 K is, therefore, recommended as a 

reasonable limit to valve operation.

Table 4-2 Percent change in strength of stainless steel from value at 600oF (589K) 18

Temperature

oF (K)

Change in ultimate 

tensile strength relative 

to value at 600oF

Change in 0.2% yield 

strength relative to 

value at 600oF

600 (589) 0.0 0.0

800 (700) -6.0% -14.9%

1000 (811) -17.7% -27.6%

1200 (922) -35.4% -34.3%

1400 (1033) -57.4% -43.3%

4.4.2.5 Pilot Valve Failure

A third possible failure mechanism results from heating of the pilot valve, which actuates the 

main valve. Small low-flow ports within the valve body direct flow from gas upstream of the 

main valve disc to the pilot valve, which is comprised of a machined bellows. According to the 

Target Rock manual, the machined bellows, “acts as a combination piston, spring and hermetic 

seal.” Expansion or extension of the bellows due to high internal pressure and movement of the 

pilot valve stem is the normal means by which the pilot valve realigns itself, and causes the main 

valve to open. It’s conceivable that thermal expansion of the bellows might cause the same 

 

18 AK Steel Corporation, West Chester, OH, “Product Data Bulletin: 304/304L Stainless Steel,” 

394/394L-B-08-01-07 (2007).
4-33

effect. However, the flow rate to the pilot valve is not known and, therefore, the heat up rate 

cannot be estimated. 

4.4.2.6 Spring Softening

It is also conceivable that flow of gases through the ports into the pilot assembly could heat the 

main valve spring or the smaller spring encapsulated in the pilot valve assembly. This might 

result in softening of the spring and a loss of valve stem closing force. The rate at which heat 

could be carried into this portion of the valve is not known for the same reasons described above 

for pilot valve failure. 

4.5 Behavior of Ex-Vessel Drywell Floor Debris 

The drywell floor is subdivided into three regions for the purposes of modeling 

molten-core/concrete interactions. The first region, which receives core debris exiting the 

reactor vessel, corresponds to the reactor pedestal floor and sump areas (CAV 0). Debris that 

accumulates in CAV 0 can flow out through a doorway in the pedestal wall19 to a second region 

representing a 90 degree sector of the drywell floor (CAV 1). If debris accumulates in this 

region to a sufficient depth, it can spread further around the annular drywell floor into the third 

region (CAV 2). This discrete representation of debris spreading is illustrated in 

Figure 4-21.

Two features of debris relocation within the three regions are modeled. The first represents bulk 

debris spill over or movement from one region to another. A control system monitors the debris 

elevation and temperature within each region, both of which must satisfy user-defined threshold 

values for debris to move from one region to its neighbor. More specifically, when debris in a 

cavity is at or above the liquidus temperature of concrete, all material that exceeds a predefined 

elevation above the floor/debris surface in the adjoining cavity is relocated (i.e., 6 inches for 

CAV 0 to CAV 1 and 4 inches for CAV 1 to CAV 2). When debris in a cavity is at or below the 

solidus temperature of concrete, no flow is permitted. Between these two debris temperatures, 

restricted debris flow is permitted by increasing the required elevation difference in debris 

between the two cavities (i.e., more debris head required to flow).

The second control system manages the debris spreading radius across the drywell floor within 

CAVs 1 and 2. Debris entering CAV 1 and CAV 2 is not immediately permitted to cover the 

entire surface area of the cavity floor. The maximum allowable debris spreading radius is 

defined as a function of time. If the debris temperature is at or above the concrete’s liquidus 

temperature, then the maximum transit velocity of the debris front to the cavity wall is calculated 

 

19 Although the drawing provided by the licensee seems to indicate the presence of a swing-door in the 

personnel opening at the base of the reactor pedestal, the analysis described here assumes this door does not 

actually exist. Years of research on the issue of drywell liner melt-through never acknowledged the 

presence of a door (e.g., NUREG/CR-5423 and NUREG/CR-6025.) It is noted in the introduction to 

NUREG/CR-5423 that the geometry of the Peach Bottom configuration was used as the template for the 

analysis. The flow of debris from the pedestal onto the outer drywell floor would not be impeded in any 

way by an obstacle in the concrete ‘doorway’ in the pedestal wall. As a result, the current SOARCA 

analysis applied the same rationale and assumed molten debris would freely flow from the pedestal onto the 

drywell floor. 
4-34

(i.e., results in 10 minutes to transverse CAV 1 and 30 minutes to transverse CAV 2). When the 

debris temperature is at or below the concrete solidus, the debris front is assumed to be frozen, 

and lateral movement is precluded (i.e., debris velocity is 0 meters per second). A linear 

interpolation is performed to determine the debris front velocity at temperatures between these 

two values.

Full mixing of all debris into a single mixed layer is assumed in each of these debris regions. 

The specific properties for concrete composition, ablation temperature, density, solidus 

temperature, and liquidus temperature are specified. The concrete composition represented in 

the MELCOR model is listed in Table 4-3. The drywell floor concrete composition includes 

13.5% rebar.

Figure 4-21 Drywell floor regions for modeling molten-core/concrete interactions

CAV 0

CAV 2

CAV 1

CAV

0

1

2

FLOOR

AREA

29.92

22.75

68.25

EQUIV

RADIUS

3.086

2.691

4.661

PERIMETER

RATIO

0.95

--

0.94

0.62

1.72

1.08

6.706 m

3.086 m

0.914 m
4-35

Other key user-defined concrete properties are selected to match defaults for limestone-common 

sand concrete including the following:

• Ablation temperature of 1,500 K

• Solidus temperature of 1,420 K

• Liquidus temperature of 1,670 K

• Density of 2,340 kilograms (kg) per cubic meter

• Emissivity of 0.6

Table 4-3 Concrete Composition

Species Mass Fraction

Al2O3 0.0091

Fe2O3 0.0063

CaO 0.3383

MgO 0.0044

CO2 0.2060

SiO2 0.3645

H2Oevap 0.0449

H2Ochem 0.0265

It must be noted that lateral debris mobility is strongly affected by geometric details of the 

drywell floor design, and these design features vary among BWRs with a Mark I containment. 

The volume of debris that can be sequestered in the sump (and, therefore, not available for lateral 

movement toward the drywell liner) spans a wide range. Further, the steel liner at the periphery 

of the drywell floor is protected or elevated above the surface of the floor by a concrete curb in 

some plants. The volume of the sump at Peach Bottom (located within the pedestal region) and 

the location at which the steel liner intersects the drywell floor were taken into account in the 

MELCOR model described above.

4.6 Containment Failure Model

The MELCOR model of the Peach Bottom Mark I containment incorporates criteria for opening 

leak pathways through the containment pressure boundary by two distinct mechanisms. The first 

mechanism, which occurs in all the calculations involving sufficient core damage and breach of 

the RPV lower head, is drywell liner melt-through. The second mechanism is leakage through 

the drywell head flange during when high internal temperatures and pressure develop within the 

containment. Each is briefly discussed in the following.

Drywell liner melt-through

If debris flows out of the reactor pedestal and spreads across the drywell floor, as described in 

Section 4.5, and contacts the outer wall of the drywell, the steel liner will fail, opening a release 

pathway to the lower reactor building. Heat transfer between the steel liner and molten core 

debris is not explicitly calculated in the MELCOR model, due to limitations of the CAV 

Package, which addresses ex-vessel model debris behavior. The model assumes an opening in 

the drywell liner occurs 15 minutes after debris first contacts the drywell wall. This time delay 
4-36

represents an average of estimates for failure time discussed in NUREG/CR-5423 [12] for 

situations in which the drywell floor is not covered with water.

Containment over-pressure

Peach Bottom has a Mark I containment that consists of a drywell and a toroidal-shaped wetwell, 

which is half full of water (i.e., the pressure suppression pool). The drywell has the shape of an 

inverted light bulb. The drywell head is removed during refueling to gain access to the reactor 

vessel. The drywell head flange is connected to the drywell shell with 68 bolts of 2 1⁄2 inch 

diameter (Figure 4-22). The flanged connection also has two 3⁄4 inch wide and 1⁄2 inch thick 

ethylene propylene diene methylene (EPDM) gaskets. The torque in the 2 1⁄2 inch diameter bolts 

range from 817 to 887 foot-pounds (ft-lb) [19][20]. An average bolt torque of 850 ft-lb was used 

in this study. 

The 68 drywell head flange bolts (see Figure 4-22) are pre-tensioned during reassembly of the 

head. This pre-tension also compresses the EPDM gaskets in the head flange. During an 

accident condition, the containment vessel may be pressurized internally. The internal pressure 

would counteract the pre-stress in the bolts. At a certain internal pressure, all of the pre-stressing 

force from the bolts would be eliminated, and the EPDM gaskets would be decompressed. 

Further increase in the internal pressure would result in leakage at the flanged connection.

The EPDM gasket manufacturers recommend a maximum squeeze (compression) of 30 percent 

for a static-seal joint. The gaskets recover about 15 percent of the total thickness after the 

compressive load is removed from the flange. However, the licensee engineers informed the 

SOARCA personnel that the gaskets for the drywell head flange are squeezed to 50 percent to 

have a metal to metal contact to ensure no leakage at a design pressure of 56 psig. In addition, 

the gaskets are exposed to constant temperature and radiation, which contribute to early 

degradation. For this reason, the gaskets are replaced during each reassembly of the reactor 

vessel head. Based on this information and actual observations, the Peach Bottom licensee 

engineers recommended a gasket recovery of 0.03 inch. 
4-37

Figure 4-22 Drywell head flange connection details

Leakage areas for different internal pressures are shown in Figure 4-23 based on a gasket 

recovery of 0.03 inches. The drywell head flange does not leak until the internal accident 

pressure is 0.660 megapascal (MPa) (i.e., P/PD = 1.35 or 82 psig). Thereafter, there is a gradual 

increase in the leakage area. 

At high temperatures (greater than 755 K, or greater than 900 o

F), upward and radial thermal 

growth of the drywell would lead to binding of small and large penetrations against the 

biological shield wall and failure. In addition, radial growth of the containment may also cause 

the seismic stabilizers to punch through the upper portion of the drywell at high temperatures

[15]. This observation is consistent with the results of previous studies that show that the 

drywell is likely to fail at the low pressure range of 0-65 psig [15]. Therefore, it can be 

concluded that the drywell is likely to fail under any appreciable pressure load at structure

temperatures of 900 o

F or greater. 

Finally, the containment can fail by drywell shell melt-through containment failure (see relevant 

discussions in Sections 4.5 and 4.8.2). 

Lb

Area = Ab
4-38

Figure 4-23 Drywell flange leakage model versus containment pressure

4.7 Radionuclide Inventories and Decay Heat

One important input to MELCOR is the initial concentration of radionuclide groups in the fuel 

and their associated decay heat. These values are important to the timing of initial core damage 

and the location and concentration of the initial radioactive source. The radionuclide groups in a 

nuclear reactor come from three primary sources: 

1. Fission products are the result of fissions in either fissile or fissionable material in the 

reactor core. 

2. Actinides are the product of neutron capture in the initial heavy metal isotopes in the fuel. 

3. Other radioisotopes are formed from the 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 

[14]. As such, their standard output does not provide the type of information needed for 

MELCOR [7]. It is important to note that changes to the TRITON sequence in SCALE were not 

needed for this analysis. The BLEND3 post- processing software extracts output from the 

TRITON sequence and combines it in a way that makes it useful for MELCOR [7].

A Global Nuclear Fuel 10x10 (GE-14C) fuel assembly was used as a typical fuel element for the 

Peach Bottom analysis. Information about assembly dimensions, enrichments, and operating 

characteristics was obtained from the licensee (with permission from the fuel vendor) and used 

for a realistic evaluation. Twenty-seven different TRITON runs were performed to model three 

different cycles of fuel at nine specific power histories. The specific power histories ranged from 

0

0.01

0.02

0.03

0.04

0.05

0.06

1 1.5 2 2.5 3 3.5

Leak Area (m2)

Pressure Ratio [P/Pdesign]
4-39

2 megawatt-days per metric ton uranium (MWD/MTU) to 45 MWD/MTU, which bounded 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 in the use of a second fuel cycle, the fuel was 

burned for its first cycle using 25.5 MWD/MTU and, allowed to decay for an assumed 30 day 

refueling outage. Then nine different TRITON calculations were performed with specific powers 

ranging from 2 to 45 MWD/MTU. The BLEND3 code was applied to each of the 50 core 

nodes20 in the MELCOR model using average specific powers derived from data for three 

consecutive operating cycles and appropriate nodal 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. Values used in the MELCOR calculations corresponded to those generated for 

equilibrium conditions, in the middle of an operating cycle. A summary of the total (core-wide) 

decay power generated by this process is listed in Table 4-4.

Table 4-4 Decay Power in Peach Bottom MELCOR Model

Time Decay Power (MW)

0.0 sec 221.36

1.0 sec 204.13

3.0 sec 185.90

7.0 sec 167.306

13.0 sec 152.86

27.0 sec 136.18

54.0 sec 120.87

1.8 min 105.85

3.7 min 92.61 

7.4 min 80.96

14.8 min 69.46 

29.8 min 57.51

60.0 min 46.19

2.0 hr 37.18

12.0 hr 22.69

24.0 hr 18.59 

48.0 hr 14.97 

4.8 Other Modeling Issues and Uncertainties

The SOARCA project is intended to provide a body of knowledge regarding the realistic 

outcomes of severe reactor accidents. To accomplish this objective, the SOARCA project used

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

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

the severe accident analysis community. 

 

20 The 50 core nodes are in five radial rings by ten axial levels
4-40

The MELCOR 1.8.6 computer code [7] 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 discussions with operators during site visits. The code models and user-specified 

modeling practices represent the current best practices.

Uncertainties remain in our understanding of the phenomena that govern severe accident 

progression and radionuclide transport. Consistent with the best-estimate approach in SOARCA, 

all phenomena were modeled using best-estimate characterization of uncertain phenomena and 

events. Important severe accident phenomena and the proposed approach to modeling them in 

the SOARCA calculations were presented to an external expert panel during a public meeting 

sponsored by the NRC on August 21 and 22, 2006 in Albuquerque, New Mexico. A summary of 

this approach is described in Section 4.8.1. These phenomena are singled out because they are 

important contributors to calculated results and have uncertainty.

Section 4.8.2 briefly describes the two other topics, steam explosions, and drywell shell 

melt-through on a wet drywell floor, that have been previously included in lists of highly 

uncertain phenomena. 

Finally, a systematic evaluation of phenomenological uncertainties for a particular sequence is a 

separate task and not discussed in this report. That task will evaluate the importance and impact 

of alternative settings or approaches for key uncertainties.

4.8.1 Base Case Approach on Important Phenomena

Sandia National Laboratories conducted a review of severe accident progression modeling for 

the SOARCA project at a public meeting in Albuquerque, NM on August 21-22, 2006 [8]. This 

review focused primarily on best modeling practices for the application of the severe nuclear 

reactor accident analysis code MELCOR for realistic evaluation of accident progression, source 

term, and offsite consequences. The scope of the meeting also included consideration of 

potential enhancements to the MELCOR code as well as consideration of the SOARCA project 

in general.

The review was conducted by five panelists21 with demonstrated expertise in the analysis of 

severe accidents at commercial nuclear power plants. The panelists were drawn from private 

industry, the U.S. Department of Energy national laboratory complex, and a company working 

on behalf of German-government ministries. The review was coordinated by Sandia National 

Laboratories and attended by NRC staff. A separate task in the SOARCA project is planned to 

address the importance of uncertainties in these modeling parameters.

 

21 The expert panel that was convened in 2006 to review the best-estimate modeling approach is a different 

panel from the peer review panel convened to review calculated results and program documentation. The 

objectives and results of the latter are described elsewhere in this report.
4-41

The following important uncertain modeling practices were presented to the expert panel. The 

expert panel provided written comments and suggestions, which were incorporated into the 

subsequent analyses. Base case approaches were identified for these uncertain and typically 

important parameters.

• Fuel degradation and relocation treatment: An additional model has been added to 

characterize the structural integrity of the fuel rods under highly degraded conditions. The 

new model acknowledges a thermal-mechanical weakening of the oxide shell as a function of 

time and temperature. As the local cladding oxide temperature increased from the Zircaloy 

melting temperature (i.e., represented as 2098 K in MELCOR) towards 2500 K, a thermal 

lifetime function accrues increasing damage from 10 hours to 1 hour until a local 

thermo-mechanical failure. Similar time-at-temperature failure criteria are applied to 

oxidized channel boxes, but at any point in time channel temperatures and degree of material 

oxidation differ from those associated with adjacent fuel assemblies. Therefore, the collapse 

of channel boxes into particulate debris typically occurs at a different time from the collapse 

of fuel assemblies in the same radial ring of the core.

• Lower plenum debris/coolant heat transfer: Following the fuel-debris slump into the lower 

plenum, there may be fuel-coolant interactions. The lower plenum heat transfer settings were 

updated to reflect the end-state thermal condition of the debris in the deep pool FARO tests 

(i.e., significant thermal interaction with the water). The resultant behavior results in debris 

fragmentation and cooling if there is a pool of water in the lower plenum. Debris 

temperatures can subsequently increase after the pool of water evaporates, which in turn

heats up the reactor vessel lower head.

• Core plate failure: The timing of core plate failure affects the relocation of the degraded core 

materials from the core region into the lower plenum. The local thermal-mechanical failure 

of the lower core plate, the flow mixer plate, and the lower support forging are calculated 

within MELCOR using the Roark engineering stress formulae. The yield stress is calculated 

based on the loading and local temperature.

• Fission product release, speciation, and volatility: First, a new ORNL-Booth fission product 

release model is used that was adjusted to match the measured responses from the 

VERCORS Test 4. VERCORS Test 4 is representative of modern, high burn-up fuel. The 

previous default model was not representative of the high burn-up release physics.

Second, the predominant speciation of cesium was changed based on detailed analysis of 

the deposition and transport of the volatile fission products in the Phebus facility tests. 

The analysis revealed molybdenum combined with cesium and formed cesium 

molybdate. Previously, the default predominant chemical form cesium was cesium 

hydroxide. As consistent with past studies, all the released iodine combines with the 

cesium. Applications of this information to the MELCOR models used in the SOARCA 

calculations are described in SAND2010-1633, “Synthesis of VERCORS and Phebus 

Data in Severe Accident Codes and Applications” [10].
4-42

Gaseous iodine remains an uncertain source term issue, especially with respect to longterm radioactive release mitigation issues after the comparatively much larger airborne 

aerosol radioactivity has settled from the atmosphere. The mechanistic modeling 

treatment for gaseous iodine behavior is a technology still under development with 

important international research programs underway to determine the dynamic behavior 

of iodine chemistry with respect to paints, wetted surfaces, buffered and unbuffered water 

pools undergoing radiolysis, and gas phase chemistry. The base case treatment under the

best practices recommendation are sufficient for the mean effects addressed in this report, 

and it is planned to investigate parameterization of the gaseous iodine fraction of total 

iodine releases in the context of the uncertainty quantification phase of this work.

• Vessel lower head failure and debris ejection: The base case approach of modeling the 

vessel lower head failure and debris ejection included some modifications in MELCOR. 

First, all particulate debris in the lower plenum is permitted to be in contact with water, if 

present. In previous versions of the code, a restrictive one dimensional counter-current 

flooding limitation (CCFL) criterion prevented water from penetrating a deep debris bed. 

This restriction has been removed, effectively assuming water steam rising upward from the 

debris bed does not totally preclude water from entering lower regions of the bed (e.g., via 

lateral flow from peripheral regions of the lower plenum). Second, the mechanical response 

of the vessel lower head is modeled using a one dimensional creep rupture model. A 

Larson-Miller failure criterion is calculated based on the one dimensional conduction and 

stress profile through the lower head. Failure of a lower head penetration prior to gross head 

failure was not explicitly modeled in the SOARCA calculations partly because MELCOR 

does not currently offer a mechanistic model for calculating penetration structural response 

and associated changes in local debris morphology. This was mentioned earlier in Section 

4.2.2. Prior analyses of BWR lower head penetration failure mechanisms [17] indicate the 

drain line is more susceptible to failure than other penetrations (e.g., CRD and in-core 

instrument penetrations), particularly for the depressurized conditions observed in the 

SOARCA accident progression analysis. The impact of local failures (rather than global 

creep rupture of the head) will be examined in the uncertainty quantification phase of this 

work. 

• Ex-vessel phenomena – Molten Corium-Concrete Interactions (MCCI ): An evaluation of 

typical MELCOR calculations of ex-vessel debris behavior when debris is submerged in a 

pool of water concluded the default treatment of heat transfer between debris and an 

overlying pool of water was not consistent with observations from the MACE tests. The 

default value for the debris-water interface heat transfer coefficient in MELCOR did not 

account for multi-dimensional effects of fissures, other surface non-uniformities, and side 

heat fluxes. An enhancement to the default value was used to more closely replicate heat 

transfer rates observed in the MACE tests.

• Hydrogen combustion: The default MELCOR hydrogen combustion model was used in 

control volumes representing the Peach Bottom reactor building. This model assumes 

ignition of a flammable mixture occurs at a volume-average hydrogen concentration of 10%, 

provided sufficient oxygen is present. Ignition at this concentration reflects the absence of a 

strong ignition source that might be present in accident sequences involving active AC 
4-43

power. Modeling options were also exercised to calculate horizontal and vertical 

propagation of combustion flames between neighboring control volumes, after a time delay 

that accounts for the time required for a flame front to span the width of the control volume. 

The time associated with flame propagation assumes a volume-centered distance between 

neighboring volumes; therefore, the propagation time varies with the size of the control 

volume.

4.8.2 Early Containment Failure Phenomena 

Two phenomenological issues not explicitly modeled in the best-estimate approach used in 

SOARCA are: (1) alpha-mode containment failure, and (2) drywell liner melt-through in the 

presence of water leading to containment failure. These phenomena, if they occurred, would

result in early failure of containment, and were included in the probabilistic studies documented 

in NUREG-1150 to quantify the risks from nuclear reactors.

The alpha-mode event is characterized by the supposition that an in-vessel steam 

explosion might be initiated during core meltdown by molten core material falling into the 

water-filled lower plenum of the reactor vessel. The concern was that the resulting steam 

explosion could impart sufficient energy to separate the upper vessel head from the vessel itself 

and form a missile with sufficient energy to penetrate the reactor containment. This would 

produce an early failure of the containment building at a time when the largest mass of fission 

products is released from the reactor fuel. In the following years, significant research was 

focused on characterizing and quantifying this hypothesized response in order to attempt to 

reduce the significant uncertainty. A group of leading experts ultimately concluded in a position 

paper published by the Nuclear Energy Agency’s 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, posing little or no significance to the overall 

risk from a nuclear power plant. Therefore, the complex processes leading to this failure mode 

were not explicitly modeled in the BWR MELCOR calculations.

The issue of Mark I drywell liner melt-through at Peach Bottom was assessed by the 

NUREG-1150 molten core-containment interaction panel. The results of expert panel elicitation 

are reported in NUREG/CR-4551, Volume 2, Revision 1, Part 2, “Evaluation of Severe Accident 

Risks: Quantification of Major Input Parameters, Experts’ Determination of Containment Loads

and Molten Core Containment Interaction Issues,” issued April 1991[11]. Analyses performed 

by the group of experts identified several areas of uncertainty in the phenomena governing debris 

transport out of the reactor pedestal toward the drywell liner. Subsequent to the completion of 

NUREG-1150, the NRC sponsored analytical and experimental programs to resolve the “Mark I 

liner attack” issue. The results of an assessment of the probability of Mark I containment failure 

by melt attack of the liner were published in NUREG/CR-5423, “The Probability of Liner 

Failure in a Mark I Containment,” issued in 1989 [12] and NUREG/CR-6025, “The Probability 

of Mark I Failure by Melt-Attack of the Liner,” issued in 1993 [13]. This work 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. In contrast, liner melt-through was 

determined to be likely under conditions in which the volume and temperature of core debris 

released to the drywell floor would support lateral debris movement toward, and contact with, 

the drywell liner.
4-44

None of the scenarios examined in the SOARCA BWR evaluation generate a pool of water on 

the drywell floor prior to vessel breach. Therefore, it was not necessary to model the effects of 

boiling heat transfer between core debris and on over-lying pool of water, and the resulting 

effects on debris mobility. However, lateral flow of a slurry of molten metallic and particulate 

oxides across a dry floor is modeled within the MELCOR framework, as described earlier in 

Section 4.5.
5-1

5.0 INTEGRATED THERMAL HYDRAULICS, ACCIDENT 

PROGRESSION, AND RADIOLOGICAL RELEASE ANALYSIS

This section describes the MELCOR accident progression analysis for the internal and external 

event scenarios described in Section 3.0 of this report. Version 1.8.6 of the MELCOR severe 

accident analysis code was used in the accident progression and radiological release calculations.

5.1 Long-Term Station Blackout – Unmitigated Response

The unmitigated scenario event progression for the LTSBO accident progression analysis 

assumes that the operators follow the actions described in Special Event Procedure SE-11 [3]. 

This document provides guidelines for managing the plant with degraded AC power sources. 

Initial operator actions would concentrate on assessing plant status. Successful reactor scram, 

containment isolation, and automatic actuation of RCIC for reactor level control would be 

verified. These checks would take approximately 15 minutes. One or more SRVs would cycle 

to control the RPV pressure.

Special Event Procedure SE-11 directs operators to align the station blackout line from 

Conowingo Dam in the event of failure of offsite power combined with the failure of all diesel 

generators to start. If this action fails to restore AC power to the plant, as assumed in this 

analysis, operators are directed to de-energize all non-essential DC loads. By removing as many 

unnecessary loads as possible from the DC bus, the station battery lifetime is extended. This 

load shedding would not affect or disable control logic to the RCIC, HPCI, main control room 

instrumentation, or SRV control.

The load shedding is expected to begin 15 minutes into the event and take approximately 

15 minutes to complete. Plant system engineers estimate the effect of load shedding would be to 

extend station battery duration from 2 to 4 hours.

One consequence of station blackout is the loss of cooling to the RCIC and HPCI corner rooms. 

Heat losses from system piping and equipment to the room atmosphere would cause these areas 

to overheat. In such an event, a step of the Special Event Procedure SE-11 is applicable. It 

directs operators to block open doors to these rooms and facilitate cross ventilation, which would 

slow the rate of room heat up. These actions are assumed to successfully prevent system 

isolation from high room temperature for the entire period of system operation.22 A subsequent 

step of the procedure directs operators to defeat high torus temperature isolation signals for HPCI 

and RCIC (if operating). MELCOR calculations presented in Section 5.1.1 indicate these signals 

would not be received before the station batteries exhaust; therefore, these actions are not 

important for the LTSBO scenario. Another step of the Special Event Procedure directs the 

operators to monitor the inventory in the CST and take actions to refill the tank23 via gravity feed 

 

22 Heat loss from RCIC (or HPCI) systems to their enclosure corner rooms is not explicitly represented in the 

MELCOR model. 23 If the seismic initiating event ruptured the CST, water would be retained in the reinforced concrete moat 

surrounding the tank and suction would continue to be available. The total (integral) volume of coolant 

drawn from the CST during the entire period of RCIC operation in the LTSBO sequence is approximately 

110,000 gal, which is considerably less than the tank capacity (approximately 200,000 gal).
5-2

from other sources if necessary. Long-term viability of the CST is therefore assumed in the 

MELCOR calculations.

Table 5-1 lists the calculated timing of key events that follow from all of these actions. The time 

at which core damage begins strongly depends on the duration of station batteries. The 

difference in time between loss of DC power and the onset of core damage increases as battery 

lifetime increases because of reductions in decay heat levels with time. In the absence of 

effective manual intervention, core damage eventually proceeds to melting and relocation of core 

material into the reactor vessel lower head, reactor vessel lower head failure, and release of 

molten core debris to the drywell floor.

Table 5-1 Timing of Key Events for LTSBO

Event

LTSBO with 

4 hr DC power

(time in hours unless 

noted otherwise)

Station blackout loss of all onsite and offsite AC power 0.0

Low-level 2 and RCIC actuation signal 10 minutes

Operators manually open SRV to depressurize the reactor vessel 1.0

RPV pressure first drops below LPI setpoint (400 psig) 1.2

Battery depletion leads immediate SRV re-closure 4.0

RCIC steam line floods with water RCIC flow terminates 5.2

SRV sticks open because of excessive cycling 8.2

Downcomer water level reaches top of active fuel 8.4

First hydrogen production 8.9

First fuel-cladding gap release 9.1

First channel box failure 9.3

Reactor vessel water level reaches bottom of lower core plate 9.3

First localized failure of lower core plate 9.6

First core cell collapse because of time at temperature 9.8

Beginning of large-scale relocation of core debris to lower plenum 10.5

Lower head dries out 13.3

Ring 5 CRGT column collapse (failed at axial level 2) 15.8

Ring 1 CRGT column collapse (failed at axial level 2) 17.4

Ring 4 CRGT column collapse (failed at axial level 2) 17.4

Ring 3 CRGT column collapse (failed at axial level 2) 17.5

Ring 2 CRGT column collapse (failed at axial level 2) 18.6
5-3

Table 5-1 Timing of Key Events for LTSBO (continued)

Event

LTSBO with 

4 hr DC power

(time in hours 

unless noted 

otherwise)

Lower head failure 19.7

Drywell head flange leakage begins 19.9

Hydrogen burns initiated in drywell enclosure region of reactor building 20.0

Refueling bay to environment blowout panels open 20.0

Hydrogen burns initiated in reactor building refueling bay 20.0

Drywell shell melt-through initiated and drywell head flange re-closure 20.0

Hydrogen burns initiated in lower reactor building 20.1

Door to environment through railroad access opens because of

overpressure

20.1

Refueling bay roof fails due to overpressure 20.2

Time iodine release to environment exceeds 1% of initial core inventory 23.6

Calculation terminated 48.0

The absence of water on the drywell floor in a transient scenario like a station blackout24 allows 

core debris ejected from the reactor vessel after lower head failure to spread laterally across the 

floor and contact the drywell wall. Past calculations have predicted drywell shell melt-through 

to occur relatively soon after vessel failure (i.e., within 30 minutes.) Fission product release 

from the containment to the reactor building and with a very short delay to the environment will 

begin at this point in time. Several release points to the environment are possible, depending on 

the response of the reactor building. Past calculations have shown that hydrogen combustion 

leads to nearly immediate opening of the refueling bay blow-out panels and the railroad doorway 

at grade level. Blow-out panels into the turbine building and personnel access doorways out of 

the reactor building might also open. However, the dominant flow path for fission products to 

the environment is through the refueling bay blowout panels.

25

5.1.1 Thermal Hydraulic Response

When plant conditions are stabilized, Special Event Procedure SE-11 calls for a controlled 

depressurization of the RPV to 125 psig using the instructions in the RC/P leg of Trip 

Procedure T-101. Depressurization would be accomplished by opening one or more SRVs or, if 

necessary, by manually opening other steam vent pathways, such as main steam line drains. The 

cooldown rate would be limited to less than 100 °F per hour. A controlled depressurization is 

initiated at 1 hour by opening a single SRV. As shown in Figure 5-1, this results in a 

 

24 As opposed to a loss-of-coolant accident, where reactor coolant effluent accumulates on the drywell floor. 25 A stable flow of air is calculated to enter the building through the open railroad doorway, rise upward 

through the open equipment hatches from grade level to the refueling bay and exit the building to the 

environment through the open blow-out panels.
5-4

stable pressure of approximately 125 psig. Reactor vessel pressure remains near this pressure for 

approximately 2 hours, while active DC power permits a SRV to hold in the open position. 

Four hours into the scenario, however, DC power from the station batteries is exhausted, and the 

solenoid valve regulating control air to the SRV operator closes, causing the SRV itself to 

reclose.26 SRV closure causes reactor vessel pressure to gradually increase back to its automatic 

(safety) lift setpoint. Reactor vessel pressure subsequently cycles about its lift setpoint for 

approximately the next 2 hours.

During the first 12 hours of the accident scenario, the reactor vessel water level is also 

undergoing significant changes (see Figure 5-2). The hydraulic transient immediately following 

reactor scram and isolation results in a gradual decrease in water level because of coolant 

evaporation and discharge through a cycling SRV to the suppression pool. RCIC automatically 

starts 10 minutes after the initiating event and begins to restore reactor water level. Two hours 

into the scenario, operators take manual control of RCIC and maintain level within the indicated 

range of +5 to +35 inches (i.e., 16 feet above the top of active fuel – TAF).

Figure 5-1 LTSBO vessel pressure

 

26 Loss of control air pressure to the valve operator might take a few minutes to affect valve position, but this 

short time is ignored in this analysis.

0

200

400

600

800

1000

1200

1400

0 2 4 6 8 10 12 14 16 18 20 22 24

Pressure [psig]

time [hr]

RPV Pressure

Batteries exhaust

- SRV recloses

Operator manually

opens 1 SRV SRV seizes open

Lower head

failure

Initial debris

relocation into

lower head
5-5

Figure 5-2 LTSBO reactor pressure vessel water level

When DC power from the station batteries expires 4 hours into the scenario, RCIC turbine speed 

is assumed to remain fixed at a nearly full-open position. The loss of electric (DC) power would 

cause the turbine inlet throttle valve to move to a full-open position. If RPV pressure is above 

approximately 400 psig, the additional steam flow resulting from further opening of the steam 

throttle valve would result in an automatic trip of the system due to steam turbine overspeed. At 

the time of battery exhaustion in this scenario, however, RPV pressure is well below 400 psig, 

and the steam inlet valve would already be positioned at or near its full-open position to permit a 

sufficient flow of steam into the RCIC turbine. As a result, a change in valve position is not 

anticipated in this situation, and a turbine overspeed trip would not be an immediate consequence 

of the loss of DC power. The functional effect of this response is that, for more than 1 hour

(i.e., until RPV pressure increases above 400 psig), the RCIC system continues to deliver coolant 

flow at approximately the same flow rate it had at the time DC power expired.

27 However, 

closure of the SRV at 4 hours means that coolant losses from the reactor vessel are temporarily 

terminated. Therefore, the reactor vessel level begins to rise (i.e., coolant injection continues, 

but losses are terminated.) A continuous rise in level is evident in Figure 5-2, between 4 hours 

and approximately 5.2 hours.

 

 

27 Steam flow to the RCIC turbine increases as RPV pressure increases following closure of the SRV. Based 

on conversations with licensee system engineers, this would lead to a turbine over-speed trip of the system 

above a pressure of approximately 400 psig. However, results of the MELCOR calculation for this 

scenario indicate flooding of the main steam line and attendant termination of RCIC operation would occur 

before RPV pressure reached this point.

0

100

200

300

400

500

600

700

800

0 2 4 6 8 10 12 14 16 18 20 22 24

Two Phase Mixture Level [in]

time (hr)

In-Shroud

Downcomer

RPV Water Level

Automatic RCIC

actuation

Operator takes manual

control of RCIC

RCIC steam

line floods

Large scale debris

relocation into

lower head

+5 to +35"

Batteries exhaust

- SRV recloses

Main Steam Line Nozzle

Top of Active Fuel (TAF)

Bottom of Active Fuel (BAF)

SRV sticks open
5-6

At 5.1 hours, the water level in the reactor vessel increases above the elevation of the main steam 

line nozzles. Water subsequently spills over into the main steam lines causing the steam line to 

the RCIC turbine to flood within a few minutes. The resulting termination of RCIC operation at 

5.2 hours causes the reactor water level to stabilize. Approximately one hour later, the average 

water temperature in the reactor vessel increases to saturation. When that occurs, the reactor 

vessel pressure is above 1000 psia and increasing. This increasing reactor vessel pressure causes 

a slight increase in the effective level of water because of decreasing average coolant density.28 

At 6.4 hours, reactor vessel pressure returns to the SRV lift pressure, and coolant losses through 

the cycling SRV resume. Without any form of coolant makeup, the reactor water level 

continuously decreases at a rate of 2 inches per minute. At 8.2 hours, the lead SRV fails to 

reclose after opening for the 270th time (refer to Section 4.4.2.1 for the technical basis for this 

event.) The stuck-open SRV reduces RPV pressure below 200 psia within one hour, and to 

equilibrium with the containment pressure in two hours.

29 The enhanced rate of coolant 

discharge from the reactor vessel causes the rate at which reactor water level decreases to 

accelerate. Water level decreases below TAF approximately 15 minutes after SRV seizure. 

One hour later, the level decreases below the bottom of the lower core plate. Within the first 

10 hours of station battery exhaustion, the entire inventory of water in the reactor vessel is 

evaporated (see Figure 5-2 and Table 5-1). 

The thermal response of fuel in the core is illustrated in Figure 5-3, which shows the calculated 

temperature of fuel cladding across the core mid-plane. Temperatures of fuel cladding at the top 

of the core begin to rise when the mixture level decreases below approximately two-thirds of the 

core height. As the mixture level decreases toward the bottom of the core, fuel temperatures 

increase rapidly due to runaway oxidation of Zircaloy cladding. The close relationship between 

the rate at which fuel cladding temperature increase and Zircaloy oxidation is shown in 

Figure 5-3, which compares clad temperatures (left-hand scale) to total in-vessel hydrogen 

generation (right-hand scale). Mechanical failure of fuel at the top of the core occurs when 

Zircaloy clad material either melts and drains to lower regions of the core, or oxides to form a 

thin, fragile ZrO2 shell around over-heated fuel. This mechanically weak material fragments into 

particulate debris, which relocate toward the lower core plate as rubble. 

 

28 The density of saturated water decreases by 4 to 5 percent as pressure increases from 900 psia to 1,150 psia. 

This causes the entire body of water within the core shroud to expand slightly, resulting in an increase in 

effective (swollen) water level. 29 If RPV depressurization did not occur due to seizure of the cycling SRV, low RPV pressure prior to vessel 

breach could be achieved by invoking plant-specific procedures in SAMG-1 (T-252), which offer means of 

depressurization that do not require electric power to position valves (i.e., manual operation).
5-7

Figure 5-3 LTSBO fuel cladding temperatures at core mid-plane and in-vessel 

hydrogen generation

-200

0

200

400

600

800

1000

0

500

1000

1500

2000

2500

3000

0 2 4 6 8 10 12 14 16 18 20 22 24

Hydrogen Generated [kg]

Temperature [K]

time [hr]

Cladding Temperatures

at Core Mid-plane

Ring 1

Ring 2

Ring 3

Ring 4

Ring 5

In-vessel H2 Generation

Indicates intact fuel no

longer present at this location
5-8

Particulate and molten debris continue to move downward in the core until 10.5 hours, when the 

lower core plate yields, releasing molten core debris into the reactor vessel lower head. The 

interaction between hot debris and residual water in the lower head increases the rate of coolant 

evaporation, as indicated in Figure 5-2 with the increased (negative) slope of the in-shroud water 

level. It also causes the molten debris to freeze on surfaces of the control rod guide tubes, which 

are submerged in the large body of water that remains in the lower plenum. The changes in core 

geometry during this time frame, which are caused by the formation and downward relocation of 

molten and particulate debris, are illustrated in Figure 5-4.

Legend

Intact Fuel (UO2) and Cladding

Particulate Debris

Molten Pool

Steam and/or hydrogen (no structure)

Steel support structures

Core boundary heat structures

Water

Figure 5-4 Relocation of core material into the lower plenum during LTSBO (Note: The 

diagrams should in this figure reflect the type(s) of core material resident in each spatial cell of the in-vessel 

MELCOR nodalization as a snapshot in time. The order in which different materials within a cell does not 

necessarily reflect their precise spatial configuration.)

The cooling of core debris as it enters the water-filled lower plenum is also evident in Figure 5-5, 

which shows the calculated temperature of debris along the inner surface of the lower head. 

When residual water in the lower plenum is completely evaporated at 13.3 hours, debris 

temperatures begin to increase. Heat transfer from debris to the inner surface of the lower head 

causes the lower head temperature to increase as well. This is illustrated in Figure 5-6, which 

depicts the calculated temperature on the inner and outer surfaces of the lower head across all 
5-9

five rings of the MELCOR model. Because reactor vessel pressure is relatively low during the 

heat up of debris in the lower plenum, the failure of the lower head is more strongly influenced 

by thermal rather than mechanical stresses.30

Figure 5-7 illustrates changes in the configuration of core debris and lower plenum structures 

between the time of RPV dryout (~13.3 hrs) and lower head failure (19.7 hours). At the time of 

lower head dryout, approximately 60% of the core fuel assemblies (i.e., the central three of five 

radial rings in the MELCOR model) have collapsed into the lower plenum. Highly oxidized, but 

vertically intact assemblies remain standing in the outer two rings of the core. Debris in the 

lower plenum surrounds a forest of intact control rod guide tubes (CRGTs). As indicated in 

Figure 5-5, the temperature of lower plenum debris steadily increases, eventually causing 

structural failure of the CRGTs, which collapse and are mixed into the growing debris bed. 

Collapse of the CRGTs supporting the outer two rings of fuel in the core causes this material to 

also fall into the debris bed. Immediately prior to lower head failure, the debris bed represents 

the mass of nearly the entire core plus structural materials below the core. This debris bed is 

composed of a mixture of molten stainless steel (~32% by mass), unoxidized zirconium (~12%) 

and particulate debris containing uranium dioxide and metallic oxides (the remainder). Failure 

of the lower head (19.7 hours) results in the rapid ejection of over 300 metric tons of core debris 

onto the floor of the reactor pedestal in the drywell. 

 

30 The inner surface temperature of the lower head (i.e., MELCOR rings 1-3) is above the melting point of 

steel at the time failure occurs.
5-10

Figure 5-5 LTSBO temperature of particulate debris on inner surface of lower head

Figure 5-6 LTSBO lower head temperature (IR-inner radius; OR-outer radius)

400

600

800

1000

1200

1400

1600

1800

2000

0 2 4 6 8 10 12 14 16 18 20 22 24

Temperature [K]

time [hr]

Lower Head Temperatures

Ring 1 - IR

Ring 1 - OR

Ring 2 - IR

Ring 2 - OR

Ring 3 - IR

Ring 3 - OR

Ring 4 - IR

Ring 4 - OR

Ring 5 - IR

Ring 5 - OR

Lower

head 

failure

Lower head dryout
5-11

Figure 5-7 Debris behavior from lower head dryout through lower head failure for 

LTSBO

Before the reactor vessel lower head fails, thermodynamic conditions in the containment are 

governed by the gradual release of hydrogen through the SRV to the torus. The large quantity of 

hydrogen (i.e., over 900 kg between 9 and 19 hours – see Figure 5-3), combined with the small 

free volume of the containment, results in significant increases in pressure. The containment 

pressure history is shown in Figure 5-8. Eighteen hours after the initiating event (i.e., 13 hours 

after the loss of all coolant injection), the containment pressure increases above the design 

pressure of 56 psig. Immediately before lower head failure (19.7 hours), containment pressure 

increased by a small amount more, to 58 psig. 

Containment atmosphere temperatures remain modest throughout the period of early increases in 

pressure because the steam/hydrogen mixture cools as it bubbles through the suppression pool. 

Changes in suppression pool temperature through the first 24 hours of the accident are shown in 

Figure 5-9. Immediately following vessel breach, containment atmosphere pressure and 

temperature increase dramatically from the accumulation of molten core debris on the reactor 

pedestal and drywell floors. The atmosphere temperature in the pedestal increases to over 

1500 K and the atmosphere in the drywell increases to a stable temperature of approximately 

490 K (420 oF). The combination of elevated pressure and temperature near the top of the 

drywell results in short-term leakage through the head flange. The leak area and discharge rate 

are assumed to be proportional to the differential pressure across the flange.31 Drywell head 

flange leakage begins almost 20 hours after the initial loss of offsite power. The initial leak rate 

is relatively small and is quickly overwhelmed by a different containment failure mode. 

 

31 The flange leak area, which is based on a structural analysis based on the containment internal pressure, is 

described in Section 4.6.
5-12

Figure 5-8 LTSBO containment pressure

Figure 5-9 LTSBO suppression pool temperature

0

10

20

30

40

50

60

70

80

90

100

0 10 20 30 40 50

Pressure [psig]

time [hr]

Containment Pressure

DW liner melt-through

(head flange leak recloses)

SRV sticks open

DW head flange

leakage begins

Debris relocation

into lower head

300

320

340

360

380

400

420

440

460

0 2 4 6 8 10 12 14 16 18 20 22 24

Temperature [K]

time [hr]

Suppression Pool 

Temperature

Pool Temperature

Saturation Temperature

DW head flange leakage begins (19.5 hrs)

DW liner melt-thru (19.7 hrs)

RCIC operation
5-13

Soon after debris is released onto the reactor pedestal floor, it flows laterally out of the cavity 

through the open personnel access doorway and spreads out across the main drywell floor. 

Lateral movement and spreading of debris across the drywell floor allows debris to reach the 

steel shell at the outer perimeter of the drywell within 10 minutes. Five minutes later, thermal 

attack of the molten debris against the steel shell results in shell penetration and opening of a 

release pathway for fission products into the basement (i.e., torus room) of the reactor building. 

The combined leakage through the drywell head flange and the ruptured drywell shell results in a 

rapid depressurization of the containment to approximately 25 psig, and then a gradual long-term 

depressurization, through the opening in the drywell liner.32 Before drywell shell melt-through 

occurs, hydrogen leaks through the drywell head flange and accumulates in the reactor building 

refueling bay.33 Within a few minutes, a flammable mixture develops and is assumed to ignite

when local hydrogen concentrations exceed 10 vol-%. The resulting increase in pressure within 

the building causes the blow-out panels in the side walls of the refueling bay to open, creating a 

release pathway to the environment. Small increases in internal pressure (0.25 psig) cause the 

blowout panels to open. Therefore, the panels in the sidewalls of the refueling bay offer a 

release pathway to the environment immediately after a hydrogen burn occurs within the 

building.

Several minutes after melt-through of the drywell liner, additional hydrogen is released from the 

drywell into the basement of the building (i.e., torus room) and is transported upward through 

open floor gratings into the ground level of the reactor building. Flammable mixtures quickly 

develop in several regions of the building, causing a sequence of several discrete combustion 

events. The pressure rise within the building causes several doorways to open, including the 

large equipment access doorway at grade level. Several other doorways also open within the 

building, including personnel access doorways into the building stairwells. The large opening at 

grade level, coupled with the open blow-out panels in the refueling bay at the top of the building, 

creates an efficient transport pathway to the environment for material released from containment. 

That is, a vertical column of airflow is created within the building whereby fresh air from outside 

the building enters through the open equipment doors at grade level, rises upward through the 

open equipment hatches at every intermediate floor within the building, and exits through the 

blow-out panels at the top of the building. As the next section will show, this chimney effect

reduces the effectiveness of the reactor building as an area for fission product retention.

5.1.2 Radionuclide Release

Volatile fission product release from fuel begins at 9.1 hours, when a portion of the fuel gap 

inventory is released due to early fuel cladding failures. As fuel temperatures rise (see 

Figure 5-3), diffusion-driven release of fission products out of the fuel matrix rapidly increases 

the amount of volatile species released into the reactor coolant system. The cumulative release 

of several volatile species from the fuel is shown in Figure 5-10.

 

32 Reduction in drywell internal pressure causes the drywell head flange leak pathway to reclose. 33 The precise leak pathway includes intermediate transport through the drywell head flange to the drywell 

head enclosure. Leakage from the enclosure into the refueling bay occurs through gaps in the concrete 

shield blocks on the refueling bay floor. This complex leak pathway is explicitly represented in the 

MELCOR model.
5-14

Figure 5-10 LTSBO In-vessel Fission Product Release from Fuel

The release of radionuclides that follows containment failure is shown in Figure 5-11 (see 

Appendix A for a detailed radionuclide core inventory). This release occurs in two steps because 

of sequential breaches in the containment boundary by two distinct failure modes. The first 

appearance of significant release to the environment begins at 19.9 hours, when leakage begins 

through the drywell head flange. The leak area associated with this failure mode is small and the 

leak pathway is open for a short time because within 5 minutes a larger leak34 develops when 

molten debris flows across the drywell floor and penetrates the steel liner after lower head 

failure. The two phases of release are, therefore, indistinguishable in Figure 5-11, which appears 

to show a single large ‘puff’ release at approximately 20 hours.

The amount (fraction of initial core inventory) of several important radionuclide species released 

to the environment continues to increase for several hours after the puff release that accompanies 

containment failure. Two processes cause the protracted release that occurs over a 4 to 6 hour 

period after containment failure, which is highlighted in the expanded view of the release 

signature shown in Figure 5-11(b). First, molten corium-concrete interactions (MCCI) on the 

drywell floor drive the residual quantity of volatile fission products from fuel debris, and release 

 

34 The maximum containment pressure that occurs prior to drywell liner melt-through is approximately 

85 psig. Therefore, based on the head flange leakage evaluation described in Section 4.6, the maximum 

leak area through the flange is approximately 2 in2 (0.001 m2

). In comparison, the assumed opening 

created by molten debris penetration of the drywell liner is 1 ft2 (0.1 m2

).

1.E-03

1.E-02

1.E-01

1.E+00

8 9 10 11 12

Fraction of Initial Inventory

time [hr]

NG, I, Cs, Te

Mo

Ba

Ru
5-15

a relatively small fraction of all nonvolatile species. Second, the combination of high drywell 

atmosphere temperatures generated as a byproduct of MCCI and heating of reactor vessel 

internal structures because of decay heating of deposited radionuclides results in a late 

revaporization release of volatile species from within the containment and RCS. The latter is 

described in greater detail below.

Figure 5-12 depicts the fraction of the initial iodine inventory that is captured in the suppression 

pool, deposited or airborne within the RPV and the drywell, and released to the environment as a 

function of time. Similar information is shown in Figure 5-13 for cesium, in Figure 5-14 for 

tellurium, and in Figure 5-15 for non-volatile cerium. Collectively, these figures provide useful 

information about the mobility of different radionuclide species and the temporal changes in 

their spatial distribution. For example, next to noble gases, iodine is the most volatile 

radionuclide group. In the SOARCA calculations, iodine is assumed to be transported in the 

form of cesium iodide (CsI), which vaporizes at relatively modest temperatures for a severe 

accident. As a result, CsI is released from fuel during the early phases of in-vessel core damage 

progression, and a significant fraction remains airborne because of the relatively high 

temperatures of structures within the reactor vessel. Airborne iodine is efficiently transported to 

the wetwell through the open SRV. In particular (see Figure 5-12), approximately 60 percent of 

the initial core inventory of iodine is discharged from the RPV to the suppression pool during the 

blowdown of the reactor vessel that occurs during the same time that volatile fission products are 

released from the fuel (see Figure 5-10). At approximately 10.5 hours, an additional 30% of the 

iodine inventory is purged from the RPV to the suppression pool by steam that is generated in 

the RPV lower plenum when a large quantity of core debris collapses into the pool of water in 

the lower head. During the succeeding 6 hours, the majority of CsI that remains deposited on 

reactor vessel internal structures after RPV blowdown evaporates from these surfaces as a result 

of decay heating, and is also carried into the suppression pool.

The release of iodine (CsI) and tellurium to the environment increases beginning after 23 hours

after initially stabilizing at values below 1% of the initial core inventory (refer to Figure 5-11). 

The source of this delayed release is CsI and Te aerosol that was initially captured in water 

trapped in the RPV downcomer (i.e., below the top of the jet pumps.) As indicated in Figure 5-2, 

the rate at which this water evaporates reduces dramatically when core debris relocates from the 

core region to the lower plenum at approximately 10.5 hours (i.e., heat losses through the core 

shroud are reduced when debris falls into the lower plenum after lower core plate failure). Total 

dryout of the downcomer occurs at 23 hours, and is indicated in Figure 5-2 by a plateau in 

downcomer level at an elevation of approximately 122 in, which corresponds to the elevation of 

the floor or ‘baffle plate’ of the downcomer. When water is totally evaporated from a control 

volume, MELCOR deposits all radioactive aerosols captured in water on the heat structure 

representing the ‘floor’ of the control volume -- in this case, the heat structure representing the 

baffle plate. Subsequent heat up of this structure due to decay heating increases the structure 

temperature, and results in the evaporation of CsI and Te into the RPV (downcomer) atmosphere. 

At the time this occurs, the RPV lower head has failed allowing airborne fission products to be 

carried directly into the drywell, and the drywell liner has failed allowing airborne fission 

products in the drywell to be transported to the environment. Therefore, late revolatilization of 

iodine and tellurium from the downcomer results in a direct increase in the environmental release 

for these species.
5-16

(a)

(b)

Figure 5-11 LTSBO Source Term to the Environment

1.E-06

1.E-05

1.E-04

1.E-03

1.E-02

1.E-01

1.E+00

0 6 12 18 24 30 36 42 48

Fraction of Initial Core Inventory

time [hr]

Fission Products 

in Environment

NG

I & Te

Cd

Ba

Sn

Ce

La

Ru

Mo

Containment failure

Cs

see next figure for 

detail in this range

1.E-06

1.E-05

1.E-04

1.E-03

1.E-02

1.E-01

1.E+00

18 19 20 21 22 23 24

Fraction of Initial Core Inventory

time [hr]

Fission Products 

in Environment

NG

I

Cd

Te

Ba

Sn

Ce

La

Ru

Mo

Containment

failure

Cs
5-17

Figure 5-12 LTSBO iodine fission product distribution

Figure 5-13 LTSBO cesium fission product distribution

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

1

0 6 12 18 24 30 36 42 48

Fraction of Initial Core Inventory

time [hr]

Iodine Distribution

Release to

env. (2.0%)

Captured in 

Suppression Pool

Deposited/Airborne

within RPV

Drywell 

(negligible)

Released

from fuel

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

1

0 6 12 18 24 30 36 42 48

Fraction of Initial Core Inventory

time [hr]

Cesium Distribution

Release to

environment (0.5%)

Captured in 

Suppression Pool

Deposited/Airborne

within RPV and 

steam lines

Drywell 

(negligible)

Released

from fuel
5-18

Figure 5-14 LTSBO tellurium fission product distribution

Figure 5-15 LTSBO cerium fission product distribution

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

1

0 6 12 18 24 30 36 42 48

Fraction of Initial Core Inventory

time [hr]

Tellurium Distribution

Release to

environment (2.2%)

Captured in 

Suppression Pool

Deposited/Airborne

within RPV

Drywell 

(negligible)

Released

from fuel

0.0E+00

2.0E-05

4.0E-05

6.0E-05

8.0E-05

1.0E-04

1.2E-04

1.4E-04

0 6 12 18 24 30 36 42 48

Fraction of Initial Core Inventory

time [hr]

Cerium Distribution

Release to

environment (0.1%)

RPV and Wetwell

(negligible)

Drywell

Released

from fuel
5-19

The spatial distribution of cesium does not change with time in the same way the distribution 

changed for iodine. This is evident in Figure 5-13, and also in the relatively small changes in 

release to the environment relative to the initial puff release at containment failure (i.e., compare, 

for example, the release signatures for iodine and cesium in Figure 5-11). A slightly larger 

fraction of the cesium released from fuel deposits on the steam separators and dryers above the 

core in the RPV than is observed for iodine (i.e., 28% versus 18%). More importantly, however, 

a larger proportion of deposited cesium is retained on these structures as in-vessel damage 

progression proceeds over time. 

These differences in iodine and cesium behavior can be attributed to differences in the physical 

properties of their dominant chemical forms. As mentioned earlier, iodine is transported as CsI. 

The cesium contribution to CsI represents only 6 percent of the total cesium inventory. The vast 

majority (approximately 90 percent)35 of the cesium inventory is transported in the form of 

cesium molybdate (Cs2MoO4). Cesium molybdate is less volatile than the iodide and remains 

deposited on in-vessel structures at significantly higher temperatures. The in-vessel temperature 

history calculated for the LTSBO creates a thermal environment that promotes a greater 

evaporation of CsI relative to that of Cs2MoO4. Therefore, iodine is preferentially transported to 

the torus, but cesium remains deposited on in-vessel structures.

The suppressed mobility of cesium compared to iodine also affects the ultimate quantity 

transported to the environment in the first few hours after containment failure. A large fraction 

of iodine (CsI) initially captured on in-vessel structure surfaces evaporates and is released 

directly into the drywell atmosphere through the ruptured RPV lower head and to the 

environment through the failure in the drywell liner. Conversely, the vast majority of cesium 

molybdate deposited on in-vessel structures remains there throughout the late phases of accident 

progression.

The behavior of tellurium (Figure 5-14) is similar to that described above for iodine, and is not 

described in further detail here. Release of the heavy non-volatile species (e.g., cerium) differs 

substantially from the trends described above for any of the volatile species. As indicated in

Figure 5-15, the release of the non-volatile refractory metals does not begin until after vessel 

breach, when MCCI occurs on the drywell floor. Release of cerium and other similar species 

(e.g., lanthanum (La) and ruthenium (Ru)) from fuel debris begins soon after vessel breach when 

MCCI is most aggressive. As indicated in Figure 5-16, the temperature of ex-vessel debris 

decreases significantly as it spreads across the drywell floor from its initial point of arrival in the 

reactor pedestal. This greatly reduces the rate at which the non-volatile species are released.

 

35 The remaining fraction is cesium located in the fuel cladding gap.
5-20

Figure 5-16 LTSBO ex-vessel debris temperatures

An earlier discussion of the physical response of the reactor building (Section 5.1.1) noted that 

hydrogen burns greatly reduce the effectiveness of building to retain fission products released 

after containment failure. The principal cause of low residence time in the building after 

containment failure is combustion of hydrogen released from containment, which generates an 

increase in internal pressure large enough to open blowout panels at the top of the reactor 

building and fail the large equipment access doorway at grade level as well as several doors 

within the building. Flow through these openings in the reactor building creates a pathway for 

air to enter the building through the open equipment access doorway, rise through the vertical 

equipment hatches through the multiple floors of the building, and out the open blowout panels. 

Relatively large flow rates along this pathway during the peak period of fission product release 

reduce the average residence time for aerosols in the building below values that permit 

gravitational setting to operate. Therefore, negligible attenuation of fission product aerosols 

released from the containment occurs before release to the environment.

Later in time when containment depressurization is complete and hydrogen release rates reduce 

to levels that no longer support combustion, a smaller but stable flow of air along the same flow 

path is sustained by buoyancy effects. The temperature of gases released from the drywell to the 

reactor building is between 700 and 1100 K.

36 This hot gas enters the reactor building at a low 

elevation37 and rises through the same vertical flow path created by the original hydrogen 

 

36 Refer to Section 5.7.2 for details on the calculation of drywell atmosphere temperature. 37 The opening in the drywell shell would likely create a flow path into the basement of the reactor building 

where the torus vent pipes are located.

0

200

400

600

800

1000

1200

1400

1600

1800

2000

0 10 20 30 40 50

Temperature [K]

time [hr]

Ex-vessel Debris Temperatures

CAV0: In-pedestal

CAV1: Quarter-floor outside doorway

CAV2: Remainder of DW floor

CAV 1 CAV 2

6.706 m 3.086 m

0.914 m

 Sump

Manway

CAV 0

Tablation

Tliquidus
5-21

combustion events. The thermal plume created by this buoyant gas rises at a velocity of 2 to 

3 m/s, which is large enough to sweep the vast majority of airborne aerosols to the environment 

with little opportunity for deposition. This low residence time created by this persistent air flow 

is compounded by the fact that the aerosol particles released late in time, due to revaporization 

from RCS surfaces, are very small. The calculated mass median diameter of airborne aerosols 

transported from the drywell through the reactor building after containment depressurization is 

less approximately 0.5 micron. These two factors combine to produce a global decontamination 

factor (DF) for the reactor building barely over 1.0. Figure 5-17 shows the calculated reactor 

building DF for iodine and cesium.

Figure 5-17 Reactor Building DF for unmitigated LTSBO

5.2 Long-Term Station Blackout with Mitigation

Table 5-2 outlines the key events for LTSBO with mitigative actions (discussed in 

Section 3.1.3). 

0

1

2

3

4

5

6

7

8

9

10

0 6 12 18 24 30 36 42 48

Decontamination Factor (DF)

time [hr]

Decontamination Factor

in Reactor Building

DF = (mass in/out)

containment failure

early leakage

from containment

Iodine

Cesium
5-22

Table 5-2 Timing of Key Events for Mitigated LTSBO

Event

Mitigated LTSBO 

with 4 hr DC power

(Time in hours unless 

noted otherwise)

Station blackout loss of all onsite and offsite AC power. 0.0

Automatic reactor scram and containment isolation 0.0+

Low-level 2 and RCIC actuation signal 10 minutes

Operators manually open SRV to depressurize the reactor vessel 1.0

RPV pressure first drops below LPI setpoint (400 psig) 1.2

Operators take manual control of RCIC; flow throttled to maintain 

level within range (+5 to +35 in) 2.0

Portable electric generator positioned, started, and connected to 

remote panel 1.0 to 4.0

Station batteries depleted 4.0

Operators position, align, and start portable pump to replace RCIC as 

injection source, if needed 4.0 to 10.0

Portable electric power and gas bottles staged near isolation valves 

for containment hard pipe vent line 10.0

Hard pipe containment vent line opened/closed at containment 

pressures of 45/25 psig

22.4 / 23.3

29.6 / 30.5

37.1 / 38.0

44.4 / 46.2

Calculation terminated 48

5.2.1 Thermal Hydraulic Response

The thermal hydraulic response during the first four hours of the mitigated case matches the 

response of the unmitigated case. RCIC automatically actuates to restore reactor water level. 

Operators manually open an SRV to reduce pressure in the RPV (see Figure 5-18), then assume 

manual control of RCIC flow to maintain water level within a specified range. Beyond four 

hours, however, the sequence of events for the mitigated case differs from the unmitigated case. 

Before DC power from the station batteries expires at 4 hours, a portable generator is engaged to 

maintain a long-term supply of control power. This permits the open SRV to remain open, 

maintaining reduced pressure in the RPV according to Special Event Procedure SE-11. Using 

the RPV water level information, which is available by use of a portable generator, operators 

manually control (locally) RCIC to maintain a stable water level in the RPV, as shown in 

Figure 5-19, and maintain fuel cooling as indicated in the plot of peak clad temperature shown in

Figure 5-20.
5-23

Figure 5-18 Mitigated LTSBO vessel pressure

Figure 5-19 Mitigated LTSBO coolant level

0

200

400

600

800

1000

1200

1400

0 6 12 18 24 30 36 42 48

Pressure [psig]

time [hr]

Operator manually RPV Pressure

opens 1 SRV 

Station batteries exhausted;

portable power supply engaged

to sustain open SRV
5-24

Figure 5-20 Mitigated LTSBO core temperature

The mitigation measures do not restore function to systems that provide containment heat 

removal. Therefore, energy discharged from the RPV through the open SRV gradually increases 

the temperature of water in the suppression pool. As shown in Figure 5-21, the temperature of 

water in the torus reaches saturation conditions in approximately 16 hours, which increases the 

rate at which containment pressure (see Figure 5-22) rises. Prior to this point in time, however, 

suppression pool temperature and pressure increase to values that would either trigger isolation 

signals for the RCIC system or invoke instructions for further RPV depressurization. For 

example, as indicated in Figure 40, RPV pressure exceeds the Heat Capacity Temperature Limit 

(HCTL) within approximately 5 hours, and would reduce below RCIC isolation setpoint for low 

steam line pressure in approximately 38 hours. Operators are instructed to take actions to ensure 

continued operation of RCIC under station blackout conditions, such as defeating isolation 

signals if they occur. The time at which one example signal is calculated to occur is indicated 

in Figure 5-19. Heating of the suppression pool reduces the HCTL, which would normally 

demand further reductions in RPV pressure. However, RPV pressure must be maintained above 

the minimum operating conditions for RCIC to ensure a sufficient steam supply to the RCIC 

turbine. Therefore, it is assumed no actions are taken to further reduce RPV pressure. 

Approximately 10 hours into the scenario, pool temperatures increase above 378 K (220 °F). 

Above this temperature, RCIC isolation on loss of Net Positive Suction Head or inadequate 

cooling of RCIC pump bearings would be possible if suction were taken from the torus. 

However, the mitigation measures assumed in this calculation maintain RCIC suction from the 

CST, thereby, eliminating challenges to RCIC operation due to high suppression pool 

temperature.

400

500

600

700

800

900

1000

0 6 12 18 24 30 36 42 48

Temperature [K]

time [hr]

Maximum Cladding Temperature in the Core
5-25

The timeline for implementing the mitigation measures (see Section 3.1.4.2) suggests portable 

electric power sources and gas bottles needed to open isolation valves on the 16-in hard pipe vent 

line would be staged for use as early as 10 hours after the initiating event. Therefore, the hard 

pipe vent is assumed to open at approximately 20 hrs, when containment pressure reaches 

45 psig, and is closed at 23 hours when containment pressure is reduced to approximately 25 

psig. Several open/close cycles of the containment vent isolation valves are necessary with a 

period of approximately 8 hours to manage containment pressure within this range. 

The principal basis for the containment venting criteria is based in part from comments received 

from the licensee during a verbal walk-through of LTSBO mitigation, which was part of a site 

visit in 2007. Based on these comments and others received from the Peer Review Panel, the 

venting criterion assumes the hard-pipe vent path will be opened at 45 psig and reclosed at 

25 psig (Figure 5-22). These values were selected based on a review of plant-specific procedures 

for containment pressure control, but also taking into consideration isolation setpoints for RCIC, 

which are sensitive to containment thermodynamic conditions. Most important in this regard is 

the setpoint for high turbine exhaust pressure. 

Figure 5-21 Mitigated LTSBO Suppression Pool Water Temperature

300

320

340

360

380

400

420

440

0 6 12 18 24 30 36 42 48

Temperature [K]

time [hr]

Suppression Pool Temperature

Tpool

Tsat

``
5-26

Figure 5-22 Mitigated LTSBO containment pressure

The Peak Containment Pressure Limit (PCPL) mentioned in containment pressure control 

procedures is 60 psig. However, a high turbine exhaust pressure isolation signal for RCIC would 

be received at a pressure of 50 psig. Since RCIC is the only operating coolant injection system 

available in this scenario, it was assumed operators would open the containment vent path at 

45 psig, thereby averting RCIC isolation. Long term containment pressure control would be 

managed by periodically opening and reclosing the vent line. The effects of vent line operation 

are clearly indicated in the pressure response shown in Figure 5-22.

The total quantity of water transferred from the CST to the RPV by continuous (throttled) 

operation of RCIC exceeds the nominal CST inventory in approximately 15 hours. Makeup 

water provided by the portable pump maintains the CST inventory for an indefinite period of 

time. The continuous supply of water from outside the containment pressure boundary ensures 

stable control of core cooling, but it also results in an accumulation of water in the torus. Steam 

evaporated from the RPV is discharged to the torus through the open SRV, where it is 

condensed. As indicated in Figure 5-23, this results in a slow increase in the suppression pool 

water level the time containment venting begins. Depressurization of the containment due to 

venting causes the torus water level to swell by approximately 1.5 ft. The torus level then 

collapses down to a value below the pre-vent level when the vent line recloses. This general 

behavior is repeated during subsequent periods of containment venting, resulting in a fluctuating 

water level that oscillates between approximately 18.4 and 20.5 ft. Although this value is well 

above target values for the torus water level (i.e., 14.5 to 15.5 ft), the peak value lies below the 

elevation at which the SRV solenoids would be flooded (~21 ft), thereby preventing any 

interference with manual SRV control.

0

5

10

15

20

25

30

35

40

45

50

0 6 12 18 24 30 36 42 48

Pressure [psig]

time [hr]

Drywell Pressure

Operators open

hard-pipe vent

Reclose

vent
5-27

Figure 5-23 Mitigated LTSBO torus water level

5.2.2 Radionuclide Release

No plots are presented for the iodine fission product distribution history, cesium fission product 

distribution history, barium fission product distribution history, cerium fission product 

distribution history, or environmental release history of all fission products resulting 

from mitigated LTSBO because the mitigated case does not result in core damage. 

5.3 Short-Term Station Blackout with RCIC Blackstart

The general response of plant equipment and operating personnel to the STSBO resembles the 

unmitigated LTSBO scenario. The reader is referred to Section 5.1 for a description of the 

actions that plant personnel would take in response to this type of event. However, a key 

difference is the assumed immediate failure of DC power, which significantly reduces the time 

available for intervention and accelerates the time line of damage progression. Successful 

manual actions to blackstart RCIC compensates for effects of the additional loss of DC power for 

a period of time. That is, the delayed (manual) actuation of coolant injection restores RPV water 

level, but RPV water level control via RCIC throttling is assumed not to be achieved. Therefore, 

although many similarities are observed in plant response between this scenario and the LTSBO 

discussed earlier, some important differences arise as well. The calculated timing of key events 

that follow is listed in Table 5-3. 

5.3.1 Thermal Hydraulic Response

The initiating event causes a prompt failure of all AC and DC power supplies to plant equipment 

and instrumentation. Reactor control blades, MSIVs, and containment isolation valves would all 

move to their fail-safe positions (inserted and closed). Isolation of the RCS causes reactor 

14

15

16

17

18

19

20

21

0 6 12 18 24 30 36 42 48

Water Level [ft]

time [hr]

Suppression Pool Level
5-28

pressure to rise to the set point of the SRVs, which open and direct coolant to the pressure 

suppression pool. As shown in Figure 5-24, reactor pressure is maintained at approximately 

1,100 psig, as the SRV with the lowest set point cycles open and closed for nearly 2.5 hours.38 

Actions taken by plant operations personnel to manually reduce reactor pressure and prevent 

frequent cycling of the SRVs are assumed to not be successful. 

Actions by plant personnel to blackstart the RCIC pump take one hour to accomplish, at which 

time coolant flow into the reactor vessel begins. As shown in Figure 5-25, the coolant inventory 

lost to evaporation during the first hour of the transient is quickly replenished and reactor water 

level is restored. The minimum reactor water level prior to RCIC actuation is approximately 

midway between the top and bottom of active fuel. The brief period of core uncovery allows 

fuel cladding temperatures to increase above 1000 K (see Figure 5-26) and a small portion of the 

cladding near the top of the core begins to oxidize. Restoration of coolant injection quenches the 

overheated fuel before clad failure occurs and temperatures throughout the core are quickly 

restored to equilibrium with reactor coolant.

 

38 A second SRV periodically opens during the first 45 minutes of the transient, when decay heat levels 

remain high. However, after this point in time, only one valve is cycling.
5-29

Table 5-3 Timing of Key Events for the short term station blackout with successful 

RCIC blackstart

Event

Time

(hours, unless 

otherwise noted)

Station blackout – loss of all onsite and offsite AC power 0.0

Low-level 2 and RCIC actuation signal (no RCIC response) 10 min

Downcomer water level reaches TAF 30 min

Hydrogen production begins 55 min

RCIC started manually and begins injection at full flow 1.0

Fuel cladding quenched and oxidation terminated (<2% oxidation) 1.2

RPV water temperature subcooled; depressurization begins 2.4

RPV pressure first drops below LPI set point (400 psig) 2.9

RPV water level above main steam line nozzles – RCIC stops due to 

steam turbine flooding 3.4

SRV sticks open due to excessive number of cycles 4.8

Downcomer water level decreases permanently below TAF 5.7

Hydrogen production begins again 6.6

First fuel-cladding gap release 6.8

First channel box failure 6.9

First core support plate localized failure in supporting debris 7.2

Reactor vessel water level reaches bottom of lower core plate 7.2

First core cell collapse due to time at temperature 7.4

Beginning of large-scale debris relocation into lower plenum 7.7

Lower head dries out 10.9

Ring 5 CRGT Column Collapse [failed at axial level 2] 13.2

Ring 1 CRGT Column Collapse [failed at axial level 2] 14.4

Ring 3 CRGT Column Collapse [failed at axial level 2] 14.6

Ring 4 CRGT Column Collapse [failed at axial level 2] 15.2

Ring 2 CRGT Column Collapse [failed at axial level 2] 15.6

Lower head failure 16.7

Drywell liner melt-through 16.9

Refueling bay to environment blowout panels open 16.9

Hydrogen burns initiated in lower reactor building 16.9

Doors to stairwell open due to overpressure 16.9

Door to environment through railroad access opens due to 

overpressure

16.9

Time Iodine release to environment exceeds 1% 23.6

Calculation terminated 48.0
5-30

Introduction of a large quantity of cold water to the RPV causes the average RPV water 

temperature to decrease below saturation conditions at approximately 2.4 hours. This reduces 

steam generation below levels necessary to sustain RPV pressure at the SRV set point. Valve 

cycling ceases at this time, and RPV pressure decreases below 400 psig within an hour. Steam 

supply to the RCIC turbine is assumed to be sufficient to permit the pump to continue operating 

at full capacity until 3.4 hours, when the RPV water level increases above the elevation of the 

main steam lines nozzles and water pours into the main steam lines. Flooding of the main steam 

line chokes the steam supply to the RCIC turbine, and pump operation ceases at 3.4 hours.

Figure 5-24 Reactor pressure for short term station blackout with RCIC blackstart

0

200

400

600

800

1000

1200

1400

0 2 4 6 8 10 12 14 16 18 20 22 24

Pressure [psig]

time [hr]

RPV Pressure

Lower head

failure

Beginning of large 

scale debris relocation 

into lower head

RCIC

operation

SRV sticks in 

open position 
5-31

Figure 5-25 Reactor vessel water level for STSBO with RCIC blackstart

Figure 5-26 Fuel cladding temperatures at core mid-plane and in-vessel hydrogen 

generation for STSBO with RCIC blackstart

0

100

200

300

400

500

600

700

800

0 2 4 6 8 10 12 14 16 18 20 22 24

Two Phase Mixture Level [in]

time (hr)

In-Shroud

Downcomer

RPV Water Level

RCIC

blackstart

RCIC steam

line floods

Initial clad failure

(first gap release)

Beginning of large scale 

debris relocation into

lower head

Lower head

dryout /

failure

Elevation of main steam line nozzle

TAF

BAF

-200

0

200

400

600

800

1000

0

500

1000

1500

2000

2500

3000

0 2 4 6 8 10 12 14 16 18

In-vessel Hydrogen Generation [kg]

Temperature [K]

time [hr]

Cladding Temperatures

at Core Mid-plane

Ring 1

Ring 2

Ring 3

Ring 4

Ring 5

H2 generation

Indicates intact fuel no

longer present at this location
5-32

Termination of steam flow to the RCIC turbine, combined with the loss of (cold) coolant 

injection to the RPV, results in a steady increase in RPV pressure. Figure 5-24 shows RPV 

pressure returning to the SRV setpoint approximately one hour after the main steam line floods 

and RCIC operation ceases. Afterward, the lead SRV begins to cycle again. However, within 

15 minutes, the cycling SRV sticks in the open position. This occurs due to stochastic failure, 

when the valve experienced a total of 270 cycles, as described in Section 4.4.2.1. RPV pressure 

decreases again as a result of the open SRV.

The permanent loss of coolant makeup following termination of RCIC operation causes the RPV 

water level to decrease below the top of active fuel at approximately 5.7 hours. Clad oxidation is 

renewed and, as shown in Figure 5-26, fuel temperatures begin to rise rapidly at approximately 

7 hours when clad temperature increase above 1500 K. Molten core debris forms near the top of 

the core and relocates downward in a configuration similar to images shown in Figure 5-4 and 

Figure 5-7 for the LTSBO scenario. Molten and particulate debris accumulates in the lower 

plenum and are initially quenched by the large body of residual water below the lower core plate. 

This water is entirely evaporated by 11 hours into the STSBO sequence and debris temperatures 

begin to rise (see Figure 5-27). Heat transferred from high temperature core debris causes the 

lower head structure to heat up as illustrated in Figure 5-28.

Figure 5-27 Temperatures of core debris along inner surface of lower head for STSBO 

with RCIC blackstart
5-33

Figure 5-28 Inner and outer surface temperatures of lower head for STSBO with 

RCIC blackstart (IR-inner radius; OR-outer radius)

Failure of the lower head (at 16.7 hours) results in the rapid ejection of over 300 metric tons of 

core debris onto the floor of the reactor pedestal in the drywell. The composition of this debris at 

the time of head failure is a mixture of molten stainless steel (about one-third by mass), 

unoxidized zirconium (~12%), and particulate debris containing uranium dioxide and metallic 

oxides (the remainder).

Before the reactor vessel lower head fails, thermodynamic conditions in the containment are 

governed by the release of hydrogen through the open SRV to the torus. The large quantity of 

hydrogen (i.e., over 960 kg within 10 hours), combined with the small free volume of the 

containment, results in a significant increase in pressure. The containment pressure history is 

shown in Figure 5-29. Immediately before lower head failure, containment pressure is 

approximately 53 psig, which is slightly lower than the corresponding base pressure in the 

LTSBO scenario. The energy accompanying the discharge of molten core debris after lower 

head failure causes this pressure to increase to 75 psig before containment failure occurs due to

thermal failure of the drywell shell as discussed below.

In contrast to the LTSBO (refer to Section 5.1.1), containment pressure immediately following 

reactor vessel failure is below the threshold for induced leakage through the drywell head flange 

(80 psig). Therefore, leakage from containment does not occur by this mechanism. This is due 

primarily to the slightly lower base pressure (i.e., prior to vessel breach) in the short-term 

scenario, which results from the shorter period of reactor steaming to the suppression pool 

(i.e., before to the onset of core damage). 

400

600

800

1000

1200

1400

1600

1800

2000

0 2 4 6 8 10 12 14 16 18 20 22 24

Temperature [K]

time [hr]

Lower Head Temperatures

Ring 1 - IR

Ring 1 - OR

Ring 2 - IR

Ring 2 - OR

Ring 3 - IR

Ring 3 - OR

Ring 4 - IR

Ring 4 - OR

Ring 5 - IR

Ring 5 - OR

Lower

head 

failure

Lower head dryout
5-34

Figure 5-29 Containment pressure history for STSBO with RCIC blackstart

Containment conditions change dramatically, when debris is released onto the reactor pedestal 

floor following lower head failure. The absence of water on the drywell floor allows debris to 

flow laterally out of the cavity through the personnel access doorway and spread out across the 

main drywell floor. Lateral movement and spreading of debris across the drywell floor allow

debris to reach the steel shell at the outer perimeter of the drywell within 10 minutes. 

Five minutes later, thermal attack of the molten debris against the steel shell results in shell

penetration and opening of a release pathway for fission products into the basement (i.e., torus 

room) of the reactor building. This results in a rapid depressurization of the containment to 

atmospheric conditions in a short time (less than 1 hour). 

Immediately following drywell shell melt-through, hydrogen is released from the drywell into 

the basement of the building (i.e., torus room) and is transported upward through open floor 

gratings into the ground level of the reactor building. Flammable mixtures quickly develop in 

these regions, which are assumed to ignite from high flammable gas concentration and high gas 

effluent temperatures exceeding 1500 K near the floor of the drywell. The resulting pressure rise 

within the building causes several doorways within the building to open, including the large 

equipment access doorway at grade level and the blow-out panels in the side walls of the 

refueling bay near the top of the building. This combination of two large openings in the reactor 

building creates an efficient transport pathway for material released from containment to the 

environment. That is, a vertical column of airflow is created within the building, whereby fresh 

air from outside the building enters through the open equipment doors at grade level, rises 

upward through the open equipment hatches at every intermediate floor within the building, and 

exits through the blowout panels at the top of the building. As was described earlier 

(Section 5.1.2), retention of fission products in the reactor building is small for most species 

because of the chimney effect that is created by this flow pattern.

0

10

20

30

40

50

60

70

80

90

0 6 12 18 24 30 36 42 48

Pressure [psig]

time [hr]

Drywell Pressure

DW liner melt-through

Initial clad

oxidation / 

RCIC starts 

RCIC

operation

Clad

oxidation

begins

again

RPV dryout

RPV lower

head failure
5-35

5.3.2 Radionuclide Release

The time-dependent release fraction for all radioactive species to the environment is shown in 

Figure 5-30. The release of radionuclides immediately accompanying containment failure (i.e., 

the puff release) is the dominant contributor to the release of activity to the environment. 

However, as indicated in the expanded view of the release fractions for volatile species shown in 

Figure 5-31, the release of some important species, iodine and tellurium in particular, continues 

for several hours. The initial ‘puff’ release accompanying containment failure primarily results 

from the discharge of fission products that were airborne in the containment atmosphere at the 

time of drywell liner melt-through.

Figure 5-30 Environmental source term for STSBO with RCIC blackstart

1.E-06

1.E-05

1.E-04

1.E-03

1.E-02

1.E-01

1.E+00

0 6 12 18 24 30 36 42 48

Fraction of Initial Core Inventory

time [hr]

Fission Products 

in Environment

NG

Ba

Ce

La

Ru

Mo

Containment failure

Cs

I, Te

Cd

Sn
5-36

Figure 5-31 Environmental source term for STSBO with RCIC blackstart: details for 

volatile species

Figure 5-32 summarizes the spatial distribution of the total iodine inventory as a function of 

time. Similar information is shown in Figure 5-33 for cesium and Figure 5-34 for tellurium. 

Collectively, these figures highlight differences in the mobility of different radionuclide species 

and temporal changes in their location. For example, all iodine is assumed to be transported in 

the form of CsI, which vaporizes at relatively modest temperatures for a severe accident and is 

readily swept from the RPV to the wetwell by the hydrogen/steam mixture that flows through the 

open SRV. The fractions of iodine, cesium, and tellurium swept into the suppression pool 

immediately are roughly in proportion to their relative volatility: iodine (90%); tellurium (83%); 

and cesium (75%). During the succeeding 6 hours, most CsI initially deposited on reactor 

vessel internal structures after RPV blowdown evaporates from the surfaces because of decay 

heating, and is swept into the suppression pool. Deposited tellurium behaves in a similar 

manner, but over a much longer time frame. Most of the cesium (i.e., the molybdate) deposited 

on in-vessel structures remains there for the duration of the 48 hour calculation.

1.E-03

1.E-02

1.E-01

1.E+00

0 6 12 18 24 30 36 42 48

Fraction of Initial Core Inventory

time [hr]

Fission Products 

in Environment

NG

I 

Cs

Ba

Te
5-37

Figure 5-32 Iodine fission product distribution for STSBO with RCIC blackstart

Figure 5-33 Cesium fission product distribution for STSBO with RCIC blackstart

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

1

0 6 12 18 24 30 36 42 48

Fraction of Initial Core Inventory

time [hr]

Iodine Distribution

Release to

environment (1.3%)

Captured in 

Suppression Pool

Deposited/Airborne

within RPV & main

steam lines

Drywell 

(negligible)

Released

from fuel

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

1

0 6 12 18 24 30 36 42 48

Fraction of Initial Core Inventory

time [hr]

Cesium Distribution

Release to

environment (0.4%)

Captured in 

Suppression Pool

Deposited/Airborne

within RPV + main stm lines

Drywell negligible

Released

from fuel
5-38

Figure 5-34 Tellurium fission product distribution for STSBO with RCIC blackstart

5.4 Short-Term Station Blackout- Sensitivity Case without RCIC Blackstart 

Another form of a STSBO scenario and a variation commonly found in BWR PRAs is one that 

does not credit actions by plant personnel to manual start steam-driven coolant injection systems 

(i.e., RCIC or HPCI). Immediate loss of all onsite and offsite electric power defeats all sources 

of emergency coolant injection, reactor pressure control and containment heat removal. As a 

result, the timeline of events in this version of a STSBO is accelerated in comparison to a

STSBO with RCIC blackstart and the LTSBO. The calculated chronology of events for the 

STSTO without RCIC blackstart is listed in Table 5-4. 

The onset of fuel damage occurs in approximately one hour. That is, oxidation of Zircaloy 

cladding in the upper region of the core leads to the release of the gap inventory of fission 

products in about one hour. Large-scale relocation of core debris to lower elevations of the core 

begins within two hours and core debris is released into the lower plenum at 2.4 hours. Lower 

head failure occurs in 8.2 hours, which releases molten core debris to the drywell floor and 

drywell liner melt-through in 8.5 hours. Significant fission product release to the environment,

39

therefore, begins at roughly ten hours of the initiating event.

 

39 ‘Significant’ in this context means the release of greater than 1% of the initial core inventory of iodine.

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

1

0 6 12 18 24 30 36 42 48

Fraction of Initial Core Inventory

time [hr]

Tellurium Distribution

Release to

environment (1.5%)

Captured in 

Suppression Pool

Deposited/Airborne

within RPV + main stm lines

Drywell (negligible)

Released

from fuel
5-39

Table 5-4 Timing of Key Events for the STSBO without RCIC blackstart

Event

Time

(hours, unless 

noted otherwise)

Station blackout – loss of all onsite and offsite AC power 0.0

Low-level 2 and RCIC actuation signal (no RCIC response) 10 min

Downcomer water level reaches TAF 30 min

First hydrogen production 55 min

First fuel-cladding gap release 1.0

First channel box failure 1.3

First core cell collapse due to time at temperature 1.6

SRV sticks open due to cycling at high temperatures 1.8

Reactor vessel water level reaches bottom of lower core plate 1.9

RPV pressure first drops below LPI setpoint (400 psig) 2.1

First core support plate localized failure in supporting debris 2.4

Lower head dries out 3.3

Ring 5 CRGT Column Collapse [failed at axial level 2] 4.9

Ring 3 CRGT Column Collapse [failed at axial level 2] 5.8

Ring 1 CRGT Column Collapse [failed at axial level 2] 6.0

Ring 2 CRGT Column Collapse [failed at axial level 2] 6.1

Ring 4 CRGT Column Collapse [failed at axial level 2] 6.3

Lower head failure 8.2

Drywell liner melt-through 8.5

Refueling bay to environment blowout panels open 8.5

Hydrogen burns initiated in lower reactor building 8.5

Doors to stairwell open due to overpressure 8.5

Door to environment through railroad access opens due to overpressure 8.5

Time Iodine release to environment exceeds 1% 9.7

Calculation terminated 48.0

5.4.1 Thermal Hydraulic Response

The loss of all AC and DC power supplies to plant equipment and instrumentation causes reactor 

control blades, MSIVs, and containment isolation valves to move to their fail-safe positions 

(i.e., inserted and closed). Isolation of the RCS causes reactor pressure to rise to the set point of 

the SRVs, which opens and directs coolant to the pressure suppression pool. As shown in 

Figure 5-35, reactor pressure is maintained at approximately 1,120 psia, as the SRV with the 

lowest set point cycles open and closed for approximately 2 hours.40 Actions taken by plant 

operations personnel to manually reduce reactor pressure and prevent frequent cycling of the 

 

40 A second SRV periodically opens during the first 45 minutes of the transient, when decay heat levels 

remain high. However, after this point in time, only one valve is cycling.
5-40

SRVs are assumed to not be successful. This is because control power (i.e., B.5.b portable 

generator and converter) to necessary equipment (e.g., SRV solenoid control valves) is assumed

not to be available and manual actions to open alternative steam relief paths41 are assumed not to 

be taken.

Figure 5-35 Reactor vessel pressure: STSBO without RCIC blackstart

Two hours after the initiating event, the single cycling SRV sticks in the open position due to 

thermal seizure of internal valve components.

42 This initiates a rapid depressurization of the 

RCS. The continuous discharge of steam through the open SRV accelerates the rate at which the 

coolant inventory is depleted from the RPV. Figure 5-36 shows the two-phase reactor mixture 

level in the downcomer and within the core shroud. Both levels show a sharp decrease at 

two hours, corresponding to the time of reactor blowdown through the open SRV. The steam 

flow produced by the flashing of residual water into steam temporarily cools over-heated fuel 

and in-core debris as shown in Figure 5-37, but also reduces the in-shroud mixture level well 

below the elevation of the lower core plate. Within an hour (i.e., less than 3 hours after the 

initiating event), the lower core plate yields, and debris begins to relocate into the lower plenum.

 

41 For example, SAMP-1 (RC/P-11) lists opening main steam line drains or HPCI and RCIC steam line drains 

as options for reducing RPV pressure. 42 Refer to Section 4.4 for details of this failure mechanism.

0

200

400

600

800

1000

1200

1400

0 2 4 6 8 10 12 14 16 18 20 22 24

Pressure [psig]

time [hr]

RPV Pressure

SRV seizes open

Lower head

failure

Beginning of 

large-scale debris

relocation to

lower plenum
5-41

Figure 5-36 Reactor vessel water level: STSBO without RCIC blackstart

Figure 5-37 Temperatures of fuel cladding at core mid-plane: STSBO without RCIC 

blackstart

0

100

200

300

400

500

600

700

0 2 4 6 8 10 12 14 16 18 20 22 24

Two Phase Mixture Level [in]

time (hr)

In-Shroud

Downcomer

RPV Water Level

Initial debris relocation 

into lower head

Main Steam Line Nozzle

Top of Active Fuel (TAF)

Bottom of Active Fuel (BAF)

Elevation of top of

jet pumps

-200

0

200

400

600

800

1000

1200

500

1000

1500

2000

2500

3000

0 1 2 3 4 5 6 7 8

In-vessel Hydrogen Generation [kg]

Temperature [K]

time [hr]

Cladding Temperatures

at Core Mid-plane

Ring 1

Ring 2

Ring 3

Ring 4

Ring 5

H2 generation

Indicates intact fuel no

longer present at this location

SRV seizes open
5-42

Debris that pours into the body of water in the lower head is cooled because of fragmentation as 

it travels through water and is then cooled by bulk boiling on surfaces of the resulting debris bed. 

This effect is shown in Figure 5-38 and Figure 5-39, which give the calculated mass and 

temperature43 of debris in the lower head, respectively. When residual water in the lower 

plenum is completely evaporated, debris temperatures begin to rise, exceeding the melting 

temperature of stainless steel (1700 K) in approximately 1.5 hours. Rising debris temperatures 

also cause the temperature of the lower head to increase, as indicated in Figure 5-40. Because

reactor vessel pressure is relatively low during this heat up, failure of the lower head is from 

creep rupture at high temperature.44

Figure 5-38 Mass of core debris in RPV lower plenum: STSBO without RCIC blackstart

 

43 The signatures shown in Figure 5-39 represent the temperature of debris along the vertical axis of the debris 

bed in the lower plenum. The axial nodalization of the lower plenum is shown in Figure 5-38. 44 The inner surface temperature of the central region of the lower head (i.e., MELCOR rings 1-4) is above 

the melting point of steel at the time failure occurs.

0

50

100

150

200

250

300

350

400

0 1 2 3 4 5 6 7 8 9 10 11 12

Mass (MT)

time [hr]

Mass of Debris 

in Lower Plenum

Lower head failure

Lower head

dryout

Lower core

plate failure
5-43

Figure 5-39 Temperature of debris in RPV lower plenum: STSBO without RCIC 

blackstart

Figure 5-40 Lower head temperature: STSBO without RCIC blackstart

(IR-inner radius; OR-outer radius)

0

500

1000

1500

2000

2500

0 1 2 3 4 5 6 7 8 9 10 11 12

Temperature [K]

time [hr]

Level 5

Level 4

Level 3

Level 2

Level 1

Lower core plate failure

Lower head failure

Lower head dryout

L.P. Debris Temperatures

Along RPV Centerline

Drainage of molten inconel & 

control rod mat'l poison

400

600

800

1000

1200

1400

1600

1800

2000

2200

0 1 2 3 4 5 6 7 8 9 10 11 12

Temperature [K]

time [hr]

Lower Head Temperatures

Ring 1 - IR

Ring 1 - OR

Ring 2 - IR

Ring 2 - OR

Ring 3 - IR

Ring 3 - OR

Ring 4 - IR

Ring 4 - OR

Ring 5 - IR

Ring 5 - OR

Lower

head 

failure

Lower head dryout
5-44

In contrast to the LTSBO (refer to Section 5.1.1), containment pressure immediately following 

reactor vessel failure is well below the threshold for induced leakage through the drywell head 

flange (80 psig). Therefore, leakage from containment does not occur by this mechanism. The 

lower pressure in the short-term scenario is the result of a reduced period of reactor steaming to 

the suppression pool before to the onset of core damage. Torus water temperature remains sub 

cooled, relative to atmospheric conditions, throughout the entire period of in-vessel core damage 

progression (see Figure 5-41). Therefore, steam contributions to containment pressure are 

negligible. Figure 5-42 shows the containment pressure for this short-term station blackout

scenario.

Containment conditions change dramatically, when debris is released onto the reactor pedestal 

floor following lower head failure. The absence of water on the drywell floor allows debris to 

flow laterally out of the cavity through the open personnel access doorway and spread out across 

the main drywell floor. Lateral movement and spreading of debris across the drywell floor allow 

debris to reach the steel shell at the outer perimeter of the drywell within 10 minutes. 

Five minutes later, thermal attack of the molten debris against the steel shell results in shell 

penetration and opening of a release pathway for fission products into the basement (i.e., torus 

room) of the reactor building. This results in a rapid depressurization of the containment to 

atmospheric conditions in a short time (i.e., less than 1 hour). 

Figure 5-41 Suppression Pool Temperature: STSBO without RCIC blackstart

Immediately following drywell shell melt-through, hydrogen is released from the drywell into 

the basement of the building (i.e., torus room) and is transported upward through open floor 

gratings into the ground level of the reactor building. Flammable mixtures quickly develop in 

300

325

350

375

400

425

450

0 2 4 6 8 10 12 14 16 18 20 22 24

Temperature [K]

time [hr]

Suppression Pool 

Temperature

Pool Temperature

Saturation Temperature

Containment failure

(drywell liner melt-through)
5-45

these regions, which are assumed to ignite from high flammable gas concentration and high gas 

effluent temperatures exceeding 1500 K near the floor of the drywell. The resulting pressure rise 

within the building causes several doorways within the building to open, including the large 

equipment access doorway at grade level and the blow-out panels in the side walls of the 

refueling bay near the top of the building. This combination of two large openings in the reactor 

building creates an efficient transport pathway for material released from containment to the 

environment. That is, a vertical column of airflow is created within the building, whereby fresh 

air from outside the building enters through the open equipment doors at grade level, rises 

upward through the open equipment hatches at every intermediate floor within the building, and 

exits through the blowout panels at the top of the building. As was described earlier 

(Section 5.1.2), retention of fission products in the reactor building is small for most species 

because of the chimney effect that is created by this flow pattern.

 

Figure 5-42 Containment Pressure: STSBO without RCIC blackstart

5.4.2 Radionuclide Release

The release of radionuclides to the environment occurs in two important phases. The first phase 

is very short and is characterized by the discharge of airborne fission products from the 

containment immediately following containment failure. A later and protracted release of begins 

24 hours into the sequence as a result of revaporization of CsI and tellurium initially deposited 

on surfaces within the RCS. Both phases are clearly evident in Figure 5-43, which shows the 

fractional release to the environment for all radioactive species. An expanded view of the release 

fractions for volatile species is shown in Figure 5-44.

0

10

20

30

40

50

60

70

80

0 2 4 6 8 10 12 14 16 18 20 22 24

Pressure [psig]

time [hr]

Drywell Pressure

DW liner melt-through

Start of clad

oxidation
5-46

Figure 5-45 through Figure 5-47 indicate changes in the spatial distribution of iodine, cesium and 

tellurium that result in the environmental releases shown in Figure 5-44. In particular, 

Figure 5-45 and Figure 5-47 clearly indicate the delayed, gradual release of iodine and tellurium, 

which begins 16 hours after containment failure,

45 is due to the movement of these species away 

from surfaces within the RCS, through the drywell to the environment. Although the transport 

pathway to the environment also involves the reactor building, retention within the building is 

negligible. Airborne fission products have a short residence time in the building due to hydrogen 

combustion immediately follows containment failure and a persistent buoyancy-driven flow 

through the building in the long term. This behavior was described earlier for the unmitigated 

LTSBO scenario (Section 5.1) and results in very low DFs for the reactor building, as illustrated 

in Figure 5-48.

Figure 5-43 STSBO without RCIC blackstart environmental source term

 

45 24 hours after the initiating event.

1.E-06

1.E-05

1.E-04

1.E-03

1.E-02

1.E-01

1.E+00

0 6 12 18 24 30 36 42 48

Fraction of Initial Core Inventory

time [hr]

Fission Products 

in Environment

NG

Ba

Ce

La

Ru

Mo

Containment failure

Cs

Cd

Sn

I Te
5-47

Figure 5-44 STSBO without RCIC blackstart environmental source term: details for 

volatile species

Figure 5-45 Spatial distribution of Iodine: STSBO without RCIC blackstart

1.E-03

1.E-02

1.E-01

1.E+00

0 6 12 18 24 30 36 42 48

Fraction of Initial Core Inventory

time [hr]

Fission Products 

in Environment

NG

I 

Cs

Ba

Te

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

1

0 6 12 18 24 30 36 42 48

Fraction of Initial Core Inventory

time [hr]

Iodine Distribution

Release to

environment 

Captured in 

Suppression Pool

Deposited/Airborne

within RPV & main

steam lines

Drywell 

(mostly airborne)
5-48

Figure 5-46 Spatial distribution of Cesium: STSBO without RCIC blackstart

Figure 5-47 Spatial distribution of Tellurium: STSBO without RCIC blackstart

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

0 6 12 18 24 30 36 42 48

Fraction of Initial Core Inventory

time [hr]

Cesium Distribution

Release to

environment (1.7%)

Captured in 

Suppression Pool

Deposited/Airborne

within RPV + main stm lines

Drywell negligible

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

0 6 12 18 24 30 36 42 48

Fraction of Initial Core Inventory

time [hr]

Tellurium Distribution

Release to

environment (10.4%)

Captured in 

Suppression Pool

Deposited/Airborne

within RPV + main stm lines

Drywell 
5-49

Figure 5-48 Reactor Building decontamination factor (DF): STSBO without RCIC 

blackstart

Figure 5-45 indicates that iodine that deposits on surfaces within the RCS during the first 6 to 

10 hours of damage progression is subsequently released from these surfaces to the environment. 

High atmosphere temperatures in the drywell prevent CsI from depositing on surfaces in the 

containment. It was noted earlier that a long-term stable flow of air through the reactor building 

prevents airborne CsI from depositing in the building. Nearly all material released from the RPV 

after 24 hours is, therefore, transported to the environment. The source of this late release of 

iodine and tellurium (shown in Figure 5-45 and Figure 5-47, respectively) within the RPV can be 

traced to the delayed evaporation of water from reactor recirculation loop piping.

Forced circulation flow through the reactor vessel is maintained by two recirculation loops 

external to the RPV (see Figure 4-2). The diagram shown in the lower right-hand corner of

Figure 5-49 shows the hydrodynamic nodalization of fluid in these loops, as represented in the 

MELCOR model shown in Figure 4-2. Water in recirculation loop piping only partially 

participates in depletion of reactor coolant (see Section 5.4.1) due to boiling in the core. 

Figure 5-49 shows the gradual evaporation of water in the recirculation loop. The level of water 

in the recirculation loops ‘risers’ (i.e., discharge piping between the recirculation pumps and top 

of the jet pumps) initially follows the decline in water level in the RPV downcomer (see 

Figure 5-36). Residual water in the downcomer (i.e., outside the jet pump diffusers) evaporates 

at approximately 6 hours; but approximately 4 meters of water is trapped in the bottom of the 

recirculation loop piping when RPV lower head failure occurs. This water in both recirculation

loops retains approximately 4.4 kg of CsI, or nearly 11% of the initial core inventory, 

representing over 10% of the initial core inventory of iodine. 

0

1

2

3

4

5

6

7

8

9

10

0 10 20 30 40

Fraction of Initial Core Inventory

time [hr]

Decontamination Factor

in Reactor Building

Cesium

Iodine

containment failure 

and open reactor building

early

containment

leakage to

an intact

reactor bldg

beginning of late 

revaporization release of CsI
5-50

It is also instructive to compare the amount of iodine and tellurium captured on in-vessel 

surfaces and in water in the recirculation loop piping in the STSBO to the amount that was 

indicated earlier for the LTSBO scenario. In-vessel retention of these species is larger in the 

STSBO than the LTSBO. At the time of RPV lower head failure, the amount of iodine retained 

in the RPV is approximately 13% for the STSBO (without RCIC blackstart) versus 3% for the 

LTSBO. This difference affects the extent to which late revaporization from the RCS affects the 

environmental release in the two scenarios. That is, more iodine and tellurium are available for 

late revaporization in the STSBO than in the LTSBO, and thus the late release from the 

recirculation loop piping observed in the STSBO has a measureable impact on the total release to 

the environment but does not impact the LTSBO. For completeness, it should be noted that total 

evaporation of water from the recirculation loop piping does not occur until long after 48 hours 

in the LTSBO scenario. However, as shown in Figure 5-12 and Figure 5-14 respectively, the

amount of volatile iodine and tellurium that is available for release (i.e., retained within the RCS) 

is not sufficiently large to significantly impact the environmental release even if it occurred 

earlier.

Figure 5-49 Water level and temperature of RPV recirculation loop-A piping: STSBO

without RCIC blackstart

After RPV lower head failure, this piping is exposed (optically) to molten core debris that 

spreads across the drywell floor. This causes the outer temperature of the outer surface of 

recirculation loop piping increase dramatically, as indicated in Figure 5-49. Conduction heat 

transfer through the recirculation loop piping evaporates residual water within the pipe, which 

dries out at approximately 22.5 hours.

MELCOR assumes that fission products retained in an evaporating pool of water remain in or on 

the surface of the pool until it fully evaporates. At that point in time, the material is transferred 

300

500

700

900

1100

1300

1500

-5

0

5

10

15

20

25

0 6 12 18 24 30 36 42 48

PIpe Wall Temperature [K]

Water Level wrt Bottom of RPV [m]

time [hr]

RPV Dwncmr Level

Suction Water Level

Discharge Water Level

Outer Surface Temp

Inner Surface Temp

Recirc 

Loop A

RPV lower 

head failure
5-51

to the surfaces of neighboring heat structures – in this case, heat structures representing the 

recirculation loop piping. Figure 5-50 shows the accumulation of CsI in water within 

recirculation loop piping during the early period of in-vessel damage progression and transfer of 

the solute onto the inner surface of the recirculation loop piping walls when the water is 

completely evaporated (at approximately 22 hours). The figure also shows the gradual 

evaporation of the deposited CsI aerosol away from the walls over the succeeding 5 hours due to 

the very high surface temperatures generated by heat transfer from neighboring core debris (pipe 

wall temperature is shown in Figure 5-49.) The total quantity of re-vaporized CsI 

(approximately 11% of the initial core inventory) corresponds directly to the amount released to 

the environment over the same time frame. The same relationship between retention and late 

revaporization from recirculation loop piping is observed with tellurium. Although similar 

fractional quantities of the dominant form of cesium (cesium molybdate) are captured in 

recirculation loop piping, sufficiently high pipe wall temperatures are not calculated within the 

48 hour time frame of these calculations to revaporize this material due to its chemical form with 

a higher vapor pressure.

Figure 5-50 CsI deposited on the surface of recirculation loop piping: STSBO without 

RCIC blackstart

5.5 SBO Sensitivity Cases with Alternate SRV Failure Criteria

The chronology of events for each of the SBO calculations described in previous sections 

includes failure of a cycling SRV. At some point in time, a continuously cycling SRV sticks in 

the open position either due to stochastic or thermally-induced failure mechanisms described in 

Section 4.4.2. The criterion for valve failure, and therefore the precise time at which valve 

0

0.02

0.04

0.06

0.08

0.1

0.12

0 6 12 18 24 30 36 42 48

Fraction of Initial Core Inventory 

time [hr]

CsI captured in water

CsI on Recirc Piping Walls

CsI inventory captured 

in residual water

Recirc piping dryout

(transfer of CsI in solution 

to pipe surfaces)
5-52

seizure would occur, is uncertain. Analyses of severe accident progression frequently identify 

RPV pressure as an important parameter affecting severe accident progression and fission 

product source term. Therefore, sensitivity calculations were conducted to quantitatively 

measure the impact of uncertainties in SRV failure criteria and discern how they affect calculated

source terms.

The calculations separately addressed uncertainties in stochastic and thermal failure criteria. 

Reasonable variations in these criteria were examined first, and results are described in 

Section 5.5.1. In this context, ‘reasonable variations’ means the alternate failure criteria are 

judged to fall within 1- or 2-sigma range of credible values for a normal distribution (i.e., they 

differ from the best estimate, but do not lie outside the 5th and 95th percentile values). The 

calculations were not explicitly designed to span this entire range, but provide a sufficient 

sampling to understand the extent to which the time of SRV seizure affects results.

As noted in the next section, all of the calculations performed within this range depressurized the 

RPV before creep rupture of the reactor coolant system pressure boundary occurred. Therefore, 

energetic containment loads associated with rapid RPV blowdown from high pressure did not 

result from any of the sensitivity calculations. Additional calculations were, therefore, 

performed to examine the manner in which creep rupture and attendant energetic loads on the 

containment pressure boundary would affect calculated source terms. Creep rupture in this 

context involves structural failure in one of two locations: reactor vessel lower head due to 

direct heating by molten core debris or main steam line (MSL) piping due to heat transfer from 

hot gases transported through cycling SRV(s). For creep rupture at either location to occur at 

high pressure, extreme values of SRV failure criteria must be defined. Section 5.5.2 describes 

results of sensitivity calculations that apply these extreme SRV failure conditions.

5.5.1 Reasonable variations in failure criteria

Alternate stochastic and thermal failure criteria for a cycling SRV were examined as follows:

• The best-estimate stochastic failure criterion of 270 cycles was changed downward as 

low as the 50th percentile (or 187 cycles) and upward to the 90th percentile (or 620 

cycles)

46.

 

• The best-estimate valve-stem temperature failure criterion of 900 K (1160 °F) was 

reduced to 811 K (1,000 °F) and as high as 1,100 K (1,500 °F). 

• The best-estimate position at which the valve is expected to fail is full open. The 

possibility of valve stem seizure at positions less than full open were examined by 

reducing the flow area after seizure by 50%.

 

46 620 cycles is never observed in the sensitivity calculations because the SRV seizes open due to thermal 

. failure criteria before this large number is reached. More discussion on this is provided later.
5-53

Also included in the comparison of results from these sensitivity variations to the best estimate 

results are preliminary calculations in which the SRV thermal failure criterion was defined on the 

basis of high gas temperatures, rather than the temperature of a simulated internal valve 

component. These results are included here primarily to document the effects of changes in 

calculation methods that resulted from the independent peer review process. The peer review 

panel recommended a more rigorous assessment of SRV failure criteria as part of their review of 

preliminary MELCOR results.

Changes in SRV failure criteria alter the time at which RPV blowdown begins, and therefore 

affect the RPV pressure and water level signatures. Sample results of a few sensitivity 

calculations are compared in the following figures for each version of the SBO scenario:

• LTSBO: Figure 5-51 and 

• Figure 5-52

• STSBO with RCIC Blackstart: Figure 5-53 and 

• Figure 5-54

• STSBO without RCIC Blackstart: Figure 5-55 and Figure 5-56

Changes in the failure criteria for a cycling SRV have a large impact on the transient pressure 

signature for all cases. Within a reasonable range of alternate failure criteria, the time SRV 

depressurization begins can shift by several hours. However, these changes in the RPV pressure 

signatures have a smaller impact on the coolant boiloff (water level) signature, and a very small 

effect on RPV failure time. The time at which the RPV water level decreases below TAF shifts 

by less than 30 minutes in the STSBO without RCIC blackstart and less than one hour in the

LTSBO. The impact on the STSBO with RCIC blackstart is more complicated. Late SRV 

failure generally occurs long after RCIC flow terminates by over-filling the RPV with water. 

However, early SRV failure can result in premature RCIC failure on low steam turbine inlet 

pressure. This difference in pressure response changes the amount of water in the RPV at the 

time RCIC flow terminates and, therefore, also impacts the coolant boiloff signature.
5-54

Figure 5-51 Comparison of RPV Pressure Response: SRV Failure Sensitivity – LTSBO

Figure 5-52 Comparison of RPV Level Response: SRV Failure Sensitivity - LTSBO

0

200

400

600

800

1000

1200

1400

0 2 4 6 8 10 12 14 16 18 20 22 24

Pressure (psia)

Time (hr)

SRV sticks at 187 cycles

SRV sticks at 270 cycles

SRV sticks at 1,100 K (497 cycles)

LTSBO - RPV Pressure

Batteries exhaust -

SRV recloses

Operator manually

opens 1 SRV SRV sticks open

Lower head

failure

Large scale debris

relocation into

lower head

0

100

200

300

400

500

600

700

800

0 2 4 6 8 10 12 14 16 18 20 22 24

Two Phase Mixture Level (in)

Time (hr)

SRV sticks at 187 cycles

SRV sticks at 270 cycles

SRV sticks at 1,100 K (497 cycles)

LTSBO - RPV Water Level

RCIC operation

Large scale debris

relocation into

lower head

MSL Nozzle

Top of Active Fuel (TAF)

Bottom of Active Fuel (BAF)

SRV sticks open
5-55

Figure 5-53 Comparison of RPV Pressure Response: SRV Failure Sensitivity – STSBO

with RCIC blackstart

Figure 5-54 Comparison of RPV Level Response: SRV Failure Sensitivity – STSBO with 

RCIC blackstart

0

200

400

600

800

1000

1200

1400

0 2 4 6 8 10 12 14 16 18 20 22 24

Pressure (psia)

Time (hr)

SRV sticks at 187 cycles (782 K)

SRV sticks at 900 K (615 cycles)

STSBO with RCIC blackstart -

RPV Pressure

Lower head failure

Beginning of large 

scale debris relocation 

into lower head

RCIC

operation

SRV sticks in 

open position 

0

100

200

300

400

500

600

700

800

0 2 4 6 8 10 12 14 16 18 20 22 24

Two Phase Mixture Level (in)

Time (hr)

SRV sticks at 187 cycles (782 K)

SRV sticks at 900 K (615 cycles)

STSBO with RCIC blackstart -

RPV Water Level

RCIC

blackstart

RCIC steam

line floods

Beginning of large scale 

debris relocation into

lower head

Lower head failure

RCIC isolates

on low steam

line pressure
5-56

Figure 5-55 Comparison of RPV Pressure Response: SRV Failure Sensitivity – STSBO 

without RCIC blackstart

Figure 5-56 Comparison of RPV Level Response: SRV Failure Sensitivity – STSBO 

without RCIC blackstart

0

200

400

600

800

1000

1200

1400

0 1 2 3 4 5 6 7 8 9 10 11 12

Pressure (psia)

Time (hr)

SRV sticks at 15 cycles

SRV sticks at 187 cycles

SRV sticks at 900 K (615 cycles)

STSBO without RCIC blackstart -

RPV Pressure

Lower head failure

Beginning of large 

scale debris relocation 

into lower head

SRV sticks in 

open position 

0

100

200

300

400

500

600

700

0 1 2 3 4 5 6 7 8 9 10 11 12

Two Phase Mixture Level (in)

Time (hr)

SRV sticks at 15 cycles

SRV sticks at 187 cycles

SRV sticks at 900 K (615 cycles)

STSBO without RCIC blackstart -

RPV Water Level

SRV sticks in 

open position 
5-57

Differences in the details of parameters that reflect the calculated progression of fuel damage, 

material relocation and debris temperature are observed among the SRV sensitivity cases 

examined here. For example, the time at which large scale debris relocation into the lower 

plenum begins from above the lower core plate is shown in Figure 5-51 through Figure 5-55 to 

span a relatively wide range (i.e., 2 to 4 hours, depending on the scenario). This is due, in part, 

to the relationship between steam flow through the debris bed that accumulates above the lower 

core plate and debris temperatures. If SRV seizure occurs early in time (i.e., before RPV water 

level reaches TAF), RPV depressurization is complete before large quantities of debris 

accumulate on the lower core plate. Consequently, debris temperatures are reduced by boiling 

heat transfer near the base of the core, and lower core plate failure occurs very soon after RPV 

water level decreases below the lower core plate. Conversely, if SRV seizure occurs late in time 

(i.e., after RPV water level is close to the bottom of active fuel, lower core plate failure is 

delayed by steam cooling of debris during RPV depressurization. Significant differences in the 

debris temperature signatures are, therefore, observed among these sensitivity cases. These 

differences, in turn, affect the calculated time between dryout of the lower head (i.e., total 

evaporation of water from the RPV) and lower head failure. The time between RPV dryout and 

lower head failure tends to be longer for cases involving early SRV seizure than those for later 

seizure. However, it is interesting to note that differences in the details of in-vessel debris 

temperature signatures and material relocation have a small effect on the calculated time of lower 

head failure. As indicated in Table 5-5, the calculated time of lower head failure does not vary 

significantly among the sensitivity cases for a particular scenario. The integral effects of 

differences in in-vessel damage behavior appears to have a small effect on the time at which 

lower head failure occurs.

Table 5-5 compares calculated values of several key metrics of the sensitivity calculations to the 

best estimate results for all three versions of the SBO sequence described in previous section. 

The listed metrics are:

• SRV failure mode: This indicates the failure criterion that occurred first in each 

calculation. ‘S’ indicates stochastic failure; ‘T’ thermal failure

• Cycles at seizure: This indicates the total number of cycles experienced by the SRV at 

the time failure occurs.

• Time of SRV seizure: This is time into the accident (after the initiating event).

• Max MSL L-M Damage Index: The maximum value of the Larson-Miller creep damage 

index calculated for the heat structure representing the SRV valve stem. A value of unity 

indicates sufficient creep for failure.

• RPV lower head failure: This is the time at which lower head failure is calculated.

• Iodine release to Env > 1%: The time at which the cumulative release of iodine to the 

environment exceeds 1% of the initial core inventory.
5-58

• Fraction of Inventory in Environment at 48 hours: The cumulative release fractions of 

iodine, cesium and barium to the environment throughout the 48 hour duration of the 

calculations.

Two important conclusions can be drawn from the results outlined in this table. First and 

foremost, however, is that reasonable variations in the criteria for SRV failure lead to relatively 

minor changes in the characteristics of the environmental source term. Over the range of SRV 

failure criteria considered here, the start time at which fission products are released to the 

environment varies from the best estimate value by only 2 hours. With only one exception, 

changes in the cumulative magnitude of fission product release to the environment are negligibly 

small.

47 This observation enhances confidence that the best estimate result is robust against 

uncertainties in the SRV failure criteria.

The second conclusion is that the margin from MSL creep rupture is large. The only condition in 

which the calculated value of the Larson-Miller (L-M) creep damage index for the MSL piping is 

close to unity is one that was judged by the SOARCA Peer Review panel to be unreasonable 

(i.e., SRV seizure occurs when the valve experiences 10 cycles with internal gas temperatures 

above 1000 K). As noted in Section 4.4, this modeling approach was replaced by one that 

explicitly calculated the thermal response of an SRV internal component. A large margin from 

MSL creep rupture is observed in all cases using the refined model. This margin can be 

expressed in at least two ways. One is the low (maximum) values of the L-M damage index for 

the MSL pipe wall. The second is the very large number of cycles that the valve experiences if 

stochastic failure is effectively neglected (i.e., if failure occurs only if the valve experiences over 

500 cycles, or twice the expected value for failure).

 

47 The single exception is the case with very early SRV seizure (15 cycles) in the STSBO, in which case the 

total environmental release fractions are much smaller than other cases.
5-59

Table 5-5 Key Metrics of the Impact of Reasonable Variations in SRV Failure Criteria
5-60

5.5.2 Extreme variations in failure criteria 

If a cycling SRV does not stick in the open position and in-vessel damage progression continue

at high pressure, the potential for high pressure creep failure of the RCS becomes a concern. As 

noted earlier, high pressure creep rupture can either occur in the RPV lower head due to direct 

heat transfer from molten core debris or by overheating of main steam line piping as hot gases 

flow to the cycling SRV. Either mechanism of high pressure failure of the RCS has the potential 

of imparting significantly higher thermodynamic loads on the containment pressure boundary 

than those observed in calculations described in earlier sections of this report. High containment 

loads could result from high pressure creep rupture of the RCS because rapid discharge of high 

temperature, non-condensable gases to the containment would not be efficiently mitigated by the 

suppression pool in the wetwell. 

To investigate this issue, sensitivity calculations was performed to determine the amount by 

which best estimate SRV failure criteria must be exaggerated for high pressure creep rupture of 

the RCS pressure boundary to occur. In particular, stochastic SRV failure occurs with 90% 

confidence if the valve cycles 615 times. The sensitivity calculations described in the previous 

section suggest thermal failure mechanisms occur well before this number of cycles would occur. 

In addition, thermal seizure of an SRV was assumed to result with one of the following two 

extreme conditions:

1. SRV cycling would continue until valve stem temperatures reached a value greater than 

1100 K – i.e., greater than the maximum value considered in the sensitivity calculations 

described in the previous section, or

2. The valve stem seizes in a position that is sufficiently large to prevent further increases in 

RPV pressure (and therefore preclude pressure relief by another available SRV), but also 

sufficiently small to maintain pressure above values necessary for MSL creep rupture to 

occur.

Trial-and-error determined that valve stem temperatures greater than 1175 K satisfied 

condition-1 for the LTSBO48 sequence, and valve open fractions of 10% of the nominal flow 

area satisfied condition-2 for the LTSBO and STSBO sequences.

Table 5-6 summarizes the same metrics listed in Table 5-5, but compares results of the best 

estimate calculations for LTSBO and STSBO to the extreme situations needed to result in MSL 

creep rupture. MSL creep rupture was calculated to occur well before lower head failure in each 

case, so MSL creep rupture is the only high pressure failure mechanism examined here. The 

comparisons shown in Table 5-5 suggest that if a MSL rupture were to occur, the environmental 

source terms would be more severe in two respects. First, significant fission product release to 

environment begins in nearly one-half the time observed in the best-estimate calculations. Iodine 

release to the environment exceed one percent of the initial core inventory in approximately 

24 hours in the best estimate LTSBO calculation, but is reduced to approximately 13 hours in the 

 

48 The STSBO calculation with an assumed SRV failure temperature of 1175 K did not result in MSL creep 

rupture, as indicated in Table 5-6. Temperatures greater than 1250 K were necessary for the STSBO, 

which is not considered credible. 
5-61

cases with MSL creep rupture. The same metric moves forward from nearly 10 hours in the best 

estimate STSBO calculation to 5.5 hours in the case with MSL creep rupture. This time occurs 

well before RPV lower head failure in the MSL creep rupture cases for both sequences. Fission 

product release to the environment begins, in such cases, immediately following MSL creep 

rupture because the calculated containment pressure is much larger than the failure criterion 

described in Section 4.6 for drywell head flange leakage. The second way in which the source 

terms are more severe for the creep rupture cases is the magnitude of the release. The 48-hour 

release fractions for iodine and cesium are substantially larger than the best estimate values. 

It should be noted that the calculations involving creep rupture of an MSL assumed the structural 

response of the MSL is a fully offset, guillotine break of one MSL. This is considered a 

conservative assumption because the large break area maximizes the hydrodynamic load to the 

containment pressure boundary and facilitates fission product transport from the RCS. 

Alternative credible responses, such as a smaller crack or fissure in the MSL or rupture at an 

alternative location (i.e., the tailpipe of the open SRV) were not considered in these sensitivity 

calculations.
5-62

Table 5-6 The Effects of Extreme Variations in SRV Failure Criteria on Key Severe 

Accident Metrics
5-63

5.6 Loss of Vital AC Bus E-12 – Sensitivity Cases without B.5.b Equipment

This scenario’s event progression (Table 5-7) assumes that operators follow the actions directed 

in Trip Procedure T-101 [4] (RPV control). The estimates of the time at which actions would be 

taken are based on a table-top exercise with plant operations personnel during a site visit in 

June 2007. The plant operations personnel would take approximately 15 minutes to assess the 

situation before taking action. During this time, the RCIC system would actuate and cycle as 

needed to maintain level. Additionally, one or more SRVs would cycle to control RPV pressure. 

These systems are assumed to operate automatically based on nominal actuation and 

termination/closure set points; manual intervention is assumed unnecessary and is not credited 

within this initial time period. It should be noted that reactor and equipment operators train on

taking manual control of RCIC and would likely do so earlier than is assumed in this analysis.

The loss of vital electric power (i.e., the initiating event) results in reactor scram, closure of the 

MSIV, and containment isolation. In response to this scenario, the operator would be directed, in 

part, by Trip Procedure T-101. This procedure provides instructions for managing reactor 

power, water level, and pressure. 

Within Trip Procedure T-101, a specific step directs the operators to restore and maintain level 

between +5 and +35 inches (+177 to +206 inches above TAF) using the following, in order of 

priority:

• feedwater

• CRDHS

• RCIC

• HPCI

• condensate

• core spray

• low pressure coolant injection

Feedwater is not available in this scenario. However, the CRDHS continues to operate at its 

nominal (post-scram) flow rate without operator intervention.49 RCIC is also available and 

would start automatically upon receipt of a ‘low-level 2’ signal (i.e., level less than -66 inches

relative to instrument zero).

50 Long-term operation of these two systems alone provides 

sufficient makeup to maintain level. However, RCIC flow terminates in 4 hours when power 

 

49 The procedure directs operators to maximize the CRDHS flow using procedure T-246, but the second 

pump is not available in this scenario. The only action that can be taken to increase flow is to open the 

pump discharge throttle valve, which is assumed to occur 1 hour after the initiating event. The licensee 

estimates this would increase maximum flow (at full pressure) from 110 gpm to 140 gpm. 50 If the MELCOR reference point is adjusted to use TAF as the reference point, Low-Level-2 corresponds to 

+106 inches above TAF. This value does not match the value of +124 inches provided by the licensee in 

response to a ‘fact check’ of this report requested by the NRC. Correcting this discrepancy would not 

change the fundamental hydraulic response of the MELCOR model to the Bus E-12 sequence or the SBO 

sequences because actuation of RCIC fully-restores RPV water level in a very short time. If the time of 

RCIC actuation is delayed slightly (i.e., based on the 18 inch difference in the modeled versus actual value 

of ‘low-level-2’ signal) the minimum value of RPV water level would change, but full level recovery 

would be nonetheless be calculated.
5-64

from station batteries depletes. Parametric MELCOR calculations (described in Section 5.6.3) 

indicate that battery durations (i.e., RCIC operation) greater than 3 hours are sufficient for 

long-term operation because after this time, the CRDHS alone can prevent core damage. 

Therefore, best-estimate MELCOR analysis concludes that this scenario would not lead to core 

damage as (conservatively) identified in the NRC SPAR model. Additional details of the 

calculation are provided below.

Table 5-7 Timing of Key Events for Loss of Vital AC Bus E-12

Event Description Time

(hr)

Loss of vital AC bus E-12 0.0

MSIV closure, reactor scram and containment isolation 0.0+

RCIC automatically starts because of low Rx water level 0.2

Operators begin manual depressurization (open 1 SRV) 1.5

Operators take manual control of RCIC to maintain level within range 2.0

Station battery supply exhausted, SRV recloses and RCIC operation terminates* 4.0

Operators secure the single CRDHS pump to prevent reactor overfill 4.3

Reactor pressure back to SRV relief setpoint (SRV automatically cycles) 6.0

Operator restarts single control rod drive hydraulic system pump to restore level 7.0

Cycling SRV fails to reclose after several hundred cycles; reactor depressurizes 13.5

Reactor water level briefly decreases below top of active fuel 13.8

Level restored above top of active fuel 16.0

Level fully recovered to nominal (sequence terminated) 21.0

* As noted in the text, termination of RCIC system operation immediately following a loss of DC power is a conservative 

assumption when the RPV is at low pressure. At low RPV pressure, it is more probable that the system would continue 

operating until reactor vessel pressure increased above approximately 400 psig. 

5.6.1 Thermal Hydraulic Response

A specific step of Trip Procedure T-101 directs operators to take manual control of the SRVs if 

they are cycling (i.e., opening automatically at their lift setpoint). Another step further directs 

operators to open the SRV until the RPV pressure decreases below 950 psig. A quantitative 

target for the RPV pressure is not prescribed in the procedure, particularly if reduced pressure 

would not challenge the viability of coolant injection.51 Preliminary MELCOR calculations 

indicated that the lead SRV would cycle approximately 50 times before automatic actuation of 

the RCIC system at approximately 20 minutes into the event. Cycling would temporarily cease 

and the RPV would briefly depressurize while the RCIC system operated because of the steam 

flow to the RCIC turbine. RCIC operation automatically and temporarily terminates at 

45 minutes when reactor water level is restored to the high-level setpoint. Subsequently, RPV 

pressure increases and SRV cycling resumes. 

Based on this information, operators are assumed to initiate manual depressurization of the 

reactor vessel 1.5 hours into the event to prevent further cycling of the SRVs. As shown in 

 

51 It is important to note that manual SRV control requires DC power. Therefore, manual opening of a SRV 

is viable only while station batteries remain active or with a B.5.b portable power supply.
5-65

Figure 5-57, reactor vessel pressure decreases below 200 psig in approximately 1 hour. Reactor 

vessel pressure stabilizes near 150 psig until 4 hours, when DC power is lost (i.e., station 

batteries exhaust), and the open SRV recloses. Reactor vessel pressure increases back to 

the minimum SRV setpoint in 2 hours. For the next 7.5 hours, reactor vessel pressure is 

maintained at approximately 1100 psig by continuous cycling of the lowest setpoint SRV. 

Figure 5-57 Loss of vital AC bus E-12 reactor vessel pressure

At 13.5 hours, after several hundred cycles, the SRV fails to reclose and the reactor vessel again 

depressurizes. This event represents a random failure of the SRV to reclose, which is calculated 

by MELCOR based on the number of valve cycles, a failure rate of 3.7x10-3 per demand, and a 

90 percent confidence level for failure.

Figure 5-58 shows the calculated reactor water level during the entire 24 hour calculation. Water 

level initially decreases in response to reactor isolation and termination of reactor feedwater. 

Twelve minutes later, the RCIC system automatically starts and begins to refill the reactor 

vessel. RCIC flow is automatically terminated at 48 minutes when the water level reaches the 

high-level setpoint. The water level subsequently decreases slowly because of evaporation 

resulting from decay heat in the core. 

At 1.5 hours, when operators open an SRV to depressurize the reactor, the increased coolant 

discharge rate through the open SRV accelerates the rate at which reactor water level decreases. 

The RCIC system automatically starts a second time, briefly stabilizing water level near the 

low-level setpoint. Two hours into the event, operators take manual control of RCIC turbine 

speed to reduce injection flow as needed to maintain the water level within range. When the

0

200

400

600

800

1000

1200

0 2 4 6 8 10 12 14 16 18 20 22 24

time [hr]

Pressure [psig]

station batteries exhaust

(loss of DC power; 

 S/RV recloses)

Cycling S/RV 

sticks open Start controlled 

depressurization
5-66

depressurization transient is completed at approximately 2 hours, the coolant effluent rate 

through the open SRV is reduced, and reactor water level increases back to the target range. 

Operators take manual control of RCIC flow during this transient and subsequently maintain the 

water level within range for approximately 1.5 hours.

Figure 5-58 Loss of vital AC bus E-12 reactor water level

Four hours into the sequence, DC power from station batteries is exhausted. As noted earlier, 

this causes the open SRV to reclose, but it is also assumed to result in termination of RCIC 

system operation. Loss of DC power would cause the RCIC turbine inlet throttle valve to move 

to a full-open position. If RPV pressure is above approximately 400 psig, the additional steam 

flow resulting from the change in steam throttle valve position would result in an automatic trip 

of the system due to steam turbine overspeed. At the time of battery exhaustion in this scenario, 

however, the steam inlet valve would already be positioned at or near its full-open position to 

permit adequate steam flow into the RCIC turbine at low RPV pressure. As a result, steam flow 

to the RCIC turbine would not increase enough to cause an immediate system trip due to turbine 

over-speed. However, as RPV pressure increases following closure of the SRV, steam flow 

would increase and an overspeed trip would eventually occur. It is conservatively assumed in 

this analysis that the overspeed trip occurs immediately following station battery exhaustion. 

Between 4 and 6 hours, the reactor vessel pressure slowly increases as shown in Figure 5-57. 

During this period, coolant losses cease, but CRDHS flow continues. This causes the reactor 

vessel water level to increase above the upper limit of the target range specified in emergency 

procedures. Observing this trend, operations personnel are assumed (based on training) to 

manually secure the CRDHS pump to slow or terminate the increase in reactor water level with 

the objective of preventing water from spilling over into the main steam lines. As indicated in 

0

100

200

300

400

500

600

700

800

0 2 4 6 8 10 12 14 16 18 20 22 24

time (hr)

Two-Phase Level [in]

termination of dc power

(S/RV recloses)

Operator takes manual control of RCIC 

& opens CRDHS throttle valve

Start controlled 

depressurization

CRDHS off

TAF

BAF

Main Steam Line Nozzles

Cycling S/RV

sticks open

termination of dc power

(S/RV recloses)
5-67

Figure 5-58, the MELCOR calculation indicates that this objective would not be met because of 

expansion of the RCS volume as energy is gradually absorbed by the isolated RCS inventory. 

The steam lines begin to flood with water approximately one hour after DC power terminates 

and the open SRV recloses.

Approximately 1 hour after steam line flooding begins, reactor vessel pressure reaches the relief 

set point of the SRVs, and cycling with its associated discharge of reactor coolant begins anew. 

The discharge of coolant through the cycling SRV and the absence of any form of active coolant 

injection (i.e., the CRDHS remains inactive) cause the reactor vessel water level to decrease, 

eventually reducing below the elevation of the main steam lines and approaching the nominal 

range. When the level reaches the upper end of the target range (7 hours), the single available 

CRDHS pump is restarted to compensate for coolant inventory lost through the cycling SRV.

From this point forward in the scenario, the CRDHS operates continuously as the only resource 

of coolant makeup to the reactor vessel. The coolant delivery rate is approximately 140 gpm 

(i.e., maximum system flow) while the reactor vessel remains at full pressure. However, 

following reactor vessel depressurization at 13.5 hours because of SRV failure to reclose, the 

coolant flow rate increases to over 180 gpm, thereby allowing the reactor water level to increase 

back to the desired range.

The minimum water level observed in the core is well above the minimum steam cooling water 

level and fuel heat up, and damage is averted. The reason that this sequence was identified in the 

SPAR model as a ‘core damage’ sequence, but the results computed here conclude otherwise, is 

that the CRDHS system is not credited in the SPAR model. Operation of this system has a 

significant impact on plant response to the initiating event and a realistic examination of reactor 

hydraulic behavior leads to a conclusion of ‘no core damage.’

5.6.2 Radionuclide Release

Because core damage is averted in this scenario, a release of radionuclides from fuel does not 

occur and no environmental source term is generated.

5.6.3 Sensitivity Analysis

The analysts performed several sensitivity calculations to examine the effects of alternative 

assumptions about key features of system performance. Several sensitivity calculations were 

performed to confirm the conclusion that adequate core cooling would be maintained in this 

scenario if alternative credible assumptions were made about key features of system 

performance. This section describes the results of these calculations. In particular, the 

sensitivity calculations studied following uncertainties:

• Duration of station batteries (DC power): As noted in Section 3.3.2, the actual duration 

of DC power from station batteries depends on several factors, including battery age and 

the effectiveness of actions taken by plant personnel to shed nonessential loads from 

the DC bus. The minimum duration required by plant technical specifications (TS) is 

2 hours. However, durations longer than the 4 hour estimate are also possible. 

Therefore, sensitivity calculations were performed to evaluate the minimum battery 

duration needed to ensure adequate core cooling in this scenario.
5-68

• CRDHS coolant delivery rate: The baseline calculation described in Section 5.6.1

assumes that operators increase the flow rate from the single available CRDHS pump to 

its maximum capacity. This involves manual actions to open a locked throttle valve in 

the pump discharge line. Sensitivity calculations were performed to evaluate plant 

response if this action is not taken.

• Manual depressurization: Plant emergency procedures call for manual depressurization 

of the reactor vessel, which is assumed to occur 1.5 hours after the initiating event in the 

baseline calculation. This action has two competing effects on hydraulic behavior in the 

RPV. Reducing reactor vessel pressure increases the coolant delivery rate from the 

CRDHS pump, but it also increases the rate at which coolant is discharged from the 

vessel during the blowdown period. The importance of this action is examined in a single 

bounding (i.e., worst case) sensitivity calculation, which assumes that manual 

depressurization does not occur and that operators fail to open the CRDHS throttle valve 

to permit maximum flow.

The results of the sensitivity calculations examining alternative values of station battery duration 

are summarized in and Figure 5-59 and Figure 5-60. The calculations considered four distinct 

values of assumed station battery duration: 2, 3, 4, and 6 hours. These calculations differ 

slightly from the baseline analysis in two ways. First, the RCIC system is assumed to operate 

entirely as an automatic system; operator actions to take control of the system to maintain water 

level within range are not credited. Second, manual actions to maximize CRDHS flow are not 

credited. The single available CRDHS pump is assumed to operate according to nominal system 

flow rates.

The calculated reactor water level for these sensitivity cases is shown in Figure 5-59 and the 

calculated peak cladding temperature in the core is shown in Figure 5-60. The minimum reactor 

water level is shown to dip below core mid-plane in the case with 2 hour battery duration, with 

gradually higher values for longer durations. The bounding thermal response of fuel in the core 

(i.e., maximum cladding temperature) is also improved with increasing battery duration. Fuel 

cladding failure and accompanying release of the gap inventory of radionuclides occurs in cases 

with 3 hour battery duration or less. Peak temperatures in cases with battery duration greater 

than 3 hours are below values at which clad failure would be anticipated (~1200 K).

Results of sensitivity calculations for the other parameters noted above are listed in Table 5-8. 

Neglecting the beneficial effects of operator actions to control CRDHS flow rate (i.e., maximize 

flow or secure the system at high reactor water levels) does not alter the conclusion of no core 

damage. The effects of manual depressurization, on the other hand, are potentially important. If 

operators fail to maximize CRDHS flow and fail to reduce reactor vessel pressure, core damage 

would not be avoided even if station batteries sustain RCIC flow for as much as 6 hours. 

However, in this case the damaged core is reflooded and vessel failure would not occur.
5-69

Figure 5-59 Sensitivity of station battery duration: Reactor water level- loss of vital AC

bus E-12

Figure 5-60 Sensitivity of station battery duration: Peak clad temperature- loss of vital 

AC bus E-12

0

100

200

300

400

500

600

700

800

0 2 4 6 8 10 12 14 16 18 20 22 24

time (hr)

Two-Phase Mixture Level [in]

2-hr dc

3-hr 4-hr

6-hr

Main Steam Line Nozzles

TAF

BAF

* RCIC operates via automatic cycling

 (no manual intervention)

* Variations in station battery duration

0

500

1000

1500

2000

2500

0 2 4 6 8 10 12 14 16 18 20 22 24

time (hr)

Maximum Clad Temp in Core [K]

2-hr dc

3-hr

4-hr

6-hr

* RCIC operates via automatic cycling

 (no manual intervention)

* Variations in station battery duration
5-70

Table 5-8 Sensitivities for Loss of Vital AC Bus E-12

Sensitivity RCIC 

Duration

Maximize 

CRDHS Flow 

(Mitigative 

Action)

CRDHS Off 

to Prevent 

RPV Overfill

Depressurize 

(Open SRV)

Re-pressurize 

(SRV Closes) Results*

Base Case 4 hrs 1 hr 4.3 – 7 hrs. 1.5 hrs 4 hrs No CD

CRD Flow 4 hrs Not Done 4.3 – 7 hrs. 1.5 hrs 4 hrs No CD

CRD Flow 

Battery Life

4 hrs Not Done Not Done 1.5 hrs 4 hrs No CD

Several cases 

spanning the 

range of 

2-6 hrs

Not Done Not Done 1.5 hrs 2-6 hrs > 3 hr life 

averts CD

Depressurize ≥3 hrs Not Done Not Done Not Done N/A CD, no VF

* Where, CD = core damage. VF = vessel failure.

5.7 Peer Review

The SOARCA peer review panel identified a few issues that motivated sensitivity calculations to 

investigate their effects. Results of many of these calculations have already been discussed in 

earlier sections of this report. Most notable is the sensitivity of core damage progression and 

fission product source terms to SRV behavior after the onset of core damage (Section 5.5). 

Sensitivity calculations were performed to examine the importance of three additional issues and 

are described below.

1. Containment leakage before failure

2. Atmospheric mixing in the drywell.

3. Leakage through TIP guide tubes

In all cases, an early (preliminary) version of the LTSBO calculation52 was used as the basis for 

investigating these issues. The following sections summarize the results of the sensitivity 

calculations for each issue.

5.7.1 Containment Leakage before Failure

The initiating event for the station blackout accident scenarios (i.e., both long- and short-term) is 

assumed to be a large beyond design-basis seismic event. The baseline calculations assume that 

containment leakage is limited to the maximum allowable by plant-specific TS. This assumption 

represents more leakage than would be anticipated during normal operation, because routine 

testing of containment leak tightness strives to control leakage well below the TS limit. 

However, this assumption does not directly account for the possibility that leakage greater than 

the TS limit could be caused by a large seismic event.

Two sensitivity calculations were performed to examine the impact of seismically induced 

increased leakage on radiological release to the environment. One case assumed a leak area 

 

52 The ‘base case’ calculation described in the following two sections differs slightly from the best estimate 

calculation of the LTSBO scenario described in Section 5.1. The difference lies solely in the assumed 

conditions for SRV failure. As described in Section 5.5, small differences in SRV failure criteria have a 

negligible effect on calculated source terms.
5-71

three times larger than the TS limit;

53 a second case assumed an area 10 times the TS limit. 

Increased levels of containment leakage have a very small impact on the final environmental 

source term, as shown in Figure 5-61. Prior to containment failure (i.e., zero to 19.5 hours in the 

LTSBO scenario), increased leakage has a small, but noticeable impact on the release of volatile 

iodine and cesium to the environment. However, the magnitude of release during this period of 

leakage is negligibly small in comparison to releases that occur shortly after containment failure. 

Increased levels of containment leakage do not affect the long term release of iodine or cesium 

because the vast majority of volatile species are swept to the suppression pool through the open 

SRV. Section 5.1.2 discusses this topic in more detail. Increases in containment leakage also 

have a negligible effect on the long-term release of iodine and cesium to the environment. 

Figure 5-61 Effect of increased containment leakage on the release of iodine to 

the environment

5.7.2 Atmospheric Mixing in the Drywell

The hydrodynamic model of the containment cannot explicitly calculate buoyancy-driven natural 

circulation flow patterns. The spatial representation of the drywell, in particular, is a simple 

series of three vertically-stacked control volumes, each connected by a single flow path. The 

possibility of rising hot gases and descending cooler gases within this region of the containment 

 

53 Containment leakage is modeled as a constant area opening in the containment pressure boundary. The 

actual leak rate, therefore, varies with internal pressure. Leakage corresponding to the TS limit is based on 

the area that would produce a leak rate of 0.5 percent of the containment free volume per day at an internal 

pressure of 56 psig. The leakage is assumed to be located in the drywell, where the largest number of 

penetrations through the containment pressure boundary is located.

Iodine

in Environment

1.E-10

1.E-09

1.E-08

1.E-07

1.E-06

1.E-05

1.E-04

1.E-03

1.E-02

1.E-01

1.E+00

0 12 24 36 48

time [hr]

Fraction of Initial Core Inventory

Base Case

3X DW Leakage

10X DW Leakage

Cesium 

in Environment

1.E-10

1.E-09

1.E-08

1.E-07

1.E-06

1.E-05

1.E-04

1.E-03

1.E-02

1.E-01

1.E+00

0 12 24 36 48

time [hr]

Fraction of Initial Core Inventory

Base Case

10X DW Leakage

3X DW Leakage
5-72

cannot be rigorously represented in the current MELCOR modeling framework. The absence of 

natural circulation in the MELCOR calculations results in an inverted, stratified temperature 

profile within the drywell after RPV lower head failure as indicated in Figure 5-62. This 

configuration is possible if the buoyancy-driven natural circulation within the drywell is 

inhibited by flow resistance associated with piping and other equipment within the drywell.

54 

Further, depressurization of the containment through the opening created in the drywell shell 

near the elevation of the drywell floor (i.e., the base of region D in Figure 5-62) would 

preferentially discharge a high-temperature atmosphere near the bottom of the drywell, perhaps 

supporting a high-temperature region near the drywell floor.

On the other hand, the vertical temperature differences between the top and bottom of the 

drywell reflected in Figure 5-62 represents a very strong driving force for upward flow. If a 

return (downward) flow path for cooler air from the top of the drywell is established, the 

resulting flow pattern would enhance mixing of the drywell atmosphere before being discharged 

to the reactor building. Analytical models for properly calculating buoyancy-driven natural 

circulation flow patterns with a large volume, such as a BWR Mark I drywell, are not available 

within MELCOR. However, the effects of atmospheric mixing can be examined by imposing a 

flow path configuration in the MELCOR model that encourages flow among the three verticallystacked control volumes representing the main body of the drywell atmosphere. This 

configuration involves replacing the single flow path connecting adjacent control volumes with 

two parallel flow paths and defining the endpoints of parallel flow paths to be slightly 

asymmetric. 

This adjustment to the baseline modeling approach was made in a sensitivity calculation. The 

result was a nearly continuous circulation velocity within the drywell of approximately 

0.5 meters per second and, as shown in Figure 5-63, a merging of atmospheric temperatures in 

the control volumes representing the main body of the drywell. The atmospheric temperature 

within the reactor pedestal (i.e., region D in Figure 5-63) is not significantly affected by this 

modeling adjustment. However, the impact of drywell atmospheric mixing on the radiological 

source term to the environment is small as indicated in Figure 5-64. The final release fractions 

of iodine are nearly the same in the baseline and sensitivity calculation. The long-term cesium 

release fraction increases by a small amount in the sensitivity calculation. 

 

54 Unlike the idealized image shown on the right-hand side of Figure 5-62, a large fraction of the internal free 

volume and horizontal cross-sectional area of the drywell is displaced by piping, valves, electrical control 

and instrument cabling and several layers of work platform grating.
5-73

Figure 5-62 Drywell atmospheric temperatures in the baseline LTSBO calculation

Figure 5-63 Drywell atmospheric temperatures with imposed drywell circulation

Drywell Atmosphere Temperatures

300

500

700

900

1100

1300

1500

1700

0 6 12 18 24 30 36 42 48

time [hr]

Temperature [K]

Lower head

failure

A

B

C

D

A

B

C

D

300

500

700

900

1100

1300

1500

1700

1900

0 6 12 18 24 30 36 42 48

time [hr]

Temperature [K]

Lower head

failure

A

B

C

D

A

B

C

D
5-74

Figure 5-64 Effect of modeling circulation flow within the drywell on iodine and cesium 

release to the environment

5.7.3 Leakage through TIP guide tubes

The description of the Peach Bottom MELCOR model in Section 4.2.4 mentioned that the TIP 

system operates for approximately one hour every four months. As a result, the baseline 

SOARCA calculations assume this system is isolated at the time the postulated accident occurs. 

However, because the TIP guide tubes offer a pathway for hydrogen and fission products to 

bypass the containment pressure boundary, a sensitivity calculation was performed to examine 

the potential consequences of leakage through the system, if it was in operation. The sensitivity 

calculation applied the additional features to the Peach Bottom MELCOR model described in 

Section 4.2.4. In particular, a hydrodynamic control volume and associated flow paths and heat 

structures were added to represent the internal volume, cross-sectional (flow) area and internal 

surfaces of the TIP guide tubes. The sensitivity calculation examines a bounding case, in which 

all three guide tubes are assumed to be active (i.e., the isolation valves are open) and all three 

probes are fully withdrawn into their respective shield chambers. This represents a physical 

configuration with maximum plausible leakage through the guide tubes.

Flow through the guide tubes begins when the instrument tubes in the core have melted, 

exposing the internal volume of the tubes to RPV pressure. The MELCOR model represents incore instrumentation tubes as miscellaneous steel structure within the core, which typically 

begins to melt when local clad oxidation reaches levels that challenge the physical integrity of 

fuel. Therefore, the MELCOR flow path from the RPV to the TIP guide tubes opens when 

damage to fuel in the center of the core exceeds the failure criterion for erect fuel (i.e., fuel pins 

collapse into particulate debris.) The actual area through which material would be released into 

the reactor building is difficult to define, as it is characterized by gaps and mechanical clearances 

in moving components within the probe shield chamber and drive mechanisms, which can 

Iodine

in Environment

1.E-04

1.E-03

1.E-02

1.E-01

1.E+00

0 12 24 36 48

time [hr]

Fraction of Initial Core Inventory

Base Case

with DW recirc

Cesium 

in Environment

1.E-04

1.E-03

1.E-02

1.E-01

1.E+00

0 12 24 36 48

time [hr]

Fraction of Initial Core Inventory

Base Case

with DW recirc
5-75

deteriorate if subjected to very high temperature gases (as would be expected in this case). The 

sensitivity calculation assumed the leak area in the reactor building is equivalent to the cross 

sectional area of the tubes. Or viewed another way, the leak rate into the reactor building is 

limited by flow resistance in the tube itself, not by the open area at the terminus of the guide 

tube. It should also be noted that the MELCOR model accounted for diversion of flow from the 

TIP guide tubes into the drywell through relief valves (or blowout valves) located in the indexing 

units. The valves are designed to open at a small differential pressure (a few psi), to protect the 

tubes and indexing units from over-pressure by the nitrogen purge system. These valves will, 

therefore, open when the tube is exposed to RPV pressure, allowing hydrogen and fission 

products to leak into the containment as well as the reactor building.

The sensitivity calculation is a permutation of the unmitigated LTSBO accident scenario 

described in Section 5.1. With the single exception of the modeling features described above to 

represent leakage through open TIP guide tubes, the MELCOR model is identical to the one used 

for the baseline unmitigated LTSBO calculation.

The calculated thermal-hydraulic response of the RPV and containment are not affected by the 

small amount of steam and hydrogen that leaks through the open TIP guide tubes. The 

signatures for RPV pressure and water level, containment pressure and temperature, as well as 

the in-vessel core damage parameters are essentially identical to those described in Section 5.1. 

Although hydrogen and fission products leak through the relief valves in the TIP indexing unit to 

the drywell atmosphere, the amounts transported to containment by this leak path are negligible 

in comparison to the amounts transported to containment through the SRV on the main steam 

line SRV. 

Differences in calculated results are observed, however, in the reactor building. Leakage through 

the TIP guide tubes begins after significant oxidation of Zircaloy cladding in the core has begun. 

As a result, hydrogen enters the open TIP tubes and is slowly discharged directly to the reactor 

building. Figure 5-65 compares total mass of hydrogen in the reactor building in the sensitivity 

calculation to the amount in the baseline LTSBO calculation. The slow accumulation of 

hydrogen in the building due to TIP tube leakage is clearly evident in the sensitivity calculation. 

In contrast, the building is nearly free of hydrogen until containment failure occurs in the 

baseline (unmitigated) LTSBO calculation.
5-76

Figure 5-65 Total Hydrogen Mass in the Reactor Building (TIP Sensitivity calculation 

versus unmitigated LTSBO baseline)

Approximately 30 kg of hydrogen is released to the reactor building over a period of 10 hours 

before containment failure. This hydrogen accumulates primarily in the atmosphere of the 

ground level floor of the building, but is also carried upward through the open equipment chase 

due to buoyancy. Although sufficient oxygen is available to react with the hydrogen, maximum 

hydrogen concentrations in these regions of the building are well below the flammability 

criterion of 10%. The maximum concentration calculated (just prior to containment failure) is 

approximately 2% in the volume closest to the release point at the TIP room. As a result, 

combustion of hydrogen in the reactor building is not observed in the sensitivity calculation until 

after containment failure.

Low hydrogen concentrations during the early period of TIP tube leakage are partly due to the 

assumption of uniform mixing within the large control volumes used to represent the building 

atmosphere. Although the reactor building is sub-divided into several hydrogen dynamic control 

volumes (two are used to represent the north and south halves of the building where the TIP 

release point is located, for example), the free volume of building is rather large. On the other 

hand, the total amount of hydrogen released (30 kg) is very small in comparison to the amounts 

released from the containment when drywell head flange leakage or (more importantly) liner 

melt-through occurs. This is evident in Figure 5-66, which compares the integral quantity of 

hydrogen released to the reactor building. The amount released prior to containment failure in 

the TIP leakage sensitivity calculation is overwhelmed by the amount released as a consequence 

of containment failure. 

0

50

100

150

200

250

300

6 12 18 24 30

Total H2 Mass in RB [kg]

time [hr]

LTSBO Baseline

LTSBO - TIP leakage

TIP tube

melts in

the core

Containment

failure
5-77

Figure 5-66 Integral Mass of Hydrogen Released to the Reactor Building (TIP Sensitivity 

calculation versus unmitigated LTSBO baseline)

Fission products are also released through the leaking TIP tubes. Figure 3 shows the long-term 

release of fission products to the environment is affected only slightly by the early release 

through leaking TIP tubes. As shown in Figure 5-67, the amount of volatile aerosol species 

released to the environment immediately following containment is larger in the sensitivity 

calculations than the baseline calculation. Early leakage of iodine, cesium and (to a lesser 

extent) molybdenum to the reactor building through the open TIP tubes creates a stable airborne 

concentration of volatile fission product aerosols in the building. For example, Figure 5-68

compares the amount of iodine in the reactor building (airborne plus deposited) in the sensitivity 

calculation to the amounts observed in the baseline LTSBO calculation. The airborne mass of 

fission product is swept out of the building by the relatively large discharge of gas accompanying 

containment failure and subsequent hydrogen burns. This increases the amount released to the 

environment immediately following containment failure, but (as indicated in Figure 5-67) has a 

negligible effect on the larger, long-term integral release of iodine to the environment. 

0

250

500

750

1000

1250

1500

1750

2000

6 12 18 24 30

Integral H2 Mass to RB [kg]

time [hr]

LTSBO - Base

LTSBO - TIP leak
5-78

Figure 5-67 Fractional Release of Key Fission Products to the Environment (TIP 

Sensitivity calculation versus unmitigated LTSBO baseline)

Figure 5-68 Iodine and Cesium in the Reactor Building (TIP Sensitivity calculation 

versus unmitigated LTSBO baseline)

1.E-04

1.E-03

1.E-02

1.E-01

1.E+00

0 6 12 18 24 30 36 42 48

Fraction of Initial Core Inventory

time [hr]

Fission Products 

in Environment

NG

Mo

Cs

I

Containment

failure

Max TIP leakage

with tube rupture

sensitivity

Baseline LTSBO

0

0.002

0.004

0.006

0.008

0.01

0.012

0.014

0 6 12 18 24 30 36 42 48

Fraction of Initial Core Inventory

time [hr]

Iodine in the

Reactor Building

Containment failure

TIP leakage

sensitivity

LTSBO baseline
5-79

Early leakage of fission products through the open TIP guide tubes shifts the time at which 

radioactive aerosols are released to the environment by several hours. As indicated in Figure 

5-69, noble gases begin to leak to the environment in small quantities soon after core damage 

begins. The baseline unmitigated LTSBO calculation assumes the containment leaks at the 

design basis leak rate of approximately 0.5 volume percent per day. This results in a very small, 

but steady, release of noble gas starting as early as 11 hours after the initiating event in the 

baseline unmitigated LTSBO calculation, and approximately one hour earlier in the sensitivity 

case with TIP tube leakage. The temporal effect on leakage of radioactive aerosols is much more 

significant. Iodine, cesium and other volatile species are released to the environment at rates 

comparable to noble gases (in terms of the fraction of core inventory released) for many hours 

prior to the time of containment failure.

Figure 5-69 Early Leakage of Key Fission Products to the Environment (TIP Sensitivity 

calculation versus unmitigated LTSBO baseline)

5.8 Barium Release Variations

In the preceding sections, some considerable variability is observed with respect to Barium 

releases, with a general trend that environmental releases for the short-term SBO are on the order 

of 8% to 10% and for the long-term SBO, on the order of 0.5% to 1%, as illustrated in 

Figure 5-70. The observed moderately large and consistent trend is due mainly to differences in 

release rate from fuel during the ex-vessel molten core-concrete interaction (MCCI) phase of the 

accidents, which in turn is strongly influenced by the MCCI temperatures realized during the 

respective MCCI phases. As seen in Figure 5-70, the MCCI temperatures are moderately higher 

in the short term SBO cases, in part due to higher in-vessel core material temperatures associated 

with the comparatively higher decay heat levels in STSBO versus LTSBO. This bias also favors 

1.E-10

1.E-09

1.E-08

1.E-07

1.E-06

1.E-05

1.E-04

1.E-03

0 2 4 6 8 10 12 14 16 18 20 22 24

Fraction of Initial Core Inventory

time [hr]

Fission Products 

in Environment

NG

Mo

Containment

failure

Cs

Max TIP leakage I

with tube rupture

sensitivity

Baseline LTSBO

Cs

I

NG
5-80

a more vigorous initial MCCI reaction rate which also drives MCCI temperatures upwards. 

Finally, the Ba release rate is strongly affected by the chemical form of Barium, with the reduced 

form of Ba metal being significantly more volatile than the oxidized form BaO. These effects 

are explained as follows.

Figure 5-70 Barium release trends in STSBO compared to LTSBO

The speciation of the Barium during MCCI between the metallic and oxide forms is strongly 

influenced by the reduction/oxidation conditions in the ex-vessel core melt mixture, and this in 

turn is strongly coupled to the presence of (or lack of) Zr metal in the melt, Zirconium being one 

of the strongest reducing agents in the periodic table. On exiting the failed vessel, the core 

material contains significant quantities of unoxidized zirconium metal. This is especially true for 

BWRs because of the large mass of Zr associated with the fuel channel boxes in comparison with 

PWRs. This Zr metal relatively quickly oxidizes to ZrO2 during the early stages of the MCCI, 

and therefore is short-lived in the overall duration of the MCCI phase; however, during the time 

that metallic Zr persists in the MCCI, the speciation of Barium is shifted towards favoring the 

reduced metallic Ba form. The speciation tendency can be roughly expressed by the following 

highly simplified net chemical reaction,

Zr + 2 BaO → ZrO2 + 2Ba

At the same time, Zr is also reacting with steam from concrete decomposition, producing 

hydrogen gas,

0

500

1000

1500

2000

2500

0

0.02

0.04

0.06

0.08

0.1

0.12

0.14

0.16

0 4 8 12 16 20 24

MCCI Temperature [K]

fractional release

time [hr]

Fuel Release - STSBO

Env Release - STSBO

Fuel Release - LTSBO

Env Release - LTSBO

Temperature - STSBO

Temperature - LTSBO

Ba Boiling Point

time period

with unoxidized Zr

in ex-vessel melt

time period

with unoxidized Zr

in ex-vessel melt
5-81

Zr + 2 H2O → ZrO2 + 2H2

further driving chemically reducing conditions as suggested by the following reaction

BaO + H2 → Ba + H2O

All of these reactions have negative Gibbs energy changes that drive the reactions in the 

direction indicated. Moreover, the high volatility of the reaction product Ba and the sparging 

effect of the evolving gases facilitate mass transfer and work to ensure that the reaction continues 

in the direction indicated in the reaction by keeping the concentration of Ba in the melt at a low 

value. As the Zr metal in the MCCI is consumed, principally by reaction with the steam evolved 

from the dehydration of the concrete, the reduction of BaO to Ba dissipates and thereafter BaO is 

thermodynamically favored.

The importance of the chemical form for Ba, determined by the presence of (or lack of) metallic 

Zr, is with the differing volatility of the metallic versus oxide form, as illustrated in Figure 5-71, 

where estimates for the vapor pressures for Ba and BaO are shown. In this graph, the Ba vapor 

pressure is taken from the MELCOR material properties, whereas the trend for BaO is simply 

approximated by interpolating between the Ba melting point and boiling point taken from 

chemical handbook data.

Figure 5-71 Vapor pressure of Ba metal and BaO (oxide vapor pressure is estimated)

As seen in Figure 5-71, the Ba metal shows a wide range of volatility between the range of 

temperatures typical of MCCI (1600 K to 2000 K), whereas, the BaO form does not have high 

volatility until temperatures exceed roughly 2250 K. 

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

1

1000 1500 2000 2500

Vapor Pressure [bar]

Temperature [K]

Ba pressure

BaO pressure

BaO

melt

BaO boil
5-82

The trends observed in Figure 5-70 can now be explained. For both the LTSBO and the STSBO, 

prior to vessel failure, the fractional release of Barium in-vessel are nearly identical at about 2% 

of the core inventory. Following lower head failure, in both the short term and long term SBO, 

there is a brief period of time when unoxidized Zr is present in the ex-vessel MCCI indicated by 

the light green shaded regions in Figure 5-70. During that time, the Ba release is seen to increase 

over the in-vessel released quantity. This renewed release is made possible by the Zr-reduction 

of BaO to the metallic form which shows significant volatility in the range of temperatures 

driven by the MCCI; however, the ex-vessel release for the STSBO case is significantly greater 

than for the LTSBO. This is because MCCI temperatures for the STSBO actually exceed the 

boiling point of the Barium metal. This difference in MCCI temperature is attributable to the 

temperatures attained by the core materials prior to lower head failure and largely attributable to 

the significantly larger decay heat level associated with the STSBO compared with the LTSBO, 

when the decay heat has dropped significantly. These trends are suggested in Figure 5-72 and 

Figure 5-73. Finally, metallic Zr release from the vessel to the drywell cavity is larger in the 

STSBO case in comparison to the LTSBO case where relocation takes place over a more 

protracted time frame and in two stages. These factors in combination would appear to explain 

the higher Ba releases in the STSBO cases relative to the LTSBO case.

Figure 5-72 Relative core damage states for STSBO versus LTSBO. (The STSBO damage 

is slightly more extensive with portions of the core forming molten pools.)

STSBO

29020 (sec)

LTSBO

67988 (sec)
5-83

Figure 5-73 Maximum core temperature for STSBO and LTSBO analyses. (At the time of 

lower head failure, core temperatures are somewhat higher in the STSBO case.)

Figure 5-74 Zr content in-vessel

As shown in Figure 5-74, relocations to cavity are evident at just after 8 hours in the STSBO and 

just before 20 hours in the LTSBO. More unoxidized Zr is released promptly in the STSBO 

case, whereas two relocations events are evident in the LTSBO case.

0

500

1000

1500

2000

2500

3000

0 4 8 12 16 20 24

Temperature [K]

time (hr)

STSBO

LTSBO

time of lower

head failure

0

10000

20000

30000

40000

50000

60000

70000

0 8 16 24 32 40 48

Mass [kg]

time [hr]

LTSBO

STSBO

Zr reduction due 

to oxidation to ZrO2

Zr relocation 

to drywell cavity

delayed second Zr 

relocation to drywell

cavity
5-84

After the metallic Zr has been depleted from the MCCI, both cases show the release of Ba to 

drop significantly, essentially ceasing as the thermodynamics thereafter favor the lower volatile 

BaO chemical form. The environmental releases simply follow the release trends from the fuel, 

mainly the MCCI-driven Ba release. In closing this discussion, it should be noted that the 

ex-vessel release of Barium is treated by the VANESA module of MELCOR, which considers 

the thermodynamic effects on Barium speciation, albeit in the context of a much larger system of 

chemical elements. This sophisticated treatment enables MELCOR to capture this transient high 

volatility of Ba during the Zr-rich phase of the MCCI. In contrast, in-vessel releases of Ba are 

treated using the less sophisticated CORSOR-Booth model for release from fuel. The in-vessel 

releases from the degrading fuel rods predicted using the Booth model are well validated against 

experimental data from testing programs. It could be argued however releases from the lower 

head molten fuel mixture are not well studied and could warrant a thermodynamic treatment as 

in VANESA as opposed to the Booth treatment as currently employed in MELCOR. That said, 

low surface to volume ratio for the in-vessel molten pool stages of core melt progression do not 

favor large release rates mainly from mass transport limitations. For this reason, a full 

thermodynamic modeling treatment is not warranted in this stage of the accident.
6-1

6.0 EMERGENCY RESPONSE

Emergency response planning is in place to protect the public health and safety in the unlikely 

event of an accident at a nuclear power plant. Advancements in the consequence model now 

allow more detailed and more realistic treatment of emergency response when performing 

consequence analyses. This includes the ability to model protective action decisions from offsite 

response organizations (OROs) and the implementation of protective actions by individual 

population segments. These advancements are significant because they facilitate more realistic 

modeling of response activities, timing of decisions, and implementation of protective actions 

across different population segments. 

Emergency response programs are developed, tested, evaluated, and established as defense 

in-depth. To support a state-of-the art approach and integrate realism into the analyses, the 

modeling of the emergency response was based on the site-specific emergency planning 

documentation. The information developed in this emergency response section was used to 

support the MACCS2 consequence analyses for the accident scenarios. These analyses are 

conducted for the unmitigated accident sequences only. Many of these response actions would 

be similar for the mitigated case because response officials initiate protective actions upon 

notification, which as described herein occurs very early in the incident, even before mitigation 

actions have been implemented. This is because emergency planning is designed to be proactive 

to remove the public prior to plume arrival when possible. For each accident scenario, evacuation 

of the plume exposure pathway emergency planning zone (EPZ) was assessed. This included 

consideration of a shadow evacuation to a distance of 20 miles from the plant. Including a

shadow evacuation, which occurs when members of the public evacuate from areas that are not 

under official evacuation orders, provides realism because these are observed in large-scale 

evacuations and have the potential to slow down the evacuation from the affected area. Also, for 

each scenario, members of the public were modeled as being relocated from any area where 

doses are projected, based on the consequence analysis model, to exceed established criteria. 

Figure 6-1 identifies the location of the Peach Bottom plant and radial distances of 10 and 20 

miles from the plant. 

Sensitivity analyses were completed to evaluate evacuating the public to distances of 16 miles 

and 20 miles from the plant. The sensitivity analysis of an evacuation to 20 miles is different 

than the shadow evacuation to 20 miles described above, because the sensitivity analysis 

evaluates the conditions under which residents of the entire 20 mile area are notified to evacuate 

and leave the area. A sensitivity analysis was also completed to assess the effect of a delay in the 

implementation of protective actions, as suggested by the peer review committee. An analysis 

was also conducted to evaluate the effects on infrastructure, emergency response, and response 

of the public due to a seismic event. 
6-2

Figure 6-1 Peach Bottom 10 and 20 mile analysis areas

As required by 10 CFR Part 50, “Domestic Licensing of Production and Utilization Facilities,” 

OROs develop emergency response plans for implementation in the event of a nuclear power 

plant accident. These plans are regularly drilled and inspected biennially through a 

demonstration exercise performed with the licensee. In biennial exercises, (every 2 years) ORO 

personnel demonstrate timely decision making and the ability to implement public protective 

actions. Emergency plans escalate response activities in accordance with a classification scheme 

based on emergency action levels (EALs). Preplanned actions are implemented at each 

classification level, including Unusual Event, Alert, Site Area Emergency (SAE), and General 

Emergency (GE). Public protective actions are required at the GE level, but ORO plans 

commonly include precautionary protective actions at the SAE level and sometimes at the Alert

level. For example, at Peach Bottom sirens are sounded at the SAE to inform the public that an 

incident has occurred and they should monitor Emergency Alert System (EAS) stations for 

updated information.

The plume exposure pathway EPZ is identified 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 [21], as the area of 

about 10 miles around a nuclear power plant. Within the EPZ, detailed emergency plans are in 

place to reduce the risk of public health consequences in the unlikely event of an accident. 

Emergency planning within the EPZ provides a substantial basis for expansion of response 
6-3

efforts if necessary [21]. During the inspected biennial exercises, ORO personnel have 

repeatedly demonstrated the ability to implement protective actions within the EPZ. Modeling of 

the expected protective action responses described in this section is consistent with exercise 

performance data. The analysis includes the State of Pennsylvania position that, if an evacuation 

is ordered, it will include the entire EPZ. This position differs from other states, where 

evacuation of downwind areas would be implemented rather than the full EPZ. For the analyses 

in this report, a full evacuation was modeled assuming that the State of Maryland OROs would 

agree with the Pennsylvania protective action decisions. Analyses were conducted for the 

accident scenarios identified in Table 6-1.

Table 6-1 Scenarios Assessed for Emergency Response

Section # Scenario

6.3.1 LTSBO- unmitigated

6.3.2 STSBO- with RCIC blackstart

6.3.3 STSBO- without RCIC blackstart

6.4.1 Sensitivity 1 for the STSBO (without RCIC 

blackstart) and evacuation to 16 miles

6.4.2 Sensitivity 2 STSBO (without RCIC blackstart) 

and evacuation to 20 miles

6.4.3

Sensitivity 3 for the STSBO (without RCIC 

blackstart) with a delay in implementation of 

protective actions

6.5.6 Seismic analysis - STSBO without RCIC 

blackstart

Analysis was conducted for the accident scenarios identified in Table 6-1. As indicated in the 

table, a seismic analysis was completed for the STSBO without RCIC blackstart. This analysis 

included the effects of damage to infrastructure, such as roadway impacts, loss of traffic 

signalization, and emergency responder priorities. The remaining analyses completed in this 

section evaluate the expected response of the public and emergency responders without regard to 

issues related to the initiating event.

6.1 Population Attributes 

The population near the Peach Bottom plant was modeled using six cohorts. A cohort is a 

population group that mobilizes or moves differently from other population groups. Cohorts

were established to represent members of the public who may evacuate early, evacuate late, 

those who refuse to evacuate, and those who evacuate from areas not under an evacuation order 

(e.g., the shadow evacuation). The consequence model does not constrain the number of cohorts, 

but there is no benefit to defining an excessive number of cohorts with little difference in 

response characteristics. The following cohorts were established for SOARCA analyses: 

Cohort 1: 0 to 10 Public. This cohort includes the public residing within the EPZ.
6-4

Cohort 2: 10 to 20 Shadow. This cohort includes the shadow evacuation from the 10 to 20 mile 

area beyond the EPZ. A shadow evacuation occurs when members of the public evacuate from 

areas that are not under official evacuation orders and typically begin when a large scale 

evacuation is ordered [22]. Although shadow evacuations are often reported and observed there 

is little quantitative data available. In a national telephone survey of residents of EPZs, questions 

were asked about evacuation, about 20 percent of people that had been asked to evacuate had 

also evacuated for situations in which they were asked not to evacuate [31]. Additional 

information used to develop a value for use in the SOARCA analysis included a review of more 

than 20 nuclear power plant evacuation time estimates (ETEs). Although not currently required, 

most of these ETEs included an analysis of a shadow evacuation. Typically, a shadow 

evacuation of 30 percent of the public outside the EPZ to a distance of 15 miles was included and 

often sensitivity analyses were provided that varied the shadow evacuation percentage to values 

as high as 60 percent. Review of these ETEs showed that increasing the shadow evacuation to 

60 percent typically did not affect the ETE. Using the above information, combined with the 

early decision in the SOARCA project to consider effects beyond the EPZ to a distance of 20 

miles, a shadow evacuation of 20 percent of the public from the area 10 to 20 miles from the 

plant was modeled. For this primarily rural area, a 20 percent shadow evacuation from the area 

beyond the EPZ has no effect on the evacuation of the EPZ residents. 

Cohort 3: 0 to 10 Schools and 0 to 10 Shadow. This cohort includes elementary, middle, and 

high school student populations within the EPZ. Schools receive early and direct warning from 

OROs and have response plans in place to support busing of students out of the EPZ. A shadow 

evacuation from within the EPZ is included with this cohort because sirens are sounded at SAE. 

This is expected to stimulate an evacuation of some of the residents from within the EPZ 

beginning about the same time as the evacuation of the schools. Because the 0 to 10 Shadow 

would evacuate at about the same time as the schools, only one cohort was used to represent the 

two population groups.

Cohort 4: 0 to 10 Special Facilities. The special facilities population includes residents of 

hospitals, nursing homes, assisted living communities, and prisons. Special facility residents are 

assumed to reside in robust facilities such as hospitals, nursing homes, or similar structures that 

provide additional shielding. Shielding factors for this population group consider this fact.

Cohort 5: 0 to 10 Tail. The 0 to 10 tail is defined as the last 10 percent of the public to evacuate 

from the 10 mile EPZ. The approach to modeling the tail is an analysis simplification to support 

inclusion of this population group. The Tail takes longer to evacuate for many reasons such as 

the need to return home from work to evacuate with the family, pick up children, shut down 

farming or manufacturing operations or performing other actions prior to evacuating. It also 

includes those who may miss the initial notification. 

Cohort 6: Non-Evacuating Public. This cohort represents a portion of the public from 0 to 10 

miles who may refuse to evacuate. It is assumed to be 0.5 percent of the population who are 

modeled as though they are performing normal activities. Research on large scale evacuations 

has shown that a small percentage of the public refuses to evacuate [22]; this cohort accounts for 

this potential group. It is important to note that emergency planning is in place to support 

evacuation of 100 percent of the public.
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6.1.1 Population Distribution

The total Peach Bottom population for the 0 to 20 mile area was obtained from SECPOP for the 

year 2000. That population value was projected to 2005 using a national level multiplier of 

1.0533 obtained from the Census Bureau. The Peach Bottom ETE presents a detailed estimate of 

the population within the 0 to 10 mile region, and was used to develop population fractions 

which are used in MACCS2. Table 6-2 summarizes the populations for each cohort used for the 

SOARCA analyses of the Peach Bottom site.

Table 6-2 Peach Bottom Cohort Population Values

Cohort Description Population

1 Public (0 to 10) 24,110

2 Shadow (10 to 20) 82,661

3 Schools (0 to 10) and Shadow (0 to 10) 16,160

4 Special Facilities (0 to 10) 261

5 Tail (0 to 10) 2,693

6 Non-Evacuating Public (0 to 10) 217

6.1.2 Evacuation Time Estimates

As provided in Appendix E, “Emergency Planning and Preparedness for Production and 

Utilization Facilities,” to 10 CFR Part 50 and 10 CFR 50.47, “Emergency Plans”, each licensee 

is required to estimate the time to evacuate the EPZ. Appendix 4 to

NUREG-0654/FEMA-REP-1, Revision 1 [21], provides information on the requirements of 

ETEs, and NUREG/CR-6863, “Development of Evacuation Time Estimate Studies for Nuclear 

Power Plants,” issued January 2005 [24], provides detailed guidance on the development of 

ETEs. A typical ETE includes many evacuation scenarios to help identify the combination of 

events for normal and off-normal conditions55 and provides emergency planners with estimates 

of the time to evacuate the EPZ under varying conditions [24]. The ETE study provides 

information on population characteristics, mobilization of the public, special facilities, 

transportation infrastructure and other information used to estimate the time to evacuate the EPZ. 

Some of the population values used in SOARCA do not directly align with ETE values because 

populations were grouped differently in the SOARCA study than in the ETE. For instance, the 

special facilities population in the ETE represents schools, nursing homes, and hospitals with an 

estimated total population of 12,989 persons for a winter weekday. In the SOARCA study, the 

schools and special facilities were represented as separate cohorts. The school population was 

combined with the 0 to 10 Shadow evacuation cohort in the SOARCA study because the school 

and shadow evacuation populations mobilize at the same time. The schools and shadow cohort 

population value used was 16,160.

The SOARCA project used a normal weather winter weekday scenario that includes schools in 

session. This scenario was selected because it presents several challenges to timely protective 

 

55 The term “off-normal condition” includes unique weather, sporting or entertainment events or other 

occurrences that would significantly disrupt normal population movements.
6-6

action implementation, including evacuating while residents are at work and mobilizing buses to 

evacuate children at school.

The most recent ETE available from the licensee was used to establish evacuation speeds and 

delay times within the EPZ. The following ETEs, rounded to the nearest quarter hour, were used 

to develop evacuation speeds. These ETEs correspond to the normal weather winter weekday 

scenario:

• 100-percent evacuation: 5 hours and 15 minutes 

• 90-percent evacuation: 4 hours and 15 minutes.

The summer weekend ETE and the winter weeknight ETE were also provided in the licensee’s 

most recent ETE study and were considered for use; however, the ETEs for these two scenarios 

were each approximately 4 hours and 45 minutes for the 100 percent evacuation. The winter 

weekday scenario with an ETE of 5 hours and 15 minutes can be considered the bounding ETE 

case for the analysis. The evacuation speed, which is an input into the consequence model, was 

developed from the ETE and is primarily influenced by population density and roadway 

capacity.

When using ETE information, it is important to understand the components of the time estimate. 

The ETE includes mobilization activities that the public undertakes upon receiving the initial 

notification of the incident [24]. These actions include receiving the warning, verifying 

information, gathering children, pets, belongings, packing, securing the home, and other 

evacuation preparations. Thus, a 5 hour ETE does not indicate that all of the vehicles are en 

route for 5 hours; rather, it is the end of a 5 hour period in which the public mobilizes and 

evacuates the area. An evacuation population does not enter the roadway system at once. Rather 

an ideal model would include a ‘road loading function’ that represents the expected movement. 

Most ETE studies use such a model. However, MACCS2 does not currently have the capability

to move populations in this manner. This being the case, cohorts are modeled to begin moving 

together at a specific time after notification. To represent the movement of the cohort evacuating 

together, a single linear value of distance divided by time (the ETE) was used. This distance 

over ETE ratio provides a slightly slower average speed than would be expected in an evacuation 

and adds some conservatism to the analysis. Adjustment factors within the consequence model 

were used to increase or decrease speeds for each cohort at the grid level.

The time to complete an evacuation can be represented as a curve that is relatively steep at the 

beginning and tends to flatten as the last members of the public exit the area. A review of more 

than 20 existing ETE studies indicated that the point at which the curve tends to flatten occurs 

when approximately 90 percent of the population has evacuated. This is consistent with research 

that has shown that a small portion of the population takes a longer time to evacuate than the rest 

of the general public and is the last to leave the evacuation area. This last 10 percent of the 

population is identified as the evacuation tail. For the analyses in this study, the 90 percent ETE 

value was used to develop evacuation speeds, and the 10 percent evacuation tail was analyzed as 

a separate cohort. 
6-7

6.2 WinMACCS 

WinMACCS, a user interface for the MACCS2 code, was used to generate input for MACCS2 

model runs. WinMACCS integrates the information described above into the consequence 

analysis. The evacuation area was mapped onto the WinMACCS radial sector grid network. The 

roadway network was reviewed against site-specific evacuation plans to determine the likely 

evacuation direction in each grid element. The results of the ETE were reviewed to determine 

localized areas of congestion and areas where no congestion would be expected. Speed 

adjustment factors were applied at the grid element level to speed up vehicles in the rural 

uncongested areas and to slow vehicles in more urban settings where the modeling indicates that 

speeds are lower than the average values used in the analyses. 

6.2.1 Hotspot and Normal Relocation and Habitability

In the unlikely case of a severe accident and radiological release, protective actions in addition to

evacuation may be implemented. For instance, residents would be relocated from areas where the 

dose exceeds protective action criteria. OROs would base this determination on dose projections 

using State, utility, and Federal agency computer models as well as measurements taken in the 

field. Hotspot relocation and normal relocation models are included in the MACCS2 code to 

reflect this activity and include dose from cloudshine, groundshine, direct inhalation, and 

resuspension inhalation. Within the MACCS2 calculation, individuals who would be relocated 

because their projected total committed dose from these pathways is projected to exceed the 

protective action criteria are prevented from receiving any additional dose during the emergency 

phase. The emergency phase is the 7 day period after the start of the release. This reflects the 

impact of the relocation of these individuals that would take place in the event of an actual 

radiological release. This relocation dose criterion is applied at a specified time after plume 

arrival at the affected area. Relocation is applied to the entire population within the analysis area 

to a distance of 50 miles, including the non-evacuating cohort within the EPZ, even though this 

small fraction of the population does not comply with previous evacuation orders. It is assumed 

these individuals will evacuate when they understand that a release has in fact occurred and they 

are informed that they are located in high dose areas.

For hotspot relocation, individuals are relocated 12 hours after plume arrival if the total lifetime 

dose commitment for the weeklong emergency phase exceeds 0.05 sievert (Sv) (5 rem). For 

normal relocation, such individuals are relocated 24 hours after plume arrival if the total lifetime 

dose commitment exceeds 0.005 Sv (0.5 rem). The relocation times of 12 hours for hotspot and 

24 hours for normal relocation were established based on review of the emergency response time 

lines, which suggest that OROs may not be available earlier to assist with relocation because of 

higher priority tasks in the evacuation area. Relocation is a process that requires identification of 

the affected areas and notification of residents within those areas. The time values represent the 

average time expected to implement each action.

Habitability is the consequence model parameter that is used to establish the dose level at which 

residents are allowed to return to the EPZ to live. Site-specific criterion are used for long-term 

habitability, and most states adhere to the U.S. Environmental Protection Agency (EPA) 

guidelines which specify a dose of 2 rem in the first year and 500 millirem per year thereafter. 

The EPA recommendation has traditionally been implemented in MACCS2 as a cumulative 

4 rem over the first 5 years (2 rem in the first year + 4 years x 0.5 rem/year) of exposure. 
6-8

Pennsylvania has a more strict habitability criterion of 500 mrem per year beginning in the first 

year, and this value was used in the Peach Bottom analysis. The hotspot and normal relocation 

values used in NUREG-1150 were 0.5 Sv (50 rem) and 0.25 Sv (25 rem) respectively. The long 

term habitability criteria used in NUREG-1150 was 0.04 Sv (4 rem) over a 5 year period. The 

values used in SOARCA were established to align with site specific response expectations and 

EPA protective action guidelines. 

6.2.2 Shielding Factors

Shielding factors vary by geographical region across the United States; those used in the Peach 

Bottom analysis are shown in Table 6-3 and are appropriate for the region. The factors represent 

the fraction of dose that a person would be exposed to when performing normal activities, 

evacuating, or staying in a shelter in comparison to a person outside with full exposure. These 

shielding factors are applied to all cohorts. Special facilities are typically larger and more robust 

structures than residential housing stock and therefore have better shielding factors, as identified 

in Table 6-3.

Table 6-3 Peach Bottom Shielding Factors

Cohort

Groundshine Cloudshine Inhalation/Skin

Normal Evac. Shelter Normal Evac. Shelter Normal Evac. Shelter

Nonspecial 

facilities

0.18 0.50 0.10 0.60 1.00 0.50 0.46 0.98 0.33

Special 

facilities 0.05 0.50 0.05 0.31 1.00 0.31 0.33 0.98 0.33

The site specific values for sheltering were obtained from NUREG-1150 [9]. An updated 

inhalation/skin evacuation shielding factor was obtained from NUREG/CR-6953, Volume 1, 

“Review of NUREG-0654, Supplement 3, ‘Criteria for Protective Action Recommendations for 

Severe Accidents’,” issued January 2005 [23]. The normal activity shielding factors have been 

adjusted to account for the understanding that people do not spend a great deal of time outdoors. 

The normal activity values are all weighted averages of indoor and outdoor values based on 

being indoors 81 percent of the time and outdoors 19 percent of the time [32]. The shielding 

factor value for indoor activities was assumed to be the same as the shielding factor value for 

sheltering shown in Table 6-3.

6.2.3 Potassium Iodide

The State of Pennsylvania potassium iodide (KI) program distributes KI tablets through several 

different means. The Pennsylvania Department of Health district offices are responsible for 

coordinating with county emergency management agency officials to make KI available to 

residents living and working within the EPZ. The distribution of KI occurs on an annual basis 

for the Peach Bottom EPZ and is preceded by public announcements. 

The purpose of the KI is to saturate the thyroid gland with stable iodine so that further uptake of 

radioactive iodine by the thyroid is diminished. If taken at the right time and in the appropriate 

dosage, KI can nearly eliminate doses to the thyroid gland from inhaled radioiodine. Factors that 

contribute to the effectiveness of KI include its availability, the timing of ingestion, and the 
6-9

degree of pre-existing stable iodine saturation of the thyroid gland, which already inhibits 

absorption of inhaled radioiodine by the thyroid. The analysis assumes that some residents will 

not remember where they have placed their KI or may not have it available and will therefore not 

take KI. It is also assumed some residents will not take their KI when directed (i.e., they may 

take it early or late which reduces the efficacy). To account for these factors, the analysis

modeled KI as taken by 50 percent of the public, and the efficacy of the KI was set at 70 percent.

6.2.4 Adverse Weather

Adverse weather is typically defined as rain, ice, or snow that affects the response of the public 

during an emergency. The affect of adverse weather on the mobilization of the public was not 

directly considered in establishing emergency planning parameters for this project because such 

a consideration approximates a worst-case evacuation scenario. However, adverse weather was 

addressed in the movement of cohorts within the analysis. The evacuation speed multiplier 

(ESPMUL) parameter in WinMACCS is used to reduce travel speed when precipitation is 

occurring as indicated from the meteorological weather file. The ESPMUL factor was set at 0.7, 

which effectively slows down the evacuating public to 70 percent of the established travel speed 

when precipitation exists.

6.2.5 Modeling Using Evacuation Time Estimates

The purpose of using the ETE to develop evacuation parameters for the consequence modeling is 

to better approximate the real time actions expected of the public. Although consequence 

modeling has evolved to allow the use of many cohorts and can address many individual aspects 

of each cohort, the approach to modeling evacuations is not direct. As stated earlier, evacuations 

include mobilizing and evacuating the public over a period of time, which is best modeled as a 

distribution of data. WinMACCS requires this distribution of data be converted into discrete 

events. For example, upon the sounding of the sirens and issuance of the EAS messaging, in the 

SOARCA analysis all members of the public are sheltered and 1 hour later all members of the 

public enter the roadway network together at the same time and begin to evacuate. Research on 

existing evacuations for technological hazards has shown that the evacuating public would 

actually enter the roadway network over a period of time. It is not realistic that all vehicles 

would load simultaneously; however, this treatment within the model is necessary because of the 

current modeling abilities of WinMACCS. 

Considering the above modeling requirement, reasonable speeds, developed from the ETE, were 

established for each cohort. The following elements factor into these speeds:

• time to receive notification and prepare to evacuate (mobilization time),

• time to evacuate, and

• distance of travel.

The time to receive notification was based on the assumption that sirens will sound when needed. 

The NRC Reactor Oversight Program data for siren performance for Peach Bottom shows an 

average siren performance indicator of 99.8 percent, indicating that sirens will perform when 

needed. It is recognized that loss of power accidents will result in the loss of power to some 

offsite areas of the EPZ. However, there is no reason to expect that the power will be out in the 
6-10

entire 10 mile EPZ. For this analysis, it is assumed that the offsite sirens would be sounded 

within much of the EPZ. For those areas where the power outage affects sirens, it is assumed 

societal notification and route alerting by OROs would alert residents within the same 

mobilization time period as estimated for the EPZ. 

A simple ratio of distance to time would show that evacuation of the 0 to 10 Public cohort from 

the 10 mile EPZ at Peach Bottom which has an ETE of 4 hours and 15 minutes, would move at a 

speed of 2.4 miles per hour (mph). However, as indicated above, notification and preparation to 

evacuate are included in the ETE.

For the general public, a 1-hour delay to shelter is assigned to reflect the mobilization time in 

which residents receive the warning and prepare to evacuate. If the 1-hour mobilization time is 

subtracted from the ETE (4 hours and 15 minutes minus 1 hour) there remains 3 hours and 

15 minutes to travel a maximum of 10 miles. Roadways are not radial out of the EPZ and actual 

travel distance is estimated, based on review of local mapping, to be 30 to 50 percent greater 

than the radial distance. 

During the evacuation, roadway congestion occurs rather quickly, and traffic exiting the EPZ 

begins to slow. A review of ETE studies shows that roadway congestion typically occurs in 1 to

2 hours, depending on the population density and roadway capacity of the EPZ. In the SOARCA 

analysis, the 0 to 10 Public is sheltered and preparing to evacuate for 1 hour. The public is then 

loaded onto the roadway and congestion is assumed to occur within 15 minutes. This total time 

of 1 hour and 15 minutes for congestion to occur was established to be consistent with observed 

data from ETE studies. 

The calculation of the speed of evacuees includes the first 15 minutes to the point when

congestion occurs. For this first 15 minutes, evacuees are assumed to travel at 5 mph. The speed 

is slow to account for the model loading all members of a cohort at one time. In the first 

15 minutes at 5 mph, a distance of 1.25 miles has been traveled. At that time congestion is heavy 

and speeds slow for the next 8.75 miles. 

The ETE is 4 hours and 15 minutes for this cohort. Having sheltered and prepared to evacuate 

for 1 hour and then traveled the first 15 minutes at 5 mph, the remaining time is 3 hours (4:15 

minus 1 hour shelter minus 15 minutes at 5 mph). To determine the speed of travel for the 

remaining 8.75 miles, the distance is divided by the time (8.75 miles / 3 hours), which provides a 

speed of 2.9 mph. The calculated speed used in the analysis for this cohort was rounded to 3 

mph. The values were rounded to avoid implying that speeds are precise and because the speeds 

represent an average for the cohort. This approach applies to evacuees travelling 10 miles. Most 

evacuees would travel less distance and some travel a greater distance because the roadways are 

not radial away from the plant. Adjustments were made to the speeds based on review of 

population densities and aerial photos of roadways. The baseline speed of 3 mph was assigned 

to the general public cohort and was then adjusted using speed adjustment factors, ranging from 

75 percent to 150 percent of the average speed, as appropriate for the rural and urban areas.
6-11

6.2.6 Cohort Modeling

The WinMACCS parameters for the cohorts are stored in multi-dimensional arrays. The 

dimensions of the arrays are defined by geographical area. WinMACCS requires the maximum 

dimensions be established with the first cohort. All subsequent cohorts must be defined within 

these array dimensions, meaning that they can extend from the origin to any distance equal to or 

less than the maximum distance established with the first cohort. 

The 0 to 10 Public was expected to be the cohort of greatest interest; therefore, Cohort 1 was 

defined as the 0 to 10 Public and has the same response characteristics as Cohort 2. The cohort 

that extends the greatest distance and defines the limits of the array is the shadow evacuation, 

which is Cohort 2. Thus, within the WinMACCS input file, Cohorts 1 and 2 were defined to 

meet the requirement that maximum distance be established with the first cohort. The 

WinMACCS model input parameters for Cohort 1 were extended from the plant out to the 

maximum array distance of 20 miles, and Cohort 2 extends from the plant out to 10 miles. In the 

WinMACCS input file, Cohort 1 is input as 20 percent of the population from 0 to 20 miles. 

This captures the 20 percent of the population between 10 and 20 miles involved in the shadow 

evacuation beyond the EPZ. As noted earlier, for this site the shadow evacuation has no affect 

on the evacuation of the residents of the EPZ. The combination of Cohorts 1 and 2 from 0 to 10 

miles in the WinMACCS model represent the 0 to 10 Public cohort defined above. For the 

remaining cohorts, application of parameters in the WinMACCS model is direct, and the 

population fractions directly correspond to the cohort descriptions.

6.3 Accident Scenarios 

An emergency response timeline was developed for each accident scenario using information 

from the MELCOR analyses, expected timing of emergency classification declarations, and 

information from the ETE. The timeline identifies points at which cohorts would receive 

instruction from OROs to implement protective actions. In practice, initial evacuation orders are 

based on the severity of the accident and in Pennsylvania would likely include an evacuation of 

the entire EPZ. The emergency planning parameters use a normal workday event, whereas the 

MELCOR analyses assumes minimum staffing consistent with an off-hours event. This adds 

some conservatism to the analysis because an off-hours, nighttime evacuation would typically 

occur a little faster because most residents are at home and schools are not in session. 

6.3.1 Long-Term Station Blackout Unmitigated

The timing of emergency classification declarations was based on Table PBAPS 3-1,

“Emergency Action Level (EAL) Matrix,” contained in site emergency plan implementing 

procedures. The emergency classification timing was reviewed with the licensee for accuracy. 

MS1 of the SAE EAL specifies that an SAE is declared 15 minutes after the initiating event (loss 

of all AC power). Sirens sound about 45 minutes after the SAE is declared. An EAS message is 

broadcast at this time providing notification to residents and transients within the EPZ that there 

is an incident and instructing them to monitor the situation for further information. A GE is 

declared, based on EAL MG1, 45 minutes into the event (coincidentally 15 minutes before the 

issue of the first EAS message related to the SAE) when it is assumed that operators have 

determined that offsite power will not be restored within 2 hours. An EAS message for the GE is 

then broadcast, and sirens sound again 45 minutes after the GE declaration which is 30 minutes 
6-12

after the siren and initial EAS message for the SAE (see Figure 6-2). The EAS message for the 

GE would include instructions for implementing protective actions.

Discussions with site representatives were held to help ensure SOARCA staff properly 

understood the EALs for each accident scenario and emergency response practices. In addition, 

staff reviewed information from formally evaluated emergency response exercise timelines that 

show the times for notification of an emergency declaration, siren activation, and broadcast of 

EAS messaging for Peach Bottom. The timing in the exercises shows approximately 50 minutes 

from notification to sirens sounding. As indicated in Figure 6-2, the analysis used an estimate of 

45 minutes from SAE declaration to SAE siren, which closely approximates the exercise values. 

The offsite emergency plans for Peach Bottom include sounding sirens for declaration of both 

SAE and GE. The emergency response timeline for the LTSBO scenario is shown in Figure 6-2. 

The first fuel cladding gap release occurs about 9 hours into the event and release to the 

environment begins about 20 hours into the event. The duration of specific protective actions for 

each cohort is summarized in Figure 6-3.

Figure 6-2 Unmitigated LTSBO emergency response timeline
6-13

Figure 6-3 Duration of protective actions for unmitigated LTSBO

Cohort 1: 0 to 10 Public. Following declaration of the SAE, sirens are sounded to alert the 

public. Sirens are again sounded following the declaration of GE, at which time the public is 

assumed to shelter. The time for the public to receive the warning and prepare to evacuate is 

assumed to be 1 hour after the siren sounds for the GE which is consistent with empirical data 

from previous evacuations [26].

 

Cohort 2: 10 to 20 Shadow. This cohort is assumed to begin movement at the same time as the 

0 to 10 Public cohort, once widespread media broadcasts are underway. Residents in the 10 to 

20 mile area begin seeing large numbers of people evacuating and initiate a shadow evacuation. 

There is no warning or notification for the public residing in this area, and the area is not under 

an evacuation order.

Cohort 3: 0 to 10 Schools and 0 to 10 Shadow. Schools are the first to take action. Upon receipt 

of the site’s declaration of an SAE, county emergency management agencies would notify the 

schools in accordance with the offsite emergency response plan. The analysis assumes that 

schools begin sheltering when notified and that this notification occurs about 15 minutes after the 

emergency declaration. Buses would be mobilized, but in accordance with the offsite emergency 

plan, evacuation would not begin until a GE is declared. The preliminary action to mobilize 

buses in response to the SAE allows for a prompt evacuation. It is assumed that schools begin 

evacuating 15 minutes after GE, and it is only coincidence that this occurs about the same time 

as the sounding of the sirens in response to the SAE. At this time in the event, roads are 

uncongested and school buses are able to exit the EPZ quickly. It is assumed that the sounding 

of sirens and broadcast of the EAS message for the SAE causes a shadow evacuation of residents 

within the EPZ (e.g., 0 to 10 Shadow cohort). Because the shadow population is grouped with 

the schools (see Section 6.1), the shadow population is also treated as if it shelters at 15 minutes

after the declaration of SAE.

Cohort 4: 0 to 10 Special Facilities. Special facilities can take longer to evacuate than the 

general public because transportation resources, some of which are specialized such as 

wheelchair vans and ambulances, must be mobilized. Special facilities evacuate individually 

with each facility responsible for obtaining resources. These resources are required to be 
6-14

established during emergency planning; therefore, it is a reasonable assumption that the 

resources will be available. These types of vehicles sometimes must make return trips until 

everyone is evacuated. For modeling convenience, the conservative assumption was made that 

the residents of these facilities remain sheltered and evacuate in a single wave beginning when 

the tail cohort begins to evacuate. It is recognized that some facilities would actually mobilize 

and evacuate earlier in the event.

Cohort 5: 0 to 10 Tail. Using the evacuation data provided in the Peach Bottom ETE study, the 

analysis assumes that 90 percent of the evacuation of the EPZ is complete at approximately 

4 hours and 15 minutes. This corresponds to the departure time for the 0 to 10 Tail cohort. 

Cohort 6: Non-Evacuating Public. This cohort group represents a portion of the public who may 

refuse to evacuate. It is assumed to be 0.5 percent of the population. Any member of the public 

who does not evacuate the EPZ is still subject to the hotspot and normal relocation criterion 

discussed earlier. 

The evacuation timing and speeds for each cohort are presented in Table 6-4. Selected input 

parameters for WinMACCS are provided in Table 6-4 to support detailed use of this study. 

More detailed information about modeling parameters is available in the MACCS2 user’s guide 

[27]. The following is a brief description of the parameters:

• Delay to shelter (parameter DLTSHL) represents a delay from the time of the start of the 

accident until cohorts enter the shelter. DLTSHL is generally referenced to alarm time, 

but the value (OALARM) is set to zero in the SOARCA analyses. 

• Delay to evacuation (parameter DLTEVA) represents the length of the sheltering period 

from the time a cohort enters the shelter until the point at which it begins to evacuate. 

• The speed (parameter ESPEED) is assigned for each of the three phases used in 

WinMACCS including Early, Middle, and Late. Average evacuation speeds were 

derived from the Peach Bottom ETE report. Speed adjustment factors were then used in 

the WinMACCS application to represent free flow in rural areas and congested flow in 

urban areas.

• Duration of beginning phase (parameter DURBEG) is the duration assigned to the 

beginning phase of the evacuation and may be assigned uniquely for each cohort.

• Duration of middle phase (parameter DURMID) is the duration assigned to the middle 

phase of the evacuation and may also be assigned uniquely for each cohort.

For the 0 to 10 Public and the 0 to 10 Tail cohorts, by definition the sum of the DLTEVA, 

DURBEG and DURMID is equal to the ETE. This is because the ETE does not include shelter 

time.
6-15

Table 6-4 Unmitigated LTSBO Cohort Timing

Cohort

Delay to 

Shelter

DLTSHL 

(hr)

Delay to 

Evacuation

DLTEVA 

(hr)

DURBEG 

(hr)

DURMID 

(hr)

ESPEED*

(early)

mph

ESPEED*

(mid)

mph

0 to 10 Public 1.5 1 0.25 3 5 3

10 to 20 Shadow 1.5 1 0.25 3 5 3

0 to 10 

Schools and Shadow 0.25 0.75 1 0.5 20 20

0 to 10 

Special Facilities 1.5 4.25 0.5 0.5 3 20

0 to 10 Tail 1.5 4.25 0.5 0.5 3 20

Non-Evac 0 0 0 0 0 0

* 20 mph was used for the late phase evacuation speed for all cohorts.

 

6.3.2 Short-Term Station Blackout with Reactor Core Isolation Cooling Blackstart

The timing of emergency classification declaration for the STSBO with RCIC Blackstart was 

based on Table PBAPS 3-1 in the site emergency plan implementing procedures. The analysts 

reviewed the emergency classification timing with the licensee for accuracy. This scenario is an 

immediate GE. With loss of offsite power and loss of DC power, operators cannot determine 

whether water level is above TAF, and a GE is declared based on EAL MG1. The emergency 

response timeline for the STSBO scenario is shown in Figure 6-4 and protective action durations 

for each cohort are shown in Figure 6-5. Core damage, as evidenced by the first fuel cladding 

gap release, is calculated at 6.8 hours into the event, with a significant radioactive release from 

containment beginning 16.7 hours into the event as indicated by the lower head failure in 

Figure 6-4. 
6-16

Figure 6-4 STSBO with RCIC blackstart emergency response timeline

Figure 6-5 Protective actions for STSBO with RCIC blackstart

Cohort 1: 0 to 10 Public. It is assumed to take 45 minutes for OROs to sound sirens following 

declaration of a GE, at which point the public is assumed to shelter. The time for the public to 

receive the warning and prepare to evacuate is assumed to be 1 hour after the siren sounds.

 

Cohort 2: 10 to 20 Shadow. This cohort is assumed to begin movement at the same time as the 

0 to 10 Public once widespread media broadcasts are underway. Residents in the 10 to 20-mile 
6-17

area begin seeing large numbers of people evacuating and initiate a shadow evacuation. There is 

no warning or notification for the public residing in this area, and area is not under an evacuation 

order. 

Cohort 3: 0 to 10 Schools and 0 to 10 Shadow. This cohort is the first to take action. Upon 

receipt of the declaration of GE, county emergency management agencies would notify the 

schools in accordance with the emergency response plan. It is assumed that schools begin 

sheltering when notified, buses are mobilized, and evacuation begins about 45 minutes after the 

GE is declared. It is coincidence that evacuation of the schools occurs at the same time that 

sirens are assumed to sound. The sounding of the sirens for GE is not directly linked to 

evacuation of the schools.

It is noted that the schools begin to evacuate in 1-hour, which is the same as in the LTSBO 

scenario. For the LTSBO, school buses are mobilized at the SAE and ready for deployment. 

Once the GE is declared, the evacuation begins within about 15 minutes because the resources 

have been prepared. For the STSBO, buses begin to be mobilized when the GE is declared, and 

the evacuation begins about 45 minutes later.

Cohort 4: 0 to 10 Special Facilities. Special facilities are assumed to depart at the same time as 

the evacuation Tail. 

Cohort 5: 0 to 10 Tail. The tail begins to evacuate approximately 4 hours and 15 minutes after 

notification to evacuate. 

Cohort 6: Non-Evacuating Public. This cohort group represents a portion of the 0 to 10 Public 

who may refuse to evacuate. It is assumed to be 0.5 percent of the population.

The delay to shelter identified in Table 6-5 represents a delay from the start of the accident, until

people enter the shelter, and the delay to evacuation represents the length of the sheltering period 

before initiating evacuation. These delays correspond to the different shielding factors that 

would be applied to each cohort during these timeframes. The speeds in this table represent 

average movements for the cohorts as derived from the ETEs. These speeds are adjusted within 

each grid element, where appropriate, when developing the WinMACCS model. 

Table 6-5 STSBO with RCIC blackstart Cohort Timing

Cohort

Delay to 

Shelter

DLTSHL 

(hr) 

Delay to 

Evacuation

DLTEVA 

(hr)

DURBEG 

(hr)

DURMID 

(hr)

ESPEED *

(early)

(mph)

ESPEED *

(mid)

(mph)

0 to 10 Public 1.00 1.00 0.25 3.00 5 3

10 to 20 Shadow 1.00 1.00 0.25 3.00 5 3

0 to 10 

Schools and Shadow 0.25 0.75 1.00 0.50 20 20

0 to 10

Special Facilities 1.00 4.25 0.50 0.50 3 20

0 to 10 Tail 1.00 4.25 0.50 0.50 3 20

Non-Evac 0 0 0 0 0 0

* 20 mph was used for the late-phase evacuation speed for all cohorts.
6-18

6.3.3 Short-Term Station Blackout without Reactor Core Isolation Cooling 

Blackstart

The timing of the emergency classification declaration for the STSBO without RCIC blackstart 

was based on Table PBAPS 3-1 in site emergency plan implementing procedures. The analysts 

reviewed the emergency classification timing with the licensee for accuracy. This scenario is an 

immediate GE. With loss of offsite power and loss of DC power, operators cannot determine 

whether water level is above TAF, and a GE is declared based on EAL MG1. The emergency 

response timeline for the STSBO without RCIC blackstart scenario is shown in Figure 6-6; 

protective action durations for each cohort are shown in Figure 6-7. Core damage, as evidenced 

by the first fuel cladding gap release, is calculated at 1 hour into the event, with a significant 

radioactive release from containment beginning 8 hours into the event, as indicated by the lower 

head failure in Figure 6-6.

Figure 6-6 STSBO without RCIC blackstart emergency response timeline
6-19

Figure 6-7 Duration of protective actions for STSBO without RCIC blackstart

The implementation of protective actions for the STSBO without RCIC blackstart is the same as 

implementation for the STSBO with RCIC blackstart.

Cohort 1: 0 to 10 Public. It is assumed to take 45 minutes for OROs to sound sirens following 

declaration of a GE, at which point the public is assumed to shelter. The time for the public to 

receive the warning and prepare to evacuate is assumed to be 1 hour after the siren sounds for the 

GE. 

 

Cohort 2: 10 to 20 Shadow. This cohort is assumed to begin movement at the same time as the 

0 to 10 Public cohort once widespread media broadcasts are underway. Residents in the 

10 to 20-mile area begin seeing large numbers of people evacuating and initiate a shadow 

evacuation. There is no warning or notification for the public residing in this area, and area is 

not under an evacuation order. 

Cohort 3: 0 to 10 Schools and 0 to 10 Shadow. This cohort is the first to take action. Upon 

receipt of the site’s declaration of a GE, county emergency management agencies would notify 

the schools in accordance with the emergency response plan. It is assumed that schools would 

begin sheltering when notified, buses would be mobilized, and evacuation would begin about 

45 minutes after the GE is declared. It is coincidence that this occurs at the same time that sirens 

are assumed to sound.

Cohort 4: 0 to 10 Special Facilities. Special facilities are assumed to depart at the same time as 

the evacuation tail. 

Cohort 5: 0 to 10 Tail. The tail begins to evacuate approximately 4 hours and 15 minutes after 

notification to evacuate. 

Cohort 6: Non-Evacuating Public. This cohort group represents a portion of the 0 to 10 Public 

who may refuse to evacuate. It is assumed to be 0.5 percent of the population.
6-20

The delay to shelter, identified in Table 6-6 represents a delay from the start of the event until

people enter the shelter. Delay to evacuation represents the length of the sheltering period before 

initiating evacuation. These delays correspond to the different shielding factors that would be 

applied to each cohort during these timeframes. The speeds in this table represent average 

movements for the cohorts. These values are adjusted within each grid element when developing 

the WinMACCS model. 

Table 6-6 STSBO without RCIC blackstart Cohort Timing.

Cohort

Delay to 

Shelter

DLTSHL 

(hr) 

Delay to 

Evacuation

DLTEVA 

(hr)

DURBEG 

(hr)

DURMID 

(hr)

ESPEED *

(early)

(mph)

ESPEED *

(mid)

(mph)

0 to 10 Public 1.00 1.00 0.25 3.00 5 3

10 to 20 Shadow 1.00 1.00 0.25 3.00 5 3

0 to 10 

Schools and Shadow 0.25 0.75 1.00 0.50 20 20

0 to 10

Special Facilities 1.00 4.25 0.50 0.50 3 20

0 to 10 Tail 1.00 4.25 0.50 0.50 3 20

Non-Evac 0 0 0 0 0 0

* 20 mph was used for the late phase evacuation speed for all cohorts.

6.4 Sensitivity Studies 

Analysis of emergency preparedness and response parameters, such as demographics, 

infrastructure, and timing, provides many opportunities for further evaluation through sensitivity 

studies. The project team selected three additional calculations to assess variations in the 

implementation of protective actions. Each of the sensitivity studies was conducted using the 

STSBO without RCIC blackstart scenario, which was selected because it represents an earlier 

release than the other scenarios.

• Sensitivity 1 – evacuation of a 16-mile area and a shadow evacuation from within the 

16 to 20 mile area. 

• Sensitivity 2 – evacuation of the 0 to 20-mile area.

• Sensitivity 3 – delay in implementation of protective actions for the public within the 

EPZ.

Sensitivity cases 1 and 2 assessed the effects of expanding the initial protective actions to 

distances of 16 and 20 miles respectively. The objective of these sensitivity analyses was to 

determine whether consequences might be reduced if the initial evacuation area were larger. 

Twenty miles was selected because it is twice the distance of the EPZ. A middle distance was 

also desired. The 16 mile distance was selected because a ‘ring’ had been established in the 

underlying nodalization network in WinMACCS. An existing ring was not available at 15 miles. 

Sensitivity 3 assessed a delay in the implementation of protective actions for the public. 

Although there is high confidence that the licensees and OROs will respond promptly and, 
6-21

follow their procedures, the SOARCA peer review committee suggested that a delay be 

considered. Such a delay might be caused by a delay in communication, a delay in the decision 

process (by licensee or OROs), or other factors. The 30-minute period was selected based on 

review of response data from exercises at Peach Bottom, which show that the response times and 

actions typically varied by only a few minutes among different exercises.

For the sensitivity analyses, the modeling of the area beyond the EPZ includes a full-scale 

evacuation, although this does not reflect likely protective action decisions. To support the 

assessment of implementing protective actions outside of the EPZ, an evacuation model was 

developed using data obtained for the 10 to 20 mile area around the nuclear power plant. 

Evacuation speeds for the cohorts in the 10 to 20 mile area were developed using OREMS 

Version 2.6. OREMS is a Windows based application used to simulate traffic flow that was 

designed specifically for emergency evacuation modeling [25]. The main features of OREMS

used in the analyses include the following:

• determining the length of time associated with complete or partial evacuation of the 

population at risk within an emergency zone, or for specific sections of the highway 

network or sub-zones

• determining potential congestion areas in terms of traffic operations within the 

emergency zone.

The OREMS model considers special conditions that may be imposed during an emergency 

evacuation. For example, intersections that normally have pre-timed controllers are assumed to 

be manned by emergency personnel to facilitate traffic flow. This function is consistent with the 

emergency response actions that would be implemented during an evacuation. Detail for road 

networks was obtained from available mapping and was input into OREMS using the standard 

intersection functions available in the model. Judgment and experience were applied in 

determining the number of nodes established for the model. OREMS can manage hundreds of 

nodes, but there is a point at which the addition of nodes and links provides little change in the 

total times. For an urban area, a nodal network would be heavily populated; for a rural area, the 

network would be lightly populated. The nodal network established for this analysis would be 

considered a moderately populated network for this code because the area is primarily rural 

mixed with some urban areas such as Lancaster, Pennsylvania.

For the Peach Bottom 10 to 20-mile ETE, 232,053 vehicles were loaded onto 118 nodes of a 

442 node network. The network loading was distributed over a 5-hour period to account for the 

trip generation time. The following evacuation times were produced from the OREMS 

calculation as plotted in Figure 6-8: 

• 100 percent evacuation: 19 hours;

• 90 percent evacuation: 12 hours and 15 minutes

These times were used to develop the evacuation speeds input into the WinMACCS model. The 

evacuation modeling conducted for the Peach Bottom plant was developed consistent with the 

characteristics observed in prior evacuations conducted for non-nuclear incidents. As described 

earlier, the analysis includes the common phenomenon of evacuations in which travelers who 
6-22

depart the threat zone the earliest experience shorter delays because the routes have yet to 

become fully used during the emergency. Evacuees who depart during the middle part of the 

evacuation, when the greatest numbers of people are seeking to depart, generally experience the 

highest congestion and longer delays because the demand on the roadway network is at its 

greatest, exceeding the available capacity in many areas. Evacuees who depart the hazard zone 

later enter the network as the demand nears, or goes below, the roadway capacity are able to 

avoid the delays associated with the peak evacuation demand period. The ETE modeling 

indicates that most congestion occurs in the more populous areas in the north near Lancaster, PA, 

and in the south near Forest Hill, Bel Air, and Fallston, MD. 

Figure 6-8 Evacuation timeline from Peach Bottom for the 10 to 20 mile region

The initial accident scenarios were evaluated for protective actions within the EPZ. Expanding 

the protective actions to distances beyond the EPZ is not readily accommodated using the 

modeling approach selected for these analyses. Therefore, although OROs may request that the 

10 to 20-mile population shelter, this population group is treated as performing normal activities 

throughout the emergency. The normal activity shielding factors are weighted averages of 

indoor and outdoor values based on being indoors 81 percent of the time and outdoors 19 percent 

of the time [32]. The hotspot and normal relocation model within MACCS2 will move affected 

individuals out of the area if the dose criteria apply.

6.4.1 Sensitivity 1 for the STSBO w/o RCIC Blackstart Evacuation to 16 Miles

For Sensitivity 1, evacuation of a 16-mile area around the nuclear power plant is assessed. A 

shadow evacuation is assumed to occur from within the 16 to 20-mile area, and the remaining 

members of the public in the 16 to 20-mile area were modeled as performing normal activities as 

Peach Bottom

0%

10%

20%

30%

40%

50%

60%

70%

80%

90%

100%

0:00 2:24 4:48 7:12 9:36 12:00 14:24 16:48 19:12

Elapsed Time

% Evacuated

12:15

19:00
6-23

described above. Figure 6-9 identifies the cohort timing for Sensitivity 1, and Figure 6-10 shows 

the durations of protective actions for each cohort.

Figure 6-9 Sensitivity 1 STSBO without RCIC blackstart - evacuation to 16 miles

Figure 6-10 Duration of protective actions for Sensitivity 1 STSBO w/o RCIC blackstart -

evacuation to 16 miles
6-24

Cohort 1: 0 to 16 Public. Following declaration of a GE, sirens are sounded and an EAS 

message is broadcast to the affected areas within the EPZ. The analysis assumes that the public

shelters when the sirens sound and that the time to receive the warning and prepare to evacuate is

1 hour. Another assumption in this sensitivity analysis is that the population from 10 to 16 miles

would be notified within the same timeframe as the EPZ via EAS messaging and route alerting. 

The ETE for the public was estimated as a linear projection between the Peach Bottom ETE 

study and the 10 to 20-mile ETE developed for the Sensitivity 2 analysis. Therefore, although the 

evacuation of the public starts at the same time as the base case, it takes longer to evacuate the 

area as indicated in Table 6-7, which shows longer travel times and slower speeds.

Cohort 2: 16 to 20 Shadow. This cohort is assumed to begin movement at the same time as the 

0 to 16 Public cohort once widespread media broadcasts are underway. Residents in the 

16 to 20-mile area begin seeing large numbers of people evacuating and initiate a shadow 

evacuation.

Cohort 3: 0 to 16 Schools and 0 to 16 Shadow. This cohort is the first to take action. Upon 

receipt of the site’s GE declaration, county emergency management agencies would notify the 

schools in accordance with the emergency response plan. It is assumed that schools begin 

sheltering in about 15 minutes and begin evacuating 45 minutes after GE. The sounding of 

sirens in response to the GE provides warning and notification to all residents and transients 

within the EPZ that there is an incident, and EAS messaging will request that people monitor the 

situation for further information. It is assumed that these actions cause a shadow evacuation 

from within the 0 to 16-mile area. 

Schools routinely practice fire drills and have call lists to quickly notify parents of an emergency. 

In the sensitivity study, it is assumed that the initial projections indicate a need for protective 

actions to a distance of 16 miles. It is assumed that within 1 hour, schools beyond the EPZ 

would notify parents and mobilize evacuation efforts.

Cohort 4: 0 to 16 Special Facilities. Special facilities are required to have evacuation plans, and 

this scenario assumes that the facilities within the 0 to 16-mile area would evacuate following 

these plans. The special facilities cohort is modeled to depart at the same time as the evacuation 

tail, although, in reality, facilities would be evacuating as resources become available. As shown 

in Table 6-7, the delay to evacuation is longer than the base case and the speed is slower.

Cohort 5: 0 to 16 Tail. An estimate of the departure for the evacuation tail is established as a 

linear projection between the Peach Bottom ETE and the OREMS 10 to 20-mile ETE developed 

for evacuation to a distance of 20 miles from the plant. 

Cohort 6: Non-Evacuating Public. This cohort group represents a portion of the public within 

the 0 to 16-mile area who may refuse to evacuate. It is assumed to be 0.5 percent of the 

population.

Table 6-7 identifies the cohort timing for Sensitivity 1.
6-25

Table 6-7 STSBO without RCIC blackstart, Sensitivity 1

Cohort

Delay to 

Shelter

DLTSHL 

(hr)

Delay to 

Evacuation

DLTEVA 

(hr)

DURBEG 

(hr)

DURMID 

(hr)

ESPEED*

(early)

(mph)

ESPEED*

(mid)

(mph)

0 to 16 Public 1.00 1.00 0.25 7.75 5 2

16 to 20 Shadow 1.00 1.00 0.25 7.75 5 2

0 to 16 

Schools and Shadow 0.25 0.75 1.00 0.50 20 20

0 to 16

Special Facilities 1.00 9.00 4.00 1.00 2 20

0 to 16 Tail 1.00 9.00 4.00 1.00 2 20

Non-Evac 0 0 0 0 0 0

* 20 mph was used for the late-phase evacuation speed for all cohorts.

6.4.2 Sensitivity 2 for the STSBO without RCIC Blackstart Evacuation to 20 Miles

Sensitivity 2 assesses the evacuation of a 20 mile area around the nuclear power plant. Because 

the initial evacuation is extended to 20 miles, no further shadow evacuation was considered. 

Table 6-8 identifies the cohort timing for Sensitivity 2. For this sensitivity case, the cohort 

timing and protective action durations are shown in Figure 6-11 and

Figure 6-12.

Figure 6-11 Sensitivity 2 STSBO without RCIC blackstart - evacuation to 20 miles
6-26

Figure 6-12 Duration of protective actions for sensitivity 2 STSBO w/o RCIC blackstart -

evacuation to 20 miles

Cohort 1: 0 to 10 Public. Following declaration of a GE, sirens would be sounded within the 

EPZ and an evacuation order would be issued for the EPZ. The time for this cohort to receive 

the warning and prepare to evacuate is assumed to be 1 hour. Although the evacuation of the 

public starts at the same time as in the base case, it takes longer to evacuate the area as indicated 

in Table 6-8 which shows longer travel times and slower average speeds. Speeds were adjusted 

using consequence model speed adjustment factors ranging from 75 percent to 150 percent of the 

average speed, as appropriate for the rural and urban areas. 

Cohort 2: 10 to 20 Public. This sensitivity analysis considers that the 10 to 20 Public cohort 

would be notified at the same time as the EPZ via EAS messaging and route alerting. The time 

to receive the warning and prepare to evacuate is still assumed to be 1 hour after the initial 

notification. The ETE for the 10 to 20 Public cohort was calculated using OREMS.

Cohort 3: 0 to 20 Schools. Upon receipt of the site’s declaration of GE, county emergency 

management agencies would notify the schools within the EPZ in accordance with the 

emergency response plan. This sensitivity study assumed that schools beyond the EPZ would 

decide, based upon media information that it is prudent to evacuate or close schools immediately.

It is assumed that within 1 hour, schools beyond the EPZ would notify parents and mobilize 

evacuation efforts. 

Cohort 4: 0 to 20 Special Facilities. This sensitivity study assumed that special facilities beyond 

the EPZ would decide, based on media information that it is prudent to evacuate. The 

0 to 20 Special Facilities cohort is modeled to depart at the same time as the evacuation tail. As 

shown in Table 6-8, the delay to evacuation is longer than in the base case and the speed is 

slower. 

Cohort 5: 0 to 20 Tail. The ETE for the evacuation tail was estimated based on the OREMS 

analysis. This cohort shelters upon hearing the sirens and begins evacuating 12 hours and 

15 minutes later. As shown in Table 6-8, the delay to evacuation is longer than in the base case 

and the speed is slower.
6-27

Cohort 6: Non-Evacuating Public. This cohort group represents a portion of the public within 

the 0 to 20-mile area who may refuse to evacuate and is assumed to be 0.5 percent of the 

population.

Table 6-8 identifies the cohort timing for Sensitivity 2.

Table 6-8 STSBO without RCIC blackstart, Sensitivity 2.

Cohort

Cohort 

Number#

Delay to 

Shelter

DLTSHL 

(hr) 

Delay to 

Evacuation

DLTEVA 

(hr)

DURBEG 

(hr)

DURMID 

(hr)

ESPEED*

(early)

(mph)

ESPEED*

(mid)

(mph)

0 to 10 

Public 1 1.00 1.00 0.25 11.00 5 1.8

10 to 20 

Public 2 1.00 1.00 0.25 11.00 5 1.8

0 to 20 

Schools 3 0.25 0.75 1.0 0.5 20 20

0 to 20 Special 

Facilities 4 1.00 12.25 5.75 1.00 1.8 20

0 to 20 Tail 5 1.00 12.25 5.75 1.00 1.8 20

Non-Evac 6 0 0 0 0 0 0

* 20 mph was used for the late-phase evacuation speed for all cohorts.

6.4.3 Sensitivity 3 for STSBO w/o RCIC Blackstart - Delay in Protective Actions

At the initiation event for the STSBO without RCIC blackstart, a GE is declared based on 

EAL MG1. Although there is a high level of confidence about the actions expected from control 

room operators and the ORO response, the peer review committee suggested that the analysis 

consider a delay in the implementation of protective actions. Such a delay could be from 

delayed communication to the public about implementation of protective actions, or other 

reasons. To address the potential for delay, an additional protective active timeline has been 

developed for the STSBO without RCIC blackstart. This timeline reflects a delay in the 

implementation of protective actions by the public within the EPZ. Because protocols and 

procedures are in place, exercised, and tested frequently, it is assumed that a delay of 30 minutes 

is adequate for this sensitivity study. The cohort timing and protective action durations are 

shown in Figure 6-13 and Figure 6-14 for this sensitivity case.
6-28

`

Figure 6-13 Sensitivity 3 STSBO w/o RCIC blackstart - delay in protective actions

Figure 6-14 Protective action durations for sensitivity 3 STSBO w/o RCIC blackstart -

delay in protective actions

The 30 minute delay was added to the ‘Delay to Shelter’ parameter for all cohorts. The 

remaining delay and speed parameters were unchanged from the base case. With the delay 

allocated to this point, all cohort actions move to the right on the timeline by 30 minutes and the 

sum of DLTEVA, DURBEG, and DURMID still equates to the ETE for the public and the tail.
6-29

Table 6-9 identifies the cohort timing for Sensitivity 3.

Table 6-9 Cohort Timing for Sensitivity 3

Cohort

Delay to 

Shelter

DLTSHL 

(hr)

Delay to 

Evacuation

DLTEVA 

(hr)

DURBEG 

(hr)

DURMID 

(hr)

ESPEED*

(early)

(mph)

ESPEED* 

(mid)

(mph)

0 to 10 Public 1.50 1.00 0.25 3.00 5 3

10 to 20 Shadow 1.50 1.00 0.25 3.00 5 3

0 to 10 

Schools/Shadow 0.75 0.75 1.00 0.50 20 20

0 to 10

Special Facilities 1.50 4.25 0.50 0.50 3 20

0 to 10 Tail 1.50 4.25 0.50 0.50 3 20

Non-Evac 0 0 0 0 0 0

* 20 mph was used for the late phase evacuation speed for all cohorts.

6.5 Analysis of Earthquake Impact

A seismic analysis was developed to assess the potential effects on local infrastructure 

(e.g., roadways and bridges), communications, and emergency response in the event of a large 

scale earthquake. The accident scenario used in the earthquake analysis is the STSBO without 

RCIC blackstart, which was selected because this scenario represents an earlier release than the 

other scenarios. Integrating the effects of the earthquake into the analysis required assessing the 

damage potential of the earthquake, identifying parameters that would be affected, and 

determining the new values for affected parameters. 

The potential for an earthquake is largely identified by the occurrence of previous earthquakes in 

the region. Understanding of where earthquake faults exist in the eastern United States is not 

robust, unlike in the West, where geological fault lines can be identified on the surface. Faults in 

the east are usually buried below layers of soil and rock and not identifiable, making prediction 

of earthquake location and magnitude difficult. The earthquakes hypothesized in SOARCA are 

assumed to be close to the plant site, and it may be assumed that severe damage is largely

localized. Housing stock would mostly survive the earthquake, with some damage. The local 

electrical grid is assumed to be out of service from the failure of lines, switch yard equipment, or 

other impacts. As there is currently no backup power system for the sirens at Peach Bottom, the 

analysis assumed that OROs would perform route alerting to notify the population of the need to 

take protective actions. This is a routine and effective method of notifying the public to 

implement protective actions [22]. Under these postulated conditions, the analysis considered 

the potential for such an earthquake to affect emergency response and public evacuation. 

6.5.1 Soils Review

To approximate the extent of damage, NRC seismic experts conducted an evaluation of the 

potential failure of infrastructure to determine which, if any, roadways or bridges may fail under 

the postulated earthquake conditions. The assessment was performed using readily available 

information and professional judgment. Existing information on the basic bedrock geology of 

the region was developed from reports and papers from the United States Geological Service, 

Pennsylvania Geological Survey, Maryland Geological Survey, and the final safety analysis
6-30

report for the Peach Bottom plant. Generalized soils information was developed from Natural 

Resources Conservation Service (NRCS) soil survey information for York and Lancaster 

counties in Pennsylvania. This analysis assumed that the generalized soil characteristics are 

applicable to the entire region.

The NRCS reports break the soils into several distinct, descriptive units. The units of interest to 

the present evaluation are the Chester-Gleneig, Mt. Airy-Manor, Grenville, and Codorus soil 

groups. These are typically well-drained soils, with the Chester-Gleneig and Mt. Airy-Manor 

units (mostly residuum from saprolite) existing on the ridges and uplands and the Glenville and 

Codorus soils (mostly alluvium and colluvium) in the low regions and valley bottoms. Based on 

the engineering properties contained in the NRCS reports, these units of interest would be either 

‘potentially liquefiable’ or ‘liquefiable’ if the water content in the soils were sufficiently high. 

Initial assumptions for the analysis included the following: (1) the general soil characteristics 

described above exist at all locations at the time of any large earthquake and the water table is 

sufficiently high that liquefaction and loss of strength would result, and (2) liquefaction of soils

beneath a roadway in flat topography would not result in any significant damage or otherwise 

compromise the evacuation route.

The region around Peach Bottom is generally flat to rolling topography with a relatively small 

number of streams and watercourses, resulting in few bridges and overpasses. The general 

region of interest near the Peach Bottom site does not have a large number of locations where 

earthquake damage would render the evacuation routes nonfunctional. Information is not readily 

available on the specific engineered features of bridges and other infrastructure with which to 

make specific assessments on the likelihood of failure in an earthquake scenario. Therefore, the 

analysis assumes that: (1) all of the bridges across the Susquehanna River fail within 20 miles of 

the plant, and (2) the road across the Conowingo Dam would be unavailable. 

6.5.2 Infrastructure Analysis

The seismic evaluation of the potential failure of roadway infrastructure identified 12 bridges 

and roadway segments that could fail under the postulated conditions. Table 6-10 provides a 

brief description of each area assumed to fail, and Figure 6-15 shows the transportation network 

and the locations of the affected roadway segments and bridges.

Table 6-10 Description of the Potential Evacuation Failure Locations

Location Description

A PA Highway 372 upstream of Susquehanna River. Two single span bridges along a 

single roadway segment. 

B PA 372 bridge across the Susquehanna River north of the plant. Two lane multi-span 

bridge.

C US 222 Robert Fulton Highway. Two lane single span bridge.

D PA 74 (Delta- bypass/Pylesville Road) south of Holtwood Road (PA 372). Single span 

two-lane bridge.

E MD 136 - Whiteford Road. Two lane segment along lakeside. Potential for slumping 

into lake.

F US 1 (Conowingo Road) east of Susquehanna River and US 222. Two lane single-span 

bridge.

G US 1 across Conowingo Dam. Two lane road.
6-31

H US 1 west of Susquehanna River, west of MD 136. Two lane road, single span bridge.

I MD 136 south of US 1. Two lane road, single span bridge.

J MD 222 Susquehanna River Road north of Main Street. Two lane road runs along river 

edge. Potential for slumping into river.

K I-95. Six lane multi span bridge.

L US 40. Four lane multi span bridge across Susquehanna River.

Figure 6-15 Roadway network identifying potentially affected roadways and bridges

Evacuations are planned and conducted to move the public radially away from the nuclear power 

plants. Evacuation routes are designated in emergency planning brochures, but all roadways 

within the EPZ serve the evacuees. Some of the bridges identified in the seismic analysis will 

have a negligible effect on the evacuation because of their location within the roadway network, 

while others may have a more pronounced effect. 

The loss of bridges and the Conowingo Dam Road crossing the Susquehanna River does not 

affect the ETE. These crossings are represented as B, G, K, and L on Figure 6-15. The EPZ 

evacuation routes identified in the emergency plan indicate that evacuees west of the river would 
6-32

typically evacuate in a westerly or southerly direction, and evacuees east of the river would 

evacuate in a northerly or easterly direction. The only river crossing (i.e., location G) identified 

as a Peach Bottom evacuation route is U.S. Route-1 (US 1) across the Conowingo Dam. 

Although this road is an evacuation route, the evacuation map indicates that travel west of the 

river would proceed westerly, and travel east of the river would proceed easterly; therefore, 

failure of the river crossing would have no effect on the ETE. 

Bridges A, C, D, and E in Figure 6-15, serve sparsely populated areas and have additional roads 

available, supporting a conclusion that evacuation delay because of loss of these bridges is 

minimal. Along Whiteford Road (i.e., location E in Figure 6-15) there is a potential for part of 

the roadway to slump into the lake. This is the only roadway failure in this area and there are 

many alternative routes north, west, and south of the location; therefore, no appreciable delay 

would be expected due to this failure. Travel along Maryland Route 136 (MD 136) may be

diverted to multiple alternative roadways in the event the single span bridge I fails.

Bridges H and I in Figure 6-15 are two lane single span bridges located south of the plant in an 

area where local roadways are available as alternate routes. Pennsylvania Route 222 (PA 222) 

becomes MD 222 at the State line; and the MD 222 part may potentially slump off into the river 

at location J in Figure 6-15. As indicated on the map, there are alternatives to route around this 

area if the roadway does slump into the river. Failure of these bridges should not appreciably 

affect the ETE.

The two bridges with the greatest potential to affect the ETE are F and H because they are 

located along the edge of the EPZ and serve larger areas. The total population identified in the 

ETE report for the area served by bridge H is about 18,000 corresponding to approximately 

6,000 passenger cars using the vehicle population factor applied in the ETE report. For this 

volume of vehicles, only about 1,200 vehicles per hour would need to exit the EPZ to stay within 

the 5 hour and 15 minute ETE. For this type of two lane roadway, a service volume of 

1,700 passenger cars per hour may be achieved [30]. The alternate routes out of the EPZ have 

more than sufficient capacity to support the evacuating population. 

The total population identified in the ETE report for the area served by bridge F on U.S. 1 is 

about 6,790, which equates to approximately 2,300 passenger cars. The Susquehanna River 

Road (PA 222) and other local roadways nearby are available as alternate routes out of the EPZ; 

therefore, no appreciable delay would be expected. 

Based on a review of the ETE report, the EPZ subareas affected by loss of bridges, and a detailed 

review of the roadway network, the analysts concluded that loss of the identified bridges will not 

increase the total ETE. This is consistent with the Peach Bottom ETE, report which shows that 

evacuation of the northeast quadrant of the EPZ controls the evacuation time for the entire EPZ. 

Only bridge B is located in this area, crossing the river as described above. Figure 6-16 shows an 

example of a bridge (i.e., Bridge C on US 222 Robert Fulton Highway) that could potentially fail 

under the earthquake conditions.
6-33

Figure 6-16 Bridge along Robert Fulton Highway

6.5.3 Electrical Power and Communications

The seismic event causes the loss of all onsite and offsite power, which can affect the response 

timing and actions of the public. Typically, sirens would sound following the declaration of an 

SAE and GE. The loss of power will affect the number of sirens that sound; however, it is 

expected that many of the sirens within the EPZ will still function. Where sirens are inoperable, 

the initial alert and notification of the public may take longer. The loss of power affects the 

number of residents receiving instructions via EAS messaging. Televisions, household radios, 

and some telephones will not operate, although battery-operated radios and car radios will 

operate. The residents within the EPZ will have felt the earthquake which will instill a sense of 

emergency. It may be expected that the residents will use multiple methods of communication, 

such as cell phones, telephones, websites, and direct interface to communicate the emergency 

message. The alert and notification will be supplemented by route alerting, which is a planned 

backup form of communication for the EPZ.

The loss of power will cause traffic signals to default to a four-way stop mode, which is less 

efficient than normal signalization [30]. Typically, emergency response personnel would 

respond to these intersections and direct traffic. A review of the roadway network within the 

EPZ indicates that there are only a few traffic signals within the EPZ and that most intersections 

are controlled with stop signs. Table 7-1, “Recommended Traffic Control Management 

Locations,” in the ETE report identified 12 key locations for emergency management traffic 

control to expedite traffic out of the EPZ. As indicated in the ETE report, these 12 locations are 

included in the county plans. The analysis assumes that the OROs will be able to provide the 

staff needed to support the few locations where traffic signals are not working; therefore, the 

analysis also concludes that the loss of signalization will not increase the total ETE.
6-34

6.5.4 Emergency Response

The analysis assumed that event timing is a mid-week winter day when residents are at work and 

children are at school. In Maryland, the primary shift of emergency responders would be on duty 

and immediately available at the time of the incident. Most of the Pennsylvania emergency 

responders are volunteers and would be at their normal place of employment. After the 

earthquake, there will be an initial need to assess damage and respond to life-threatening needs. 

These initial priorities for emergency response personnel may delay implementation of traffic 

control to support an evacuation. It is assumed that responders will realize early that damage to 

local infrastructure is not severe, because there are very few bridges or large structures within the 

EPZ, and will focus efforts on communicating with the public via route alerting, where 

necessary. Route alerting would not be appreciably delayed because damage to local 

infrastructure is not severe. 

During large scale emergencies, OROs routinely supplement staff with on call and off duty

personnel. Although communications are assumed to be initially limited, radios are available to 

contact the needed staff, and off duty responders may be expected to report for duty during such 

emergencies, particularly if they have experienced the earthquake. It is expected by the time an 

evacuation is ordered, OROs would have been augmented with additional staff. 

6.5.4.1 Evacuation Time Estimate

Roadway capacity is a key parameter in developing ETEs. Reducing the number of roadways 

will reduce the available capacity. The evacuation times can be affected when bridges fail, traffic 

signals do not operate effectively, and EAS messaging is not disseminated in a timely manner to 

inform evacuees of protective actions and preferred evacuation routes. Although there are a 

number of factors that can increase evacuation time, the effect on the ETE is expected to be 

limited. This is because residents will have experienced the magnitude of the earthquake, 

therefore they have some level of understanding that an emergency exists, prior to any formal 

warning. Secondly, emergency response personnel would begin route alerting and establish 

traffic control.

The roadway network beyond the EPZ was also evaluated to determine if loss of infrastructure 

might delay evacuees traveling through this area. How this loss of infrastructure affects the 

evacuation time depends on the location of the facilities and the evacuating public who may be 

expected to use these routes. As described earlier, loss of the bridges and roadways identified in

Table 6-10 are not expected to appreciably affect the ETE. This is because only 12 locations in 

the 20 mile radius around the plant, (1,256 square miles) were identified as potentially failing. 

Within such a large area, there are many alternate routes for evacuation.

Another consideration in this analysis is that no large-scale evacuation is expected for the 

10 to 20 mile area. Therefore, the roadways are assumed to be substantially available to serve 

the evacuating public from the EPZ. Furthermore, almost all of the bridges and the roadway 

sections that are assumed to fail are located in the southern section of the 10 to 20 mile area 

beyond the EPZ. This area has a smaller population than the area to the north of the plant. The 

results of the OREMS analysis used to develop the ETE for the 10 to 20 mile area demonstrated 

that longer evacuation times occurred in the northern section because of congestion experienced 

near Lancaster, Pennsylvania. 
6-35

The ETE was used to develop the speeds for the evacuating public. The following ETEs 

provided for Peach Bottom were used for the base case analyses: 

• 100 percent evacuation: 5 hours and 15 minutes

• 90 percent evacuation: 4 hours and 15 minutes

For the seismic analysis, it was assumed that because the earthquake is felt by the public, the 

0 to 10 Shadow cohort evacuation increased from the 20 percent used in the base case to 30 

percent of the population. This effectively removes 30 percent of the total vehicles from the 

roadway network shortly after the incident, which reduces potential for traffic congestion. The 

wide availability of roadway infrastructure at these distances from the plant provides ample 

access for evacuees leaving the EPZ and should not appreciably affect evacuation times.

6.5.5 Development of WinMACCS Parameters 

Modeling the effects of seismic events required adjusting additional parameters. To account for 

the potential loss of bridges and roadway sections, the routing patterns in the WinMACCS model 

were adjusted to divert traffic around these locations by using routes not impacted by damaged 

bridges. This adjustment was completed for each grid element with an impacted roadway. 

Routing was manually adjusted to travel around rather than through the impacted grid elements. 

The relocation parameters used in the earthquake analysis are the same as those used in the base 

case analyses. Shielding factors are also the same as those used in the base case analyses. It 

may be expected that the damage to structures caused by an earthquake of this magnitude would 

include broken windows and some structural damage. However, because residents within the 

seismic area are assumed to shelter only for a short period of time, any adverse affect on 

sheltering protection factors is irrelevant because evacuation begins before the plume arrives. 

6.5.6 Seismic Short-Term Station Blackout without RCIC Blackstart

The timing of emergency classification declarations for the STSBO without RCIC blackstart was 

based on the EALs contained in the site emergency plan implementing procedures. The timing 

of emergency classification declarations for the STSBO without RCIC blackstart was based on 

Table PBAPS 3-1 in the site emergency plan implementing procedures. The analysts reviewed 

emergency classification timing with the licensee for accuracy, and this scenario was identified 

as an immediate GE. With loss of offsite power and loss of DC power, operators cannot 

determine whether water level is above TAF, and a GE is declared based on EAL MG1. The 

emergency response timeline for the STSBO without RCIC blackstart scenario is shown in 

Figure 6-17. The duration of specific protective actions for each cohort is shown in Figure 6-18. 

Core damage, as evidenced by the first fuel cladding gap release, is calculated at 1 hour into the 

event, with a significant radioactive release from containment beginning 8 hours into the event as 

indicated by the containment failure.
6-36

Figure 6-17 STSBO w/o RCIC blackstart emergency response timeline (seismic analysis)

Figure 6-18 Protective action durations - STSBO w/o RCIC blackstart (seismic analysis)

The timeline identifies points at which cohorts would receive instruction from OROs to 

implement protective actions. Cohorts would then implement the protective actions. While 

protective actions within the EPZ can be modeled in accordance with procedures, the analysts 

made assumptions that approximate those actions that could be taken due to the effects of the 

earthquake. The evacuation is assumed to include the full EPZ, which is consistent with 
6-37

emergency planning in Pennsylvania. For this analysis, a full evacuation was modeled assuming 

that the State of Maryland OROs would agree with the Pennsylvania protective action decisions. 

The analysis assumed that the large earthquake will be felt by everyone within the EPZ, instilling 

a heightened preparedness. It is assumed that the public is ready to respond to protective actions 

once they receive information, and that some individuals will begin to prepare for an evacuation 

before receiving official notice. 

Cohort 1: 0 to 10 Public. The 0 to 10 Public cohort is assumed to begin evacuating upon receipt 

of notification which is provided primarily by route alerting. The analysis assumes that the 

effects of the earthquake are so severe that members of the public, knowing they live within an 

EPZ, begin preparations for evacuation shortly after the earthquake. 

Cohort 2: 10 to 20 Shadow. This cohort is assumed to begin movement at the same time as the 

0 to 10 Public cohort once widespread media broadcasts are underway. The analysis assumes 

that the shadow population increases to 30 percent (from 20 percent for the base case 

calculations) of the public in the area beyond the EPZ. 

Cohort 3: 0 to 10 Schools. Although communication systems may have been affected, this 

analysis assumes that, after receipt of the GE declaration, county emergency management 

agencies notify the schools promptly. The analysis also assumes that, having felt the earthquake, 

schools take the initiative to prepare to evacuate as they would for an emergency. Buses would 

be mobilized, and it is assumed that schools would begin evacuating about 1 hour after the start 

of the incident. The limited effect on infrastructure within the EPZ is not expected to 

appreciably delay bus mobilization. The analysis also assumed that, given the magnitude of the 

earthquake, parents in the vicinity of the schools would pick up their children, reducing the need 

for a full complement of buses. 

Cohort 4: 0 to 10 Shadow. This cohort is assumed to begin movement first. They experience the 

earthquake and quickly begin to evacuate, avoiding the traffic congestion.

Cohort 5: 0 to 10 Special Facilities. The special facilities cohort is assumed to depart at the 

same time as the evacuation tail. Inbound lanes on roadways will be useable for emergency 

support vehicles, but localized congestion will delay the arrival of specialized vehicles. Special 

facilities are assumed to leave at the same time as the evacuation tail; however, as discussed 

earlier for other scenarios, this is a simplification of the analysis because special facilities would 

realistically evacuate individually as resources are available. 

Cohort 6: 0 to 10 Tail. The tail takes longer to evacuate for many reasons, such as the need to 

return home from work to evacuate with the family; the need to shut down farming or 

manufacturing operations before evacuating, and for the earthquake, the need to move rubble or 

other items before evacuating. 

Cohort 7: Non-Evacuating Public. This cohort represents the portion of the 0 to 10-mile Public

who may refuse to evacuate. It is assumed to be 0.5 percent of the population.
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Table 6-11 summarizes the evacuation timing for each cohort. In general, the cohorts in the 

seismic study have faster mobilization times because of their immediate awareness and have the 

same evacuation speeds. 

Table 6-11 Cohort Timing for STSBO without RCIC blackstart

Cohort

Delay to 

Shelter

DLTSHL 

(hr)

Delay to 

Evacuation

DLTEVA 

(hr)

DURBEG

(hr)

DURMID

(hr)

ESPEED *

(early)

(mph)

ESPEED *

(mid)

(mph)

0 to 10 Public 1.00 0.00 0.25 4.00 5 3

10 to 20 

Shadow 1.00 0.00 0.25 4.00 5 3

0 to 10 

Schools 0.25 0.75 0.25 4.00 5 3

0 to 10 

Shadow 0.50 0.00 1.00 0.50 20 20

0 to 10

Special 

Facilities 1.00 4.25 0.50 0.50 3 20

0 to 10 Tail 1.00 4.25 0.50 0.50 3 20

Non-Evac 0 0 0 0 0 0

* 20 mph was used for the late-phase evacuation speed for all cohorts.

6.6 Accident Response and Mitigation of Source Terms

The Peach Bottom SOARCA study has concluded that scenarios can be mitigated by the licensee 

through the use of safety and security enhancements, including SAMGs and 10CFR50.54(hh) 

mitigation measures. Analyses were conducted of the consequences that may result if the onsite 

emergency response organization (ERO) takes insufficient mitigative action to prevent core 

damage and radiological release to the environment. It is expected that mitigative actions would 

be attempted. The staff believed it appropriate to perform the consequence analyses to develop 

an understanding of core melt sequences, source term evolution, and offsite response dynamics

and to compare this to previous studies. However, neither a human reliability assessment nor a 

detailed seismic damage assessment for implementation of mitigative measures were performed. 

The following analysis describes an expected national level response to a severe nuclear power 

plant accident and provides a basis for truncating the release no later than 48 hours after the 

accident begins. Note that past studies, including PRA such as NUREG-1150, typically truncated 

releases after 24 hours.

Mitigative actions during an accident are intended to:

• prevent the accident from progressing;

• terminate core damage if it begins;

• maintain the integrity of the containment as long as possible; and

• minimize the effects of offsite releases.

Response to a General Emergency would begin with the onsite ERO and would expand as 

needed to include utility corporate resources, State and local resources, and resources available 

from the Federal government, should these be necessary. It is most likely that plant personnel 
6-39

would attempt to mitigate the accident before core melt, but if their efforts were unsuccessful the 

national level response would provide resources to support mitigation of the source term. This

discussion presents a timeline for bringing resources onto the Peach Bottom site to support 

mitigative actions, including as a last effort, flooding the reactor building to a level above a 

hypothetical hole in containment to minimize the effects of an offsite release. The approach 

described herein also discusses, to the extent practical at this time, the NRC Task Force review 

of the response to the nuclear power plant accident at the Fukushima Diaichi facility in Japan. 

On March 11, 2011, the Great East Japan Earthquake caused a large tsunami estimated to have 

exceeded 14 meters (45 feet) in height at the Fukushima Daiichi nuclear power plant site. The 

earthquake and tsunami produced widespread devastation across northeastern Japan, resulting in 

approximately 25,000 people dead or missing, displacing tens of thousands of people, and 

significantly impacting the infrastructure and industry in the northeastern coastal areas of Japan. 

The earthquake and tsunami caused accidents at Units 1, 2 and 3 of the Fukushima Daiichi 

nuclear power plant facility and caused concern for the remaining units and spent fuel pools at 

the site. Amid the vast devastation and competing health and safety priorities in the region, 

onsite and offsite response agencies worked diligently to bring the accidents under control. This 

effort took many weeks. 

Shortly after the nuclear accident, the NRC established a task force to conduct a methodical and 

systematic review of the NRC’s processes and regulations to determine whether the agency 

should make additional improvements to its regulatory system. The Task Force report, 

“Recommendations for Enhancing Reactor Safety in the 21st Century,” [35] identified that 

prolonged SBO and multiunit events present challenges to Emergency Preparedness facilities 

that were not considered when the NRC issued NUREG–0696, “Functional Criteria for 

Emergency Response Facilities,” in 1981. The Task Force report also states that an overarching 

lesson is that major damage to infrastructure in the area surrounding the plant might challenge an 

effective emergency response. A number of recommendations are presented in the report that 

address physical, administrative, and regulatory enhancements to further reduce the risk of 

similar challenges occurring among the US fleet of nuclear power plants. As a state-of-art 

analyses, the SOARCA project included a degree of depth in the analyses beyond the scope of 

many previous studies. In this regard, some of the recommendations of the Task Force report 

had already been considered in SOARCA.

For instance, SOARCA investigated the challenges of potential damage to infrastructure within 

and beyond the EPZ as a result of an earthquake. The site specific analysis showed that 

12 bridges may fail. Four of these bridges were located beyond the EPZ and the remaining eight 

were dispersed within the 10 mile EPZ (314 square miles). The analysis showed that because 

there were relatively few bridge failures within such a large area, use of the roadways would be 

relatively unimpeded. The seismic analysis described in Section 6.5 quantified the offsite effects 

of challenges to both infrastructure and resources within the Peach Bottom EPZ. The Task Force 

report also recommends further enhancement of current capabilities for onsite emergency actions 

by requiring that licensee’s modify the EOP technical guidelines to include SAMGs. This would 

enhance current capabilities, but would not change the manner in which SAMGs have already 

been considered in SOARCA.
6-40

With regard to the emergency response, the Peach Bottom analyses applied the emergency 

classification scheme identified in the site EAL matrix. For Peach Bottom the declaration of a 

GE would occur at 15 minutes for the STSBO scenarios. For the LTSBO scenario, declaration of 

an SAE occurs at 15 minutes and declaration of a GE would occur about 30 minutes later. 

Licensees are required by regulation to notify OROs within 15 minutes of declaring an 

emergency, and the OROs then initiate a planned response by offsite agencies who are able to 

direct necessary resources upon request, such as fire trucks, to support mitigation of the accident. 

The declaration by the licensee is not only a notification, this declaration initiates an ongoing 

communication between the control room, licensee staff, OROs, NRC, and other response 

agencies. 

As supported by the SOARCA analyses, it is shown that the accidents evaluated could be 

mitigated through the actions of the onsite and offsite response agencies. The evaluation of the 

mitigation of source term and truncation of the accident at 48 hours further expands upon the 

response resources through identification of corporate, local, State, and Federal offsite resources. 

The responsibilities and resources of each of these organizations are described in onsite and 

offsite emergency response plans. These response organizations would mobilize upon request 

and as needed to support a severe nuclear power plant accident. These resources are in addition 

to the mitigative actions by the licensee through the use of safety and security enhancements, 

including SAMGs and 10CFR50.54(hh) mitigation measures. 

Although the response to the Fukushima Daiichi accident has taken weeks and challenges still 

remain, it is expected that the regulatory structure, protocols, and resources available to support a 

response in the US are sufficient to mitigate the release of the accident scenarios identified in this 

study within 48 hours. Implementation of recommendations from the Task Force report would 

further improve the capabilities to mitigate the accident itself in a timely manner to help prevent 

a release from occurring.

For purposes of this analysis, the timeline begins when the plant has declared an emergency. 

The State would be notified by the plant within 15 minutes after the emergency declaration and 

the plant would also notify their corporate headquarters. The NRC would be notified, following 

plant procedures, within one hour of the declaration; although, for a seismic event, it may be 

assumed that the NRC would be informed sooner than one hour by the onsite NRC resident 

inspectors, if not the licensee. The NRC resident inspectors are available to provide knowledge 

of accident conditions and would recognize the urgency of the situation should an earthquake 

occur. There are multiple onsite communications systems available to make these notifications 

and the communication paths are exercised and tested. The NRC headquarters would notify 

other Federal agencies as appropriate. 

Upon receipt of notification of the emergency classification, the NRC would activate the 

Headquarters Operations Center (HOC). NRC response teams reporting to the HOC include the 

Reactor Safety Team, Protective Measures Team, Executive Team, and other teams that support 

response related activities. Plant drawings and procedures are available in the HOC. Data that is 

typically communicated to the HOC via the Emergency Response Data System (ERDS), the 

Emergency Notification System bridge line, and other communication bridges, would likely be 

communicated via the resident inspectors satellite phones and via site battery operated systems. 
6-41

Licensees are required to provide guaranteed power to the emergency communications 

equipment per NRC Bulletin 80-15, “Possible Loss of Emergency Notification System (ENS) 

with Loss of Offsite Power.”The NRC Region 1 office would activate, with similar response 

teams, and would mobilize and deploy a Site Team to the licensee’s response facilities to 

independently review response actions. The Site Team would include reactor safety engineers to 

review actions taken to mitigate the accident and protective measures specialists to review 

protective action recommendations to assure that the most appropriate actions are taken. Arrival 

of the Site Team may take several hours. The HOC, Regional Operations Center, and Site Team 

include liaisons to support coordination of resources when requested by the licensee.

Once notified, the plant corporate headquarters would activate the corporate response resources. 

Peach Bottom is part of the Exelon fleet, which includes a remote EOF that would be activated. 

Exelon has access to fleet-wide emergency response personnel and equipment, including 

equipment from sister plants, following 10CFR50.54(hh) reactor security requirements to 

mitigate the effects of large fires and explosions. This equipment would support multiple 

response needs and include such items as generators, pumps, compressed gas, etc. In addition, 

the Institute for Nuclear Power Operations (INPO) and the Nuclear Energy Institute (NEI) would 

activate their emergency response centers to assist the site. These organizations would make 

available additional knowledgeable personnel and an extensive array of equipment. 

Concurrent with the NRC and industry response, the National Response Framework (NRF) 

which establishes a coordinated response of national assets would be implemented. As described 

in the Nuclear/Radiological Incident Annex to the NRF, NRC is typically the Coordinating 

Agency for incidents occurring at NRC-licensed facilities. As Coordinating Agency, NRC has 

technical leadership for the Federal Government’s response to the incident. Under established 

agreement with the NRC, the Department of Homeland Security (DHS) would be the 

Coordinating Agency56 for an event in which a GE is declared. In this case, NRC retains the 

Federal technical leadership role but does not coordinate overall Federal response. Under the 

NRF, the DHS would likely deploy the radiological monitoring and radiological response assets 

of the Department of Energy (DOE) for this type of an emergency. Some of the other agencies 

cooperating in an incident include the EPA, the Federal Emergency Management Agency, the 

U.S. Department of Health and Human Services, and any other federal agency that may be 

needed. 

The above organizations would be developing onsite and offsite mitigation strategies with 

different objectives. These strategies would be implemented concurrently. An onsite mitigation 

strategy relies upon onsite resources and is expected to be immediate in order to prevent core 

melt and radiological release. As mentioned earlier, for the unmitigated cases the SOARCA 

project assumes onsite mitigative efforts are not effective. Offsite mitigation strategies utilize 

onsite and offsite equipment and resources and take more time to develop. 

 

56 The coordinating agency supports the U.S. Department of Homeland Security incident management 

mission by providing the leadership, expertise, and authorities to implement critical and specific aspects of 

the response. 
6-42

6.6.1 External Resources

The initiating event for the reactor accident is a beyond design-basis earthquake close to the 

plant. A General Emergency is declared shortly after the event, and the DHS would be the 

Coordinating Agency bringing to bear any needed Federal resources. The State remains 

responsible for protecting the public health and safety and remains directly involved in the 

response. State, local and Federal resources would be available for use in mitigating the 

accident. This review describes how resources within the region might be acquired and used to 

support accident mitigation. 

The States and counties have road departments with access to heavy equipment to clear debris 

from roadways and facilitate plant access, if needed. The EPZ is largely agricultural, and there 

are multiple access roads to the plant. The six-lane I-95 and four-lane U.S. 40 bridges south of

the Peach Bottom site that cross the Susquehanna River are assumed to fail in the earthquake. 

There are equipment suppliers on both sides of the river; therefore, loss of these bridges would 

only have a limited affect on delivery of equipment. A heavy airlift capacity is also available in 

the region if needed to support prompt delivery of equipment. 

It is expected that primary equipment needs for accident mitigation would be electrical 

generators and additional pumps to deliver water to the facility. Response staff will have 

identified that these resources must be obtained from offsite. There are numerous fire stations in 

the region with water trucks and pumper trucks that would respond. This would be the first 

option for portable water and pumping resources if onsite systems are not functional. Local fire 

trucks could be onsite within minutes of a request, and larger regional fire station resources could 

be onsite within 2 to 3 hours based on the proximity of these resources to the plant and the time

of the request. The local fire departments are primarily volunteer rural fire departments that are 

familiar with operating in areas without public water supplies and are experienced with drafting 

water from ponds and rivers should this be necessary.

For additional or larger capacity resources, Philadelphia is approximately 70 miles east and 

Baltimore is about 50 miles south of the plant. A review of heavy equipment supply companies 

in these cities has found that electrical generators and pumping equipment are widely available. 

As equipment needs are identified, INPO, NEI and utility personnel would be coordinating 

acquisition and delivery of industry resources or commercially available resources, as needed. It 

is assumed that large portable generators and pumps, if needed, could be obtained and 

transported via truck to the site within 8 hours of the start of the accident.

If roadways were inaccessible, the Pennsylvania National Guard air wing in Fort Indiantown 

Gap, PA, is less than 100 miles from the site and has Chinook-47 helicopters rated to lift about 

26,000 pounds. Trailer mounted 600 kilowatt generators weigh about 22,000 pounds and may be 

the largest generators that can be airlifted to the site in a timely manner. These generators are 

large enough to support a variety of onsite power needs, including pumps. The Public 

Information Officer for the air wing confirmed that there are typically five to 25 operational 

helicopters available, and there are other air wings in the region. Given national response to a 

GE, it is assumed heavy lift helicopters would be made available within about 6 hours to support 

delivery of electrical generators, piping, heavy equipment, and other resources that may be 
6-43

needed. It would take additional time for the helicopters to pick up and deliver the equipment to 

the site.

6.6.2 Mitigation

The above evaluation explains why it is expected that the management of the emergency, 

technical expertise, and resources likely needed to mitigate the accident will be available. The 

next step is applying these resources to mitigate the accident. If other actions are not effective, 

the mitigative actions identified in the plant Severe Accident Management Plan (SAMP)

ultimately direct primary containment flooding to a level of 4 feet above the drywell floor. The 

EROs for the site and utility will be identifying methods to achieve the primary containment 

flooding, and the NRC would be independently reviewing the approaches being considered. At 

this stage of the accident, there are many parallel activities underway which focus on injecting 

water into the drywell. The current analysis assumes efforts to restore coolant injection directly 

to the drywell are not successful. Ad hoc means are considered for flooding the reactor building 

to mitigate the release to the environment. The flooding actions would provide some scrubbing 

of the source term and would reduce the release.

Several strategies are envisioned for flooding the reactor building. The mitigation strategy 

considered in this evaluation is one that floods the Reactor Building with water to cover the 

release pathway out of containment, thereby mitigating the release by scrubbing fission products 

that emerge through the failed drywell liner. If the opening in the liner is sufficiently large, water 

could also flow into the containment, covering and cooling core debris. This particular strategy 

would utilize portable pumps and fire hoses to transfer water from the cooling tower basin or 

other nearby sources to the basement of the reactor building. Direct access to the basement is not 

necessary. The compartments in the lower portion of the Reactor Building, such as the Torus 

room and ECCS pump rooms, are designed for internal flooding to an elevation of approximately 

111 feet, which is below the drywell floor elevation. Under normal operations, the Torus room 

doors are open. If the Torus room is flooded and doors are closed, water that rises above an 

elevation of 111 feet may flow into adjacent rooms through openings in the walls. This could 

result in the flooding of compartments that contain the major coolant pumps, such as Core Spray 

and might also result in leakage into adjacent buildings such as the Radwaste Building. Leakage 

to adjacent buildings would need to be controlled57. 

The approach to flooding the Reactor Building lower elevations described here includes directing 

fire hose streams through the open truck bay doors. This would result in the water entering the 

Torus room through the open grating at elevation 135-ft of the Reactor Building (see 

Figure 6-19). This approach avoids the need for Core Spray pumps and the need to enter the 

building to align the Core Spray system valves. The outside equipment hatch to the Torus room 

on the west side of the building could also be removed to provide an additional location for 

injection and monitoring of the water level. Calculations developed using building dimensions 

and depth of water show that approximately 2.5 million gallons would be necessary to fill the 

building to a level sufficiently deep to generate a static head for water to flow onto the drywell 

 

57 The mitigation scheme examined here assumes no leakage out of the reactor building into adjacent 

buildings. It is assumed that the initiating event does not cause subterranean structural damage to the 

building, nor would other leak paths out of the building preclude flooding up to grade level
6-44

floor via the opening in the drywell liner, if containment failure occurred as calculated in the 

STSBO or LTSBO scenarios evaluated in Section 5.

There are three potential water sources available. The primary source would be the onsite 

3.55 million gallon cooling tower basin, which is assumed to survive the earthquake. The 

structure is made of reinforced concrete and is designed to Seismic Class I requirements. Water 

would also be available from the Susquehanna River. Water tanker trucks supplied by local fire 

departments could also augment water supplies, but is not a significant source. If the river 

recedes due to downstream affects from the earthquake, water would need to be drafted from 

existing or man-made pools that may be away from the existing shoreline. This may require 

field fabrication of pontoons or temporary platforms to support suction piping across the now 

exposed river bottom. This could be accomplished with local equipment and available supplies. 

There are no physical impediments to prevent installation of these temporary fixtures; therefore, 

it is reasonable to assume this could be in place within a few hours of the decision to draft water 

from the river.

Portable pumps and fire trucks would be used to deliver water into the building. Piping and 

hoses may be directed through the open truck bay doors. If radiation dose is an issue, placement 

may require remote techniques, such as pushing pipe segments through the doorway from a 

distance, but it is reasonable to expect that mechanical expertise is onsite to accomplish these 

actions and that in-field designs are developed while plant staff are awaiting delivery of offsite 

resources. Materials are staged onsite to control radiological runoff from a 10CFR50.54(hh) 

event that could be used to promote water flowing into the lower elevations, rather than draining 

out the open doorway. A monitor nozzle could be used on one of the hose lines set on the fog 

spray mode of operation to provide some scrubbing of the release as it progressed from the Torus 

room up through the open equipment hatches in the vicinity of the truck bay. 
6-45

Figure 6-19 Reactor Building

The water would flood the Torus Room and adjacent rooms until it reaches a level above the 

hole in the reactor. At this point, the water may enter the reactor vessel and continue rising until 

it covers the molten core on the drywell floor, mitigating the accident. Alternately, there may be 

pressure from the core melt that could prevent water from entering the hole. In this case, the 

water would continue rising outside of the reactor vessel and the hole would become submerged. 

Any release would then be scrubbed through the water covering the hole. As indicated in Figure 

6-20, about 2,500,000 gallons would be necessary to fill the building sufficiently to cover the 

molten core in the Peach Bottom reactor building. With a pumping capacity of 6,000 gpm,

equivalent to four to six pumps, 2,500,000 gallons of water could be delivered in about 7 hours 

after pumping starts as shown in Figure 6-21. This would be sufficient to submerge the molten 

Section A-A

Truck bay

N

A

A

B

B

El. 91' 6"

El. 135'

El. 165'

El. 195'

El. 234'
6-46

core to a depth of about one meter and mitigate the release. If the onsite water tank is available, 

a greater flow rate may be achieved and the pumping time would be less. If water must be 

drafted from the river, six pumps/pumper trucks could be configured in a reasonable time. 

Figure 6-20 Volume Needed to Fill the Peach Bottom Reactor Building

Figure 6-21 Peach Bottom Pumping Capacity and Time
6-47

The above assessment is not definitive and describes a response that may occur should all onsite 

efforts to mitigate the accident fail. Some of the actions taken are ad hoc and the availability of 

equipment is likely but not certain. It should be recognized that the proceduralized mitigative 

actions can be successful in preventing core damage. In summary, the key activities and 

associated timing described above include:

• Fire trucks will be onsite within minutes with additional support available from nearby 

towns and cities;

• Large capacity pumps and electrical generators transported to the site within a few hours;

• Flooding the Reactor Building at 6,000 gpm takes approximately 6 hours from the time 

pumping begins.

The types of resources needed are multi-use and would begin being acquired or established early 

in the response to support restoration of cooling, depressurizing the RCS, injecting into the RCS, 

etc. Based on the above, flooding of the Reactor Building would need to begin about 40 hours 

after the start of the accident. It is reasonable to assume that within 40 hours response personnel 

would be prepared to make a decision to flood the Reactor Building. The approach described 

provides a supporting basis for truncating the release at 48 hours. 

6.7 Emergency Preparedness Summary and Conclusions

Advancements in consequence modeling provide an opportunity to include detailed emergency 

planning and response activities in the assessment of an accident. This includes the ability to 

model protective action decisions from OROs and the implementation of these decisions by 

individual population segments. To support the modeling effort, detailed demographic data was 

obtained from local and regional sources, and emergency response planning information was

obtained from offsite emergency response plans and through discussions with OROs. This 

information supported the evaluation of differences in the implementation of protective actions 

for the population segments. 

The modeling advancements are significant because they now allow detailed modeling of 

response activities, timing of decisions, and implementation of protective actions across different 

population segments. The Peach Bottom EALs were obtained from site procedures for each of 

the accident scenarios. This provides the best estimate for the timing of the declaration of an

SAE and a GE. The response of OROs with regard to sounding sirens, notifying the public, 

coordinating the evacuation efforts, and establishing return criteria, was obtained from 

discussions with the OROs and review of emergency plans. The response timing for the public 

was based on emergency plans, the site specific ETE, and research of public actions in response 

to large scale evacuations [22]. Previous consequence analyses, such as Sample Problem A [27], 

NUREG-1150 [9], and NUREG-6953 Volume 1 [23], typically used only one cohort to represent 

the EPZ public and one cohort for the non-evacuating public. In the NUREG-1150 analysis, the 

EPZ population was first sheltered and then evacuated as a single group at one speed. 

The SOARCA analyses included multiple cohorts and considers sheltering and speeds 

representative of each specific cohort:
6-48

• For the general public, shielding factors appropriate for the region were applied for 

normal activities, sheltering, and evacuation times. Evacuation speeds were developed 

from the Peach Bottom ETE study.

• Schools were modeled assuming they are notified directly in accordance with offsite 

emergency response plans. 

• Special facilities were modeled assuming they are notified early, but respond differently 

than schools because of the need to mobilize specialized transportation resources 

including wheelchair vans and ambulances. In actual practice, evacuation of these 

facilities would be on a facility-by-facility basis as resources become available. In 

SOARCA, these facilities were modeled as evacuating together after an extended 

sheltering period. The shielding protection values for these facilities are better than those 

for standard housing, and this was considered in the analysis.

• The evacuation tail includes those members of the public who take longer to evacuate 

and are the last to leave the area. Indoor shielding values were applied to this cohort. 

They were evacuated late in the emergency and move at faster speeds because of the 

lower volume of traffic on the roadways at the latter stages of the evacuation. The timing 

of the evacuation tail was developed from the ETE.

• A shadow evacuation, which occurs when people evacuate from areas that are not under 

an evacuation order, was represented in the area beyond the EPZ to account for additional 

vehicles on the roadway network. Including a shadow evacuation adds realism to the 

analysis and allows consideration of the impact of the shadow evacuation vehicles on the 

evacuation of the EPZ.

• A non-evacuating cohort representing 0.5 percent of the population was modeled in the 

analysis consistent with NUREG-1150. Normal activity shielding values were applied to 

this cohort.

• For the seismic analysis, it was assumed that a shadow evacuation of residents from 

within the EPZ occurs before the issuance of an evacuation order. This additional 

shadow evacuation cohort was included in the analysis.

Three sensitivity analyses were performed to understand the effects of expanding the limits of 

the evacuation and to understand any effects caused by a delay in implementation of protective 

actions. The sensitivity analyses were performed using the STSBO without RCIC blackstart

scenario because this has an earlier release than other scenarios. In the first two sensitivity 

analyses, the limits of the evacuation were extended to 16 miles and 20 miles. An ETE was 

developed using OREMS to establish evacuation response parameters for the consequence 

model. A third sensitivity analysis was conducted maintaining all of the base case parameters 

and increasing the delay of response actions by 30 minutes. The peer review committee had 

suggested the sensitivity of the response timing be evaluated. 
6-49

An analysis of the effects of a seismic event was also completed and showed that at this site there 

are relatively few roadway sections and bridges that fail in an earthquake. The roadway sections 

and bridges that might fail were dispersed enough that local traffic would be able to detour 

around these areas and exit the area without appreciable delay. The six-lane I-95 and four-lane 

U.S. 40 bridges south of the Peach Bottom site that cross the Susquehanna River are assumed to 

fail in the earthquake; however, this would only have a minor affect on the evacuation. Because 

the public is immediately aware of the earthquake, having experienced it, when notification to 

evacuate is received, they are prepared and respond promptly. As a result, timing of the public 

response occurs more quickly because the delay to evacuation is reduced while the travel time is 

about the same. 

The staff expects that plant actions would be successful in mitigating the important severe 

accident scenarios considered in this study. The truncation evaluation describes some of the 

types of resources, potential response times, and a conceptual approach that might be considered 

to mitigate an accident should onsite mitigation not be successful. 

The parameters developed for the base case and sensitivity analyses developed in this section 

provide input to the MACCS2 consequence model presented in Section 7.

7-1

7.0 OFF-SITE CONSEQUENCES

7.1 Introduction

The MACCS2 consequence model (Version 2.5.0.0) was used to calculate offsite doses and their 

effect on members of the public for the unmitigated accident scenarios. There was no need to 

compute consequences for the mitigated sequences because there were no source terms released 

to the environment in these scenarios. Updates to the current version of the MACC2 code used 

for SOARCA offsite consequence predictions are discussed NUREG-1935, Section 5. MACCS2 

was developed at Sandia National Laboratories for the NRC for use in PRAs for commercial 

nuclear reactors to simulate the impact of accidental atmospheric releases of radiological 

materials on humans and on the surrounding environment. The principal phenomena considered 

in MACCS2 are atmospheric transport using a straight-line Gaussian plume model of short-term 

and long-term dose accumulation through several pathways including cloudshine, groundshine, 

inhalation, deposition onto the skin, and food and water ingestion. The ingestion pathway was 

not treated in the analyses reported here because uncontaminated food and water supplies are 

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

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

• cloudshine during plume passage

• groundshine during the emergency and long-term phases from deposited aerosols

• inhalation during plume passage and following plume passage from resuspension of 

deposited aerosols. Resuspension is treated during both the emergency and long-term 

phases.

The SOARCA project made additional enhancements to MACCS2 [8]. In general, these 

enhancements reflect recommendations obtained during the SOARCA external review and also 

reflect needs identified by the broader consequence analysis community. The code 

enhancements done for SOARCA were primarily to improve fidelity and code performance and 

to enhance existing functionality. These enhancements are anticipated to have a significant 

effect on the fidelity of the analyses performed under the SOARCA project.

MACCS2 previously allowed up to three emergency-phase cohorts. Each emergency-phase 

cohort represents a fraction of the population who behave in a similar manner, although response 

times can be a function of radius. For example, a cohort might represent a fraction of the 

population who rapidly evacuate after officials instruct them to do so. To create a high-fidelity 

model for SOARCA, the number of emergency-phase cohorts was increased, as described in 

Section 6 of this report. This allowed significantly more variations in emergency response 

(e.g., variations in preparation time before evacuation to more accurately reflect the movement of 

the public during an emergency). In a similar way, modeling evacuation routes using the 

network-evacuation model added a greater degree of realism than in previous analyses that used 

the simpler, radial-evacuation model.
7-2

7.2 Peach Bottom Source Terms

Brief descriptions of the source terms for the Peach Bottom accident scenarios are provided in 

Table 7-1. For comparison, the largest source term (SST1) from the 1982 NUREG/CR-2239, 

“Technical Guidance for Siting Criteria Development” (referred to hereafter as “The 1982 Siting 

Study” or just “The Siting Study”) [28] is also shown. Of the Peach Bottom source terms shown 

in the table, the unmitigated STSBO without RCIC blackstart is the largest in terms of release 

fractions and the release begins at the earliest time; the unmitigated STSBO with RCIC 

blackstart and the unmitigated LTSBO are comparable in terms of release fractions; the release 

begins at the latest time for the LTSBO; the mitigated STSBO with RCIC blackstart is 

intermediate in terms of timing but slightly lower in terms of release fractions, with the exception 

of barium, which is slightly greater than the release fraction for the unmitigated LTSBO. 

The fission product inventory used in these analyses is presented in Appendix A. The inventory 

data were evaluated specifically for the SOARCA work and reflect realistic fuel cycle data from 

Peach Bottom. The inventory data are consistent with those used in MELCOR for decay heat 

values and for fission product masses.

In comparison, the SST1 source term is significantly larger in magnitude, especially for the 

cesium group, than any of the Peach Bottom source terms. Moreover, it begins only 1.5 hours 

after accident initiation. The current understanding of accident progression has led to a very 

different characterization of release signatures than was assumed for the 1982 Siting Study.

Table 7-1 Brief Source-Term Description for Unmitigated Peach Bottom Accident 

Scenarios and the SST1 Source Term from the 1982 Siting Study

Scenario CDF

(Events/yr)

Integral Release Fractions by Chemical Group

Atmospheric

Release 

Timing

Xe Cs Ba I Te Ru Mo Ce La Start

(hr)

End

(hr)

PB LTSBO 3x10-6 0.978 0.005 0.006 0.020 0.022 0.000 0.001 0.000 0.000 20.0 48.0

PB STSBO

w/ BS 3x10-7 0.979 0.004 0.007 0.013 0.015 0.000 0.001 0.000 0.000 16.9 48.0

PB STSBO

w/o BS 3x10-7 0.947 0.017 0.095 0.115 0.104 0.000 0.002 0.007 0.000 8.1 48.0

SST1 1x10-5 1.000 0.670 0.070 0.450 0.640 0.050 0.050 0.009 0.009 1.5 3.5

For comparison, a set of consequence analyses using the old SST1 source term is presented in 

this chapter. This allows a direct comparison, using the same modeling options and result 

metrics, of the SST1 source term and the current, best-estimate source terms. 

7.3 Consequence Analyses

The results of the consequence analyses are presented in terms of risk to the public for each of 

the three accident scenarios identified for Peach Bottom. Both conditional and absolute risks are 

tabulated. The conditional risks assume that the accident occurs and show the risks to 

individuals as a result of the accident. The absolute risks are the product of the CDF and the 

conditional risks. The absolute risks are the likelihood of receiving a fatal cancer or early fatality 

for an average individual living within a specified radius of the plant per year of plant operation.
7-3

The risk metrics are latent-cancer-fatality and early-fatality risks to residents in circular regions 

surrounding the plant. They are averaged over the entire residential population within the 

circular region. The risk values represent the predicted number of fatalities divided by the 

population for four choices of dose-truncation level. These risk metrics account for the 

distribution of the population within the circular region and for the interplay between the 

population distribution and the wind rose probabilities. 

Risk results are presented for three dose-response assumptions including: 1) linear no threshold 

(LNT); 2) US average natural background dose rate combined with average annual medical 

exposure as a dose truncation level (US BGR), which is 620 mrem/yr; and 3) a dose truncation 

level based on the Health Physics Society’s Position that there is a dose below which, due to 

uncertainties, a quantified risk should not be assigned (HPS), which is 5 rem/yr with a lifetime 

limit of 10 rem. A 10 mrem/yr dose truncation level was investigated, but it produced results 

that were just slightly lower than with the LNT assumption, and thus were not included in the 

final version of this document.

In addition to the base case mitigated and unmitigated accident scenarios, four additional 

sensitivity analyses are reported in this chapter. A sensitivity analysis for the unmitigated 

STSBO without RCIC blackstart scenario shows the influence of the size of the evacuation zone 

or a delay in evacuation on predicted risk. Another sensitivity analysis considers the effect of 

seismic activity on emergency response. This sensitivity is considered because the base case 

results account for the effect of the seismic event on the plant but do not account for its effects 

on evacuation. This sensitivity analysis takes the latter effects into account as well. A separate 

analysis of the SST1 source term [28] (summarized in Table 7-1) allows older source-term 

assumptions to be compared with the current state-of-the-art methods for source-term evaluation 

using otherwise equivalent assumptions and models. This analysis does not try to reproduce the 

1982 Siting Study results; it merely overlays the older source term onto what are otherwise 

SOARCA assumptions for dose-response modeling, emergency response, and other factors. The 

final sensitivity analysis attempts to replicate the 1982 Siting Study using the current version of 

MACCS2.

In this section, the risk tables represent rounded values obtained from the full data sets. The 

plots were developed from the full data sets and slight differences may be noticed due to this 

rounding. 

7.3.1 Unmitigated Long-Term Station Blackout Scenario

Table 7-2 displays the conditional, mean, latent-cancer-fatality (LCF) risks to residents within a 

set of concentric circular areas centered at the Peach Bottom site for the unmitigated LTSBO 

scenario. Three values of dose-truncation level are shown in the table: linear, no threshold 

(LNT); the average, annual, US-background radiation (including average medical radiation) of 

620 mrem/yr (US BGR); and a dose truncation level based on the Health Physics Society’s 

Position that there is a dose below which, due to uncertainties, a quantified risk should not be 

assigned (HPS), which is 5 rem/yr with a lifetime dose limit of 10 rem.
7-4

Dose-truncation based on the HPS Position is more complex than the US BGR truncation 

because it involves both annual and lifetime limits. According to the recommendation, annual 

doses below the 5-rem truncation level do not need to be counted toward health effects; however, 

if the lifetime dose exceeds 10 rem, all annual doses, no matter how small, count toward health 

effects. Because of the 10 rem lifetime limit, risks predicted with the criterion based on the HPS 

Position can sometimes exceed those using the background radiation level for dose truncation.

Table 7-2 Mean, Individual, LCF Risk per Event (Dimensionless) for Residents within 

the Specified Radii of the Peach Bottom Site for the Unmitigated LTSBO

Scenario, which has a Mean, Core Damage Frequency (CDF) of 3x10-6 pry

Radius of Circular Area 

(mi) LNT US BGR HPS

10 8.9x10-5 7.4x10-7 3.7x10-7

20 7.6x10-5 1.9x10-5 2.2x10-6

30 5.3x10-5 1.1x10-5 8.9x10-7

40 3.3x10-5 5.0x10-6 3.7x10-7

50 2.7x10-5 3.4x10-6 2.4x10-7

Table 7-3 is analogous to Table 7-2, but displays the absolute rather than the conditional risks. 

In the case of the Peach Bottom LTSBO, the mean CDF of 3x10-6 pry is used, a frequency that is 

based on the assumption that B.5.b mitigation does not succeed (see Section 3.1.4). The absolute

risk is the product of the conditional risk and this CDF.

Table 7-3 Mean, Individual, LCF Risk per Reactor-Year (1/yr) for Residents within the 

Specified Radii of the Peach Bottom Site for the Unmitigated LTSBO 

Scenario, which has a Mean, CDF of 3x10-6 pry

Radius of Circular Area

(mi) LNT US BGR HPS

10 2.7x10-10 2.2x10-12 1.1x10-12

20 2.3x10-10 5.8x10-11 6.5x10-12

30 1.6x10-10 3.3x10-11 2.7x10-12

40 1.0x10-10 1.5x10-11 1.1x10-12

50 8.0x10-11 1.0x10-11 7.1x10-13

The values in Table 7-3 are shown in Figure 7-1. The figure shows that for LNT, the risks are 

greatest for those closest to the plant and diminish monotonically as distance increases. On the 

other hand, for either of the dose-truncation levels, the risk reaches a maximum outside the 10

mile evacuation zone. The explanation for this counterintuitive trend is provided in the 

following discussion of the risks incurred during the emergency versus the long-term phases. The 

NRC Safety Goal for latent cancer fatality risk from nuclear power plant operation (i.e., 2x10-6 or 

two in one million) which is shown in Figure 7-1 and subsequent figures 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 lower than the NRC Safety Goal and millions of times lower 

than the general U.S. cancer fatality risk.
7-5

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

annual US cancer fatality risk from all causes are provided to give context that may help the 

reader to understand the contribution to cancer risks from these nuclear power plant accident 

scenarios. However, such comparisons have limitations for which the 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. In fact, any analytical technique, including PRAs, will have 

inherent limitations of scope and method. As a result, comparison of SOARCA’s scenariospecific 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.

Figure 7-2 shows the absolute LNT risks for the Peach Bottom unmitigated LTSBO for the 

emergency and long-term phases. The entire height of each column shows the combined (total) 

risk for the two phases. The emergency response is very effective within the evacuation zone 

(10 miles) during the early phase, so those risks are very small and entirely represent the 

0.5 percent of the population that are modeled as refusing to evacuate. The peak emergency

phase risk is at 20 miles, which is the first location in the plot outside of the evacuation zone.

Figure 7-1 Mean, individual, LCF risk per Reactor-Year (1/yr) from the Peach Bottom, 

unmitigated, LTSBO scenario for residents within a circular area of specified 

radius from the plant for three values of dose-truncation level

0.0E+0

5.0E-11

1.0E-10

1.5E-10

2.0E-10

2.5E-10

3.0E-10

10 20 30 40 50

Individual LCF Risk per Reactor Year 

(pry)

Radius of Circular Area (mi)

LNT

US BGR

HPS

Note: The NRC Safety Goal for LCF Risk is 2.0 x 10-6 pry
7-6

Figure 7-2 Mean, individual, LNT, LCF risk per Reactor-Year (1/yr) from the Peach 

Bottom, unmitigated, LTSBO scenario for residents within a circular area of 

specified radius from the plant for the emergency and long-term phases

The long-term phase risks dominate the total risks for the accident scenario when the LNT 

dose-response assumption is made. These long-term risks are controlled by the habitability 

(return) criterion, which is the dose rate at which residents are allowed to return to their homes 

following the emergency phase. For Peach Bottom, the habitability criterion is an annual dose 

rate of 500 mrem/yr. This dose rate is below the truncation levels based on US background 

(620 mrem/yr) and based on the HPS Position; therefore, most of the doses received during the 

long-term phase are below the dose-truncation limit and are not counted toward health effects 

when using these criteria. Thus, most of the risks associated with either of the truncation levels

are from doses received during the first year. 

To better understand this explanation, it is important to understand the differences between 

exposure periods, commitment periods, and the periods of time when doses are actually received.

For external dose pathways, the time over which doses are received is concurrent with the 

exposure period. External dose pathways include cloudshine and groundshine. 

The exposure period for internal pathways, inhalation and ingestion, is the period of time when 

the inhalation or ingestion occurs; however, doses continue to be received over a person’s entire 

lifetime following the exposure. A person’s lifetime is obviously variable, depending upon the 

age of the person at the time of exposure among other things. The period of time over which 

doses are received from an internal pathway is accounted for in the construction of dose 

conversion factors by integrating the doses over a finite period called a dose commitment period, 

which is usually taken to be 50 years when calculating internal-pathway dose conversion factors 

0.0E+0

5.0E-11

1.0E-10

1.5E-10

2.0E-10

2.5E-10

3.0E-10

10 20 30 40 50

Individual LNT LCF Risk per Reactor 

Year (pry)

Radius of Circular Area (mi)

Long-Term Phase

Emergency Phase

Note: The NRC Safety Goal for LCF Risk is 2.0 x 10-6 pry
7-7

for adults. The implicit assumption is that the average adult lives for an additional 50 years

following the exposure, which is most likely a conservative assumption. 

Since ingestion doses are taken to be negligible in SOARCA, inhalation is the only internal 

pathway that is treated. A significant portion of the exposures during the emergency phase are 

from inhalation. As explained above, these exposures are assumed to lead to doses over the 

commitment period, which is the next 50 years following the exposure. However, depending on 

the isotope inhaled, the doses received may diminish rapidly and become negligible for most of 

the dose commitment period.

Most of the exposures during the long-term phase are from groundshine; a small fraction is from 

inhalation of resuspended aerosols. Since groundshine is an external pathway, doses received 

are concurrent with the exposure. On the other hand, exposures from inhalation during each year 

of the long-term phase contribute to doses received over the subsequent 50-yr commitment 

period.

Doses received in the first year thus correspond to:

• all of the dose from external exposure during the emergency phase,

• most of the dose from internal exposure during the emergency phase,

• all of the dose from external exposure during the first year of the long-term phase, and

• most of the dose from internal exposure during the first year of the long-term phase.

Doses received in the second and subsequent years correspond to:

• a fraction of the dose from internal exposure during all previous years plus most of the 

dose from internal exposure during that year, and

• all of the dose from external exposure during that year. 

Following a single exposure, internal doses decrease more slowly from one year to the next when 

the isotopic half life is relatively long (i.e., on the order of a year or longer) and the solubility of 

the dominant chemical form of the isotope is low so that the removal rate from the human body 

is low (i.e., the biological half life is long). A good example is 90Sr, for which the second-year 

effective dose from inhalation is 60% of the first-year dose. The isotopic half life is 29 years, so 

most of the reduction from year one to year two results from the biological half life. The internal 

doses decrease more rapidly from one year to the next when either the isotopic half life is short 

or when the solubility of the dominant chemical form of the isotope is high so that the human 

body tends to excrete it rapidly. A good example of this is 131I, for which the second-year 

effective dose from inhalation is essentially zero. This isotope has a short isotopic half life (i.e., 

8 days) and a short biological half life because of its high solubility. For comparison, the 

second-year effective dose from inhalation for 137Cs is about 10% of the first-year dose, so it is 

intermediate between the previous examples.
7-8

Because the internal doses from inhalation diminish with time, most of the doses in the second 

and subsequent years are from the exposures during that year. But these doses are limited by the 

habitability criterion to be less than 500 mrem in any year. The 500 mrem limit is for all dose 

pathways, in this case groundshine and inhalation from resuspended aerosols. The inhalation 

dose used in this criterion is a committed dose (i.e., it accounts for doses received over the next 

50 years). Because the annual doses allowed by the habitability criterion are less than truncation 

levels based on US background and the HPS Position, nearly all of the risk is from doses 

received during the first year. These doses include most of emergency-phase doses and a 

fraction of the long-term phase doses. This explains why the risk profiles for these dosetruncation criteria in Figure 7-1 are similar to the emergency-phase profile in Figure 7-2. 

The prompt-fatality risks are zero for this accident scenario. This is because the release fractions 

(shown in Table 7-1) are too low to produce doses large enough to exceed the dose thresholds for 

early fatalities, even for the 0.5 percent of the population that are modeled as refusing to 

evacuate. The largest value of the mean, acute dose for the closest resident (i.e., 0.5 to 1.2 

kilometers from the plant) for this scenario is about 0.1 gray (Gy) to the red bone marrow, which 

is usually the most sensitive organ for prompt fatalities, but the minimum acute dose that can 

cause an early fatality is about 2.3 Gy to the red bone marrow. Clearly, the calculated doses are 

all well below this threshold. 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.

7.3.2 Short-Term Station Blackout with Reactor Core Isolation Cooling Blackstart

Table 7-4 displays the conditional, mean, latent-cancer-fatality risks to residents within a set of 

concentric circular areas centered at the Peach Bottom site for the STSBO scenario with 

successful RCIC blackstart. The RCIC blackstart delays the beginning of release and provides 

more time for evacuation before release than in the subsequent scenario, in which RCIC 

blackstart is not attempted or fails. 

Table 7-4 Mean, Individual, LCF Risk per Event (Dimensionless) for Residents within 

the Specified Radii of the Peach Bottom Site for the STSBO Scenario with 

RCIC blackstart, which Has a Mean CDF of 3x10-7 pry

Radius of Circular Area 

(mi) LNT US BGR HPS

10 7.1x10-5 6.5x10-7 3.0x10-7

20 6.5x10-5 1.6x10-5 1.2x10-6

30 4.6x10-5 9.2x10-6 4.4x10-7

40 2.9x10-5 4.4x10-6 1.8x10-7

50 2.4x10-5 3.0x10-6 1.1x10-7

Table 7-5 is analogous to Table 7-4 but shows absolute rather than conditional risks. These risks 

are shown graphically in Figure 7-3. In the case of the Peach Bottom STSBO with RCIC 

blackstart, the mean CDF of 3x10-7 pry is used, a frequency that is based on the assumption that 

B.5.b mitigation does not succeed (see Section 3.2.4).
7-9

Table 7-5 Mean, Individual, LCF Risk per Reactor-Year (1/yr) for Residents within the 

Specified Radii of the Peach Bottom Site for the Unmitigated STSBO 

Scenario with RCIC blackstart, which has a Mean CDF of 3x10-7 pry

Radius of Circular Area 

(mi) LNT US BGR HPS

10 2.1x10-11 2.0x10-13 9.0x10-14

20 2.0x10-11 4.8x10-12 3.6x10-13

30 1.4x10-11 2.8x10-12 1.3x10-13

40 8.8x10-12 1.3x10-12 5.3x10-14

50 7.1x10-12 8.9x10-13 3.4x10-14

Figure 7-3 Mean, individual, LCF risk per Reactor-Year (1/yr) from the Peach Bottom, 

STSBO scenario with RCIC blackstart for residents within a circular area of 

specified radius from the plant for three values of dose-truncation level

Figure 7-4 shows the individual, LNT, LCF risks for the Peach Bottom STSBO with RCIC 

blackstart for the emergency and long-term phases. The height of each column indicates the 

combined (total) risk for the two phases. The emergency response is very effective within the 

evacuation zone (10 miles) during the emergency phase, so those risks are very small and 

entirely represent the 0.5 percent of the population that are modeled as refusing to evacuate. The 

peak in the EARLY risk curve is at 20 miles, which is the first location in the plot outside of the 

evacuation zone.

0.0E+0

5.0E-12

1.0E-11

1.5E-11

2.0E-11

2.5E-11

10 20 30 40 50

Individual LCF Risk per Reactor Year 

(pry)

Radius of Circular Area (mi)

LNT

US BGR

HPS

Note: The NRC Safety Goal for LCF Risk is 2.0 x 10-6 pry
7-10

Figure 7-4 Mean, individual, LNT, LCF risk per Reactor-Year (1/yr) from the Peach 

Bottom, STSBO scenario with RCIC blackstart for residents within a 

specified radius from the plant for emergency and long-term phases

The trends for this accident scenario are very similar to those for the unmitigated LTSBO 

scenario. The long-term-phase risks for this scenario are greater than the emergency-phase risks, 

especially within the evacuation zone (10 miles), where the emergency-phase risks are very 

small. The long-term risks are controlled by the habitability or return criterion, which is an 

annual dose limit of 500 mrem.

Because the annual dose limit of the habitability criterion (500 mrem/yr) is lower than the dose 

truncation levels based on the US background (620 mrem/yr) and HPS Position, those two risk 

profiles (shown in Figure 7-3) are similar to the emergency-phase profile shown in Figure 7-4. 

In other words, the long-term doses are largely excluded by either of the dose-truncation criteria, 

so the health effects are dominated by doses received during the emergency phase. As a result, 

those risk profiles are similar to the emergency-phase profile in Figure 7-4.

The prompt-fatality risks are identically zero for this accident scenario. This is because the 

release fractions are too low to produce doses large enough to exceed the dose thresholds for 

early fatalities, even for the 0.5 percent of the population modeled as refusing to evacuate.

7.3.3 Unmitigated Short-Term Station Blackout without RCIC Blackstart

Table 7-6 displays the conditional, mean, latent-cancer-fatality risks to residents within a set of 

concentric circular areas centered at the Peach Bottom site for the unmitigated STSBO scenario 

without RCIC blackstart. The releases for this scenario are larger and earlier than those for 

either of the previous ones. 

0.0E+0

5.0E-12

1.0E-11

1.5E-11

2.0E-11

2.5E-11

10 20 30 40 50

Individual LN T LCF Risk per Reactor 

Year (pry)

Radius of Circular Area (mi)

Long-Term Phase

Emergency Phase

Note: The NRC Safety Goal for LCF Risk is 2.0 x 10-6 pry
7-11

Comparing Table 7-4 and Table 7-6 reveals that the risks are larger for the STSBO when RCIC 

blackstart does not succeed. Table 7-7 is analogous to Table 7-6 but shows absolute rather than 

conditional risks. In the case of the Peach Bottom STSBO without RCIC blackstart, the mean 

CDF of 3x10-7 pry is used, a frequency that is based on the assumption that B.5.b mitigation does 

not succeed (see Section 3.2.4).

Table 7-6 Mean, Individual, LCF Risk per Event (Dimensionless) for Residents within 

the Specified Radii of the Peach Bottom Site for the Unmitigated STSBO 

Scenario without RCIC blackstart, which Has a Mean CDF of 3x10-7 pry

Radius of Circular Area 

(mi) LNT US BGR HPS

10 2.1x10-4 1.2x10-5 1.3x10-5

20 5.7x10-4 3.7x10-4 3.7x10-4

30 3.9x10-4 2.4x10-4 2.2x10-4

40 2.4x10-4 1.3x10-4 1.1x10-4

50 1.9x10-4 9.7x10-5 7.3x10-5

Table 7-7 Mean, Individual, LCF Risk per Reactor-Year (1/yr) for Residents within the 

Specified Radii of the Peach Bottom Site for the Unmitigated STSBO 

Scenario without RCIC blackstart, which Has a Mean CDF of 3x10-7 pry

Radius of Circular Area 

(mi) LNT US BGR HPS

10 6.2x10-11 3.7x10-12 3.8x10-12

20 1.7x10-10 1.1x10-10 1.1x10-10

30 1.2x10-10 7.3x10-11 6.6x10-11

40 7.1x10-11 4.0x10-11 3.2x10-11

50 5.6x10-11 2.9x10-11 2.2x10-11

Table 7-7 is plotted in Figure 7-5. Due to rounding of the full data set, the plot does not align 

precisely with the table values. The plot shows that predicted risks reach a maximum beyond the 

EPZ (10 miles) for all choices of dose truncation level. Risks within the 10 mile evacuation zone 

are very small for either of the dose-truncation criteria because long-term annual doses are below 

these truncation levels.
7-12

Figure 7-5 Mean, individual, LCF risk per Reactor-Year (1/yr) from the Peach Bottom, 

unmitigated, STSBO scenario without RCIC blackstart for residents within a 

specified radius from the plant for three values of dose-truncation level

Figure 7-6 shows the LNT latent-cancer fatality risks for the Peach Bottom unmitigated STSBO 

scenario without RCIC blackstart for the emergency and long-term phases. The height of each of 

the columns shows the combined (total) risk for the two phases. The emergency response is very 

effective within the evacuation zone (10 miles) during the emergency phase, so those risks are 

very small and mostly represent the 0.5 percent of the population that are modeled as refusing to 

evacuate. The peak in the emergency-phase risk profile is at 20 miles, which is the first location 

in the plot outside of the evacuation zone. 

0.0E+0

2.0E-11

4.0E-11

6.0E-11

8.0E-11

1.0E-10

1.2E-10

1.4E-10

1.6E-10

1.8E-10

Individual LCF Risk per Reactor Year 

10 20 30 40 50

(pry)

Radius of Circular Area (mi)

LNT

US BGR

HPS

Note: The NRC Safety Goal for LCF Risk is 2.0 x 10-6 pry
7-13

Figure 7-6 Mean, individual, LNT, LCF risk per Reactor-Year (1/yr) from the Peach 

Bottom, unmitigated, STSBO scenario without RCIC blackstart for residents 

within a specified radius for the emergency and long-term phases

The long-term phase risks for this scenario are significantly smaller than the emergency-phase 

risks except within the evacuation zone (10 miles), where emergency-phase risks are very small. 

The long-term risks are controlled by the habitability or return criterion, which is an annual dose 

limit of 500 mrem. Because the overall risks are dominated by the emergency-phase risks, the 

risk profiles for all truncation levels have peaks at 20 miles, as shown in Figure 7-5.

Because the annual dose limit for the habitability criterion is lower than either of the dose 

truncation levels, those two risk profiles (shown in Figure 7-5) are mainly influenced by risks 

during the emergency phase, as shown in Figure 7-6 and explained in more detail in Section 

7.3.1. 

The contribution of the emergency phase to the overall risk is much greater for the Peach Bottom 

unmitigated STSBO without RCIC blackstart than for other Peach Bottom scenarios discussed 

above and for all of the Surry Power Station scenarios presented in Section 7 of the companion 

Surry report. The uniqueness of the unmitigated STSBO without RCIC blackstart scenario 

appears to be related to the relatively large releases for the Ba, I, Te, and Ce classes compared 

with the Cs class for this scenario (see Table 7-1). While the Cs class release fraction is about a 

factor of 3 greater than the other Peach Bottom scenarios, the Ba class release fraction is more 

than an order-of-magnitude greater. The cesium group, especially 137Cs, tends to dominate the 

long-term doses following an accident. Most of the other isotopes, e.g., the iodine, tellurium, and 

barium isotopes, tend to contribute more to short-term doses.

0.0E+0

2.0E-11

4.0E-11

6.0E-11

8.0E-11

1.0E-10

1.2E-10

1.4E-10

1.6E-10

1.8E-10

10 20 30 40 50

Individual LNT LCF Risk per Reactor 

Year (pry)

Radius of Circular Area (mi)

Long-Term Phase

Emergency Phase

Note: The NRC Safety Goal for LCF Risk is 2.0 x 10-6 pry
7-14

The prompt-fatality risks are zero for this accident scenario. This is because the release fractions 

are too low to produce doses large enough to exceed the dose thresholds for early fatalities, even 

for the 0.5 percent of the population modeled as refusing to evacuate.

7.3.3.1 Sensitivity Analyses on the Size of the Evacuation Zone and the Evacuation Start 

Time

The base case analysis included evacuation of the 10-mile EPZ and a shadow evacuation cohort 

between 10 and 20 miles. For the unmitigated STSBO scenario without RCIC blackstart, three 

additional calculations were performed to assess variations in the protective actions.

Sensitivity 1 - Evacuation of a 16-mile Circular Area

In this calculation, the evacuation zone is expanded to 16 miles. Shadow evacuation occurs from 

within the 16 to 20 mile area.

 

Sensitivity 2 - Evacuation of a 20-mile Circular Area 

In this calculation, the evacuation zone is expanded to 20 miles. No shadow evacuation beyond 

the evacuation zone is considered. 

Sensitivity 3 – Delayed Evacuation of a 10-mile Circular Area

This calculation is identical to the base case described above except that implementation of 

protective action is delayed by 30 minutes.

The results of all three sensitivity analyses are compared with the base case in Table 7-8. The 

results for the case with delayed evacuation are identical to those for the base case; the other two 

sensitivities are slightly different than the base case, especially within 10 and 20-mile radii.

These results are also shown in Figure 7-7. Since the delayed evacuation case is identical to the 

base case, it is omitted from the figure.

Table 7-8 Effect of Size of Evacuation Zone on Mean, Individual, LNT, Latent Cancer

Fatality Risk per Reactor-Year (1/yr) for Residents within the Specified 

Radii of the Site for the Unmitigated STSBO Scenario w/o RCIC blackstart

Radius of

Circular Area 

(mi)

Base Case

10-mile 

Evacuation

Sensitivity 1

16-mile

Evacuation

Sensitivity 2

20-mile 

Evacuation

Sensitivity 3

10-mile Delayed

Protective Action

10 6.2x10-11 1.8x10-10 1.7x10-10 6.3x10-11

20 1.7x10-10 1.3x10-10 7.2x10-11 1.7x10-10

30 1.2x10-10 1.0x10-10 8.4x10-11 1.2x10-10

40 7.1x10-11 6.6x10-11 5.7x10-11 7.2x10-11

50 5.6x10-11 5.1x10-11 4.8x10-11 5.7x10-11
7-15

Figure 7-7 Mean, individual, LNT, LCF risk per Reactor-Year (1/yr) from the Peach 

Bottom, unmitigated, STSBO scenario without RCIC blackstart for residents 

within a circular area of specified radius from the plant showing the effect of 

the size of the evacuation zone

Although expanding the size of the evacuation zone decreases the latent cancer fatality risk 

beyond the 10 mile radius for the unmitigated STSBO without RCIC blackstart, the risk within 

10 miles increases with this change. This is because evacuating a larger area increases the time 

to evacuate the 10-mile region due to increased traffic congestion. For circular areas with greater 

than a 20-mile radius, the risk reduction associated with increasing the size of the evacuation 

zone is slight. Prompt-fatality risk remains zero for all three of these sensitivity cases.

7.3.4 Evaluation of the Effect of Seismic Activity on Emergency Response

Earlier sections of this chapter provide offsite health consequence estimates for unmitigated 

sensitivity cases that reflect the effects of the seismic event on emergency response for 

mitigation of the accident. However, these earlier sections do not reflect the effects of the 

seismic event on public evacuation. This section provides consequence estimates that also 

include the effects of the seismic event on public evacuation. These consequence estimates were 

developed for the STSBO without RCIC blackstart. Although this has a lower frequency and 

lower absolute risk than the LTSBO scenario, this scenario was chosen because it was believed 

to be the most likely to show an increase in risk. Seismic effects on emergency response (ER) are 

site-specific but, as the results in Table 7-9 demonstrate, they have no substantial effect on health 

consequences at Peach Bottom. Although sirens fail, alternative notification is adequate and a 

larger shadow evacuation is expected. Although bridges fail, they are not significant for 

evacuation; an adequate road network remains, and evacuation speeds are unchanged. In 

addition, accident progression timing predicted by realistic analysis is relatively slow so that 

there is some margin for emergency preparedness activation and execution. 

0.0E+0

2.0E-11

4.0E-11

6.0E-11

8.0E-11

1.0E-10

1.2E-10

1.4E-10

1.6E-10

1.8E-10

2.0E-10

10 20 30 40 50

Individual LNT LCF Risk per Reactor 

Year (pry) 

Radius of Circular Area (mi)

Base Case

10-Mile Evacuation

Sensitivity #1

16-Mile Evacuation

Sensitivity #2

20-Mile Evacuation

Note: The NRC Safety Goal for LCF Risk is 2.0 x 10-6 pry
7-16

The risk of a prompt fatality is unaffected in this sensitivity analysis and remains zero.

Table 7-9 Mean, Individual, LNT, LCF Risk per Event (Dimensionless) for Residents 

within the Specified Radii of the Peach Bottom Site for the Unmitigated, 

STSBO Scenario without RCIC blackstart Comparing the Unmodified 

Emergency Response and Emergency Response Adjusted for the Effect of 

Seismic Activity on Evacuation Routes and Human Response

Radius of Circular Area 

(mi)

Unmodified 

ER

ER Adjusted for

Seismic Effects

10 2.1x10-4 2.1x10-4

20 5.7x10-4 5.1x10-4

30 3.9x10-4 3.7x10-4

40 2.4x10-4 2.3x10-4

50 1.9x10-4 1.8x10-4

7.3.5 Evaluation of SST1 Source Term

An objective of the SOARCA project is to update and compare the quantification of 

consequences reported in earlier studies, such as the 1982 Siting Study. One of the dramatic 

differences between the SOARCA study and the 1982 Siting Study is the character of the 

radiological releases in terms of magnitude and timing. Because of these dramatic differences, it 

is useful to characterize and compare the risk to the public that derives from these releases.

The project evaluated the updated effects of the SST1 source term previously used in the 

1982 Siting Study, and then compared these effects to the most severe accidents in SOARCA. 

Although previous studies believed that the SST1 source term was important, we now understand 

the accident progressions leading to such large early releases are either much more unlikely than 

previously thought or physically impossible. 

The approach used in this section is to substitute the SST1 source term in place of the SOARCA 

source term into the MACCS2 input files for the unmitigated STSBO without RCIC blackstart

scenario. This comparison does not attempt to replicate NUREG/CR-2239; nor is it a 

comparison to its results. Such a comparison is provided in the next section, 7.3.6. Because it is 

true that many of the parameters and models used at the time of NUREG/CR-2239 differed from 

those being used in the current SOARCA study, this sensitivity is intended to focus solely on the 

influence of the source term on predicted consequences. 

The MACCS2 input files chosen were the ones developed for the unmitigated LTSBO and 

STSBO without RCIC blackstart scenarios. These sensitivity analyses show the impact of the 

improvements made in the source term methods and practices on the consequence results.

The characteristics of the SST1 source term as previously described in the 1982 Siting Study 

report are as follows:

• severe core damage

• essentially involves loss of all installed safety features

• severe direct breach of containment
7-17

An exact scenario and containment failure mechanism (e.g., hydrogen detonation, direct 

containment heating, or alpha-mode failure) are not specified in the report. 

Notification time (i.e., sounding a siren to notify the public that a GE has been declared) for the 

Peach Bottom unmitigated STSBO without RCIC blackstart occurs at 1.0 hour, as shown in 

Figure 6-6. Evacuation of the general public begins 1 hour later, or 2 hours after accident 

initiation, which is the same time as the first cohort was assumed to evacuate in the 1982 Siting 

Study. The beginning of release for the SST1 source term (see Table 7-1) occurs 1.5 hours after 

accident initiation, which is 30 min. before evacuation of the general public. 

While the 1982 Siting Study treated emergency response very simplistically, a major emphasis of 

the SOARCA project is to treat all aspects of the consequence analysis as realistically as 

possible. No attempt was made in this sensitivity analysis to reproduce the treatment of 

emergency response used in the 1982 Siting Study.

Table 7-10 shows the conditional latent-cancer-fatality risks for a release corresponding to the 

SST1 source term occurring at Peach Bottom based on the unmitigated STSBO scenario without 

RCIC blackstart. This comparison shows that dose truncation level has a very minor influence 

on predicted risk. That is because the doses with the SST1 source term are large enough to 

exceed the dose truncation levels for most of the affected population.

Table 7-10 Mean, Individual, LCF Risk per Event (Dimensionless) for Residents within 

the Specified Radii of the Peach Bottom Site for the SST1 Source Term

previously used in the 1982 Siting Study. All Parameters Other than for 

Source Term Are Taken from the Unmitigated STSBO Scenario without 

RCIC blackstart

Radius of Circular Area 

(mi) LNT US BGR HPS

10 3.3x10-3 3.2x10-3 3.1x10-3

20 1.8x10-3 1.6x10-3 1.5x10-3

30 1.0x10-3 9.0x10-4 8.2x10-4

40 6.1x10-4 4.9x10-4 4.2x10-4

50 4.6x10-4 3.5x10-4 3.0x10-4

Table 7-10 compares the LNT risks using the SST1 source term with those for the largest source 

term calculated for Peach Bottom in this study, the unmitigated STSBO. The LNT risk within 

10 miles for the SST1 source term using the STSBO ER timing is about a factor of 15 higher 

than the risk for the unmitigated STSBO. The 10-mile risk using a US BGR dose-truncation 

criterion is a factor of 250 higher when comparing the SST1 result using the STSBO ER timing 

with the unmitigated STSBO. At larger distances, the risks are less disparate. For example, the 

ratio is about a factor of 2.5 for a 50-mile area when comparing the LNT risks for the SST1 

source term using the STSBO ER parameters with those for the unmitigated STSBO.
7-18

Table 7-11 Mean, Individual, LNT, LCF Risk per Event (dimensionless) for Residents 

within the Specified Radii of the Peach Bottom Site for the SST1 Source 

Term previously used in the 1982 Siting Study Using Emergency Response 

Parameters from the STSBO Scenario. The Final Column of the Table Shows 

the SOARCA Results for the Unmitigated STSBO without RCIC blackstart

Radius of Circular Area 

(mi)

SST1 Using

STSBO w/o RCIC 

Blackstart ER

Unmitigated 

STSBO w/o RCIC 

Blackstart

10 3.3E-03 2.1E-04

20 1.8E-03 5.7E-04

30 1.0E-03 3.9E-04

40 6.1E-04 2.4E-04

50 4.6E-04 1.9E-04

The maximum risk is within 10 miles for the SST1 source term, which is largely because the 

release is very early and emergency response is not rapid enough to prevent exposures within the 

EPZ during the emergency phase. 

A notable feature of the risks presented in Table 7-10 is that the choice of dose truncation 

criterion has a minor influence on risk. This is very different than the SOARCA accident 

scenarios discussed in preceding sections. Figure 7-8 provides some insights into this behavior. 

For the SST1 source term, nearly all of the risk, especially at short distances from the plant, is 

from exposures that occur during the emergency phase. Because a significant fraction of these 

doses are received over a short period of time, and the doses are large due to the large source 

term, the range of dose truncation values considered in this study have little influence on 

predicted risks. Again, this is a very different trend than is observed for the current, state-of-theart source terms.
7-19

Figure 7-8 Mean, individual, LNT, LCF Risk per Reactor-Year (1/yr) from the SST1 

source term for residents within a circular area of specified radius from the 

Peach Bottom plant using the SOARCA STSBO ER timing and showing the 

risks from the emergency and long-term phases

Table 7-12 shows the individual early fatality risk for several circular areas of specified radii 

centered at the plant for the two SST1 cases. The individual early fatality risk is approximately 

1x10-2 near the plant. Unlike the source terms presented in the preceding sections, the predicted 

prompt-fatality risks are greater than zero. The SST1 release fractions are more than large and 

early enough to induce prompt fatalities for members of the public who live close to the plant. 

Early fatality risks for the unmitigated SOARCA scenarios are zero, as discussed earlier. 

The NRC quantitative health object (QHO) for prompt fatalities (5x10-7 pry) is generally 

interpreted as the absolute risk within 1 mile of the exclusion area boundary (EAB). For Peach 

Bottom, the EAB is 0.5 mile from the reactor building from which release occurs, so the outer 

boundary of this 1-mile zone is at 1.5 miles. The closest MACCS2 grid boundary to 1.5 miles 

used in this set of calculations is at 1.3 miles. Evaluating the risk within 1.3 miles should 

reasonably approximate the risk within 1 mile of the EAB. The core damage frequency stated 

for the SST1 source term in the 1982 Siting Study [28] is 10-5 pry, so the absolute risk of a 

prompt fatality for this source term is approximately 1.4x10-7 pry using the LTSBO ER model 

and 9.8x10-8 pry using the STSBO ER model. Even for this very large source term, the prompt 

fatality risk is below the QHO value of 5x10-7 pry.
7-20

Table 7-12 Mean, Individual, Prompt-Fatality Risk per Event (Dimensionless) for 

Residents within the Specified Radii of the Peach Bottom Site for the SST1 

Source Term previously used in the 1982 Siting Study Using Emergency 

Response Parameters from the SOARCA STSBO without RCIC blackstart 

Scenario

Radius of Circular Area 

(mi)

Probability of a Prompt 

Fatality Using LTSBO ER 

(dimensionless)

Probability of a Prompt 

Fatality Using STSBO ER 

(dimensionless)

1.0 1.4x10-2 9.9x10-3

1.3 1.4x10-2 9.8x10-3

2.0 7.1x10-3 4.2x10-3

2.5 4.9x10-3 2.8x10-3

3.0 3.3x10-3 1.8x10-3

3.5 1.8x10-3 1.0x10-3

5.0 5.0x10-4 2.8x10-4

7.0 2.2x10-4 1.2x10-4

10.0 9.7x10-5 5.4x10-5

7.3.6 Comparison with the 1982 Siting Study

This subsection discusses the results of a comparison of scenario-specific risk between SOARCA 

and the 1982 Siting Study analysis for the SST1 source term. Since the 1982 Siting Study does 

not provide latent cancer results at distances that are meaningful and comparable to those 

provided in the SOARCA study or to the NRC safety goal, an effort was made to reproduce the 

Sandia Siting Study results for Peach Bottom using the SST1 source term in order to produce 

results that are directly comparable to the SOARCA results. An exact reproduction of those 

results was not feasible because the CRAC2 code is no longer available and some of the models 

and modeling choices used in the 1982 Siting Study cannot be readily reconstructed. The current 

successor to the CRAC2 code, MACCS2, shares a number of models with its ancestor, but other 

models have been improved and therefore produce different results. However, those model 

parameters that were known or presumed to have been used in the 1982 Siting Study were 

chosen in an effort to reproduce the results of that study. The results presented in this report

were all computed with MACCS2 version 2.5.

The motivation for this calculation is to establish a point of comparison between the 1982 Siting 

Study and SOARCA. This subsection seeks to compare the results of the 1982 Siting Study and 

SOARCA by attempting to reproduce all aspects of the 1982 Siting Study, as best as can be 

done. Key aspects of the modeling are discussed in the following subsection, Section 7.3.6.1. 

Table 7-13 compares the release fractions from the Peach Bottom unmitigated STSBO scenario

without RCIC blackstart and the SST1 source term. The unmitigated STSBO without RCIC 

blackstart scenario was chosen for this comparison because it is the largest of the source terms 

for Peach Bottom that were evaluated as part of the SOARCA investigation. Its frequency is 

only 3x10-7

/yr compared with the frequency assigned to the SST1 source term of 10-5/yr, about a 

factor of 30 lower. 
7-21

Table 7-13 Total Release Fractions by Chemical Group Comparison between the SST1 

and the SOARCA Unmitigated STSBO without RCIC blackstart Peach 

Bottom Scenarios

Xe Cs Ba I Te Ru Mo Ce La

SOARCA 0.9470 0.017 0.095 0.115 0.104 0.000 0.002 0.007 0.000

SST1 1.000 0.670 0.070 0.450 0.640 0.050 0.050 0.009 0.009

7.3.6.1 Comparison of Modeling Choices

Table 7-14 compares key modeling choices and parameters used in the 1982 Siting Study with 

those used in SOARCA for Peach Bottom. This table reflects our understanding of the 

differences in key modeling aspects between these two studies. Some of the modeling choices 

listed in the table could be established with a reasonable degree of certainty from the 1982 Siting 

Study documentation; others represent best judgments as to how consequence analyses were 

performed at the time of the 1982 Siting Study. Generally, those judgments were based on 

NUREG-1150 or WASH-1400 modeling practices. Best judgments or approximations are 

denoted with an asterisk in the table. Each of the modeling choices shown in the table are 

discussed below.

Weather Sampling: The exact strategy that was used in the 1982 Siting Study is unknown. The 

Siting Study does show a binned representation of each of the weather files used in the study, so 

it is highly likely that weather binning was used. Also, the exact weather data that were used in 

the study are unknown. The 1982 Siting Study used meteorological data files from 29 regional 

weather stations to represent the weather at the 91 sites considered in the study. The 

consequence analysis for Peach Bottom used one of these files, but the documentation does not 

specify which one. Weather sampling used in this reconstruction uses the current Peach Bottom 

weather file and the NUREG-1150 choices for weather bin structure and samples per bin.

Habitability Criterion: The habitability criterion used in the Siting Study was 25 rem over an 

exposure period of 30 years. This criterion leads to higher long-term doses than the one used in 

SOARCA for Peach Bottom, which is 500 mrem over 1 year.

Emergency Response: Emergency response was treated simplistically and conservatively in the 

1982 Siting Study; the SOARCA treatment of emergency response is more detailed and realistic.

For example, 30% of the population began to evacuate by 2 hours after accident initiation in the 

1982 Siting Study; whereas, almost 93% of the population have begun to evacuate by 2 hours in 

the STSBO scenario. Also, SOARCA uses the more realistic network evacuation model to 

represent traffic on designated emergency routes. This model was not developed until after the

Siting Study. The evacuation speed, 10 MPH, used in the Siting Study, however, is faster than 

the travel speed used in the SOARCA representation of Peach Bottom once evacuation begins.

KI Ingestion: KI was not distributed at the time of the 1982 Siting Study. Because it was not 

distributed, no model for the effect of KI ingesting had been developed. Distribution of KI is 

relatively common now and is realistically accounted for in the SOARCA study.
7-22

Number of Sectors: The only option available at the time of the 1982 Siting Study was to model 

wind directions using 16 compass sectors. That capability has been extended, and SOARCA 

takes advantage of the full 64-sector capability in the current version of MACCS2.

Table 7-14 Comparison of Modeling Choices and Parameters Used to Reconstruct 

1982 Siting Study Results with the Peach Bottom Unmitigated STSBO

without RCIC blackstart from SOARCA

Modeling Choice or 

Parameter Siting Study SOARCA

Weather Sampling 142 Trials* 984 Trials

Habitability Criterion 25 rem in 30 yr 0.5 rem in 1 yr

Emergency Response 3 Cohorts 6 Cohorts

30% Evacuate at 2 hr 37.2% Evacuate at 1.0 hr

40% Evacuate at 4 hr 55.5% Evacuate at 2.0 hr

30% Evacuate at 6 hr 6.8% Evacuate at 5.25 hr

0.5% Do Not Evacuate

KI Ingestion No One Takes KI 50% Take KI with 70% Efficacy

Number of Sectors 16 64

Fission Product Inventory Low Burnup Mid-Cycle High Burnup

Deposition Velocity 1 cm/s 0.05 to 1.7 cm/s

Mixing Height Annual Ave. Day & Night Seasonal Ave.

Risk Factors for Cancers BEIR III* BEIR V

Population Basis Year 1970 2005

Groundshine Weathering WASH-1400* MACCS2

Relocation Criteria

Normal 25 rem / 24 hr* 1 rem / 24 hr

Hot Spot 50 rem / 12 hr* 5 rem / 12 hr

Plume Meander Model MACCS2* None

Dose Conversion Factors ICRP-26, -30* FGR-13

Food Ingestion Model COMIDA2* None

Fission Product Inventory: Burnups at the time of the 1982 Siting Study were much lower than 

today. The Siting Study report provides the fission product inventory used in that study. The 

inventory used for the SOARCA evaluation of Peach Bottom was based on current fuel cycle 

practices at Peach Bottom and assumes that the accident occurs mid-cycle. The values are laid 

out in Appendix A. For comparison, the inventory of 137Cs, the most important isotope for longterm doses, using in the 1982 Siting Study was about 65% of that used for Peach Bottom in the 

SOARCA study.

Deposition Velocity: Dry deposition of aerosol particles is represented through a set of aerosol 

size bins. Each size bin represents a range of aerosol sizes, usually characterized by a mass 

median diameter. Each aerosol bin is assigned a dry deposition velocity. The set of dry 

deposition velocities are used by MACCS2, along with airborne aerosol concentrations that are 

calculated using the Gaussian plume approximation, to determine the ground concentrations.
7-23

Common practice from the time of the 1982 Siting Study through NUREG-1150 was to treat a 

single aerosol bin using a representative deposition velocity of 1 cm/s. This single-bin practice is 

still common today. The practice, used in SOARCA is to use all of the aerosol data from 

MELCOR. These data are for 10 aerosol bins, each representing a range of aerosol sizes. The 

representative deposition velocities for the 10 bins range from 0.05 cm/s for the smaller particles 

to 1.7 cm/s for the larger ones. The dominant or average deposition velocity in SOARCA is 

about 0.3 cm/s, a factor of 3 lower than the single value used in the 1982 Siting Study.

Mixing Height: The 1982 Siting Study report shows mean annual daytime mixing heights for 

each representative weather station. Apparently, a single mixing height was used to represent the 

entire year. In SOARCA, seasonal average daytime and nighttime mixing heights are used.

Risk Factors for Cancer: Cancer risk factors used in the 1982 Siting Study are presumed to have 

come from the BEIR III report, which would have been the latest available at the time. Cancer 

risk factors in the SOARCA study are based on BEIR V [37]. BEIR V was chosen rather than 

BEIR VII because the treatment of tissues is consistent with the FGR-13 dose conversion factors. 

The BEIR V risk factors are about a factor of 2.7 higher than those from BEIR III.

Population Basis Year: To simplify recreation of the Siting Study results for Peach Bottom, the 

NUREG-1150 site file, which is for 1980, was used. Data provided in the Siting Study report 

give population densities at low resolution and would have been difficult to convert into a site 

file. This NUREG-1150 site file is based on the year 1980 rather than basis year for the Siting 

Study, which is believed to be 1970. However, individual risks only depend on the relative 

locations of the population, not on the total population. From that standpoint, using the 1980 

population data to reconstruct the Siting Study should have a minor effect on the comparison

presented below provided that the locations of the population centers did not change much.

Groundshine Weathering: The Siting Study report did not document the parameters used in the 

groundshine weathering model. It was judged that the model might have been the same as the 

one used in WASH-1400, which predated the Siting Study. The SOARCA model for 

groundshine weathering is the same as the one used in NUREG-1150. The specific model used 

turns out to play a small role for a large, early release like the SST1 source term because most of 

the doses are during the emergency phase. Weathering occurs during the long-term phase.

Relocation Criteria: The values used for normal and hot-spot relocation were not described in 

the Siting Study report, so the values were assumed to be the same as those used in 

NUREG-1150. The SOACA dose values to trigger relocation were much smaller, but the 

relocation times were the same.

Plume Meander Model: The plume meander model used in the Siting Study was assumed to be 

the same as the one used in NUREG-1150. Plume meander was not treated in SOARCA. We

considered using the NRC Reg. Guide 1.145 plume meander model, but it would have had a 

minimal impact on the predicted doses for even the closest residents to the Peach Bottom site

because this model only affects the plume dimensions at relatively short distances.
7-24

Dose Conversion Factors: The original version of MACCS2 was distributed with a set of dose 

conversion factors (DCFs) using tissue weighting factors from ICRP-26 and organ-specific 

DCFs from ICRP-30. These publications predated the 1982 Siting Study, so it is reasonable to 

expect that they were also used in the Siting Study. These DCFs were used in the reconstruction 

of the Siting Study SST1 results.

Food Ingestion Model: No details of the ingestion pathway are provided in the Siting Study

report, but it does mention that ingestion of contaminated food and milk were treated. The food 

ingestion model that would have been used certainly predates the implementation of the 

COMIDA2 food model, which first became available in MACCS2. Since the food model used 

in the Siting Study would be difficult or impossible to reconstruct, the COMIDA2 model was 

used as a stand in. For comparison, the food pathway was not treated in the SOARCA analyses.

Making all of the changes listed above plus replacing the Peach Bottom unmitigated STSBO 

source term with the SST1 source term resulted in a best-effort attempt to reproduce the 

1982 Siting Study results. However, this effort over-predicted the Siting Study latent cancer 

results using the SST1 source term for Peach Bottom by about a factor of 2 at long distance

(e.g., 500 miles). Thus, there are other changes in the models and parameter choices that were 

not captured in the attempt to reproduce this result. Nonetheless, even with this imprecision in 

recreating the 1982 Siting Study and a residual factor-of-2 bias in the results, this 

characterization of the Siting Study at shorter distances that can be compared directly with the 

SOARCA results provides a useful comparison. 

7.3.6.2 Comparison of Results

Table 7-15 compares the 1982 Siting Study conditional probabilities of an excess, individual 

latent cancer fatality using the SST1 source term with those for the unmitigated STSBO scenario

without RCIC blackstart evaluated in SOARCA. The comparison shows that the ratio of 

conditional probabilities within 10 miles of the plant is about 36. Accounting for a potential 

factor-of-2 bias, the ratio is about 20 within a 10-mile radius. Therefore, at the distance 

associated with the NRC Safety Goal for latent cancers, the risk predicted for SOARCA is 

substantially smaller than that predicted in the 1982 Siting Study. This ratio diminishes with 

increasing radius, becoming about a factor of 2 within a 50-mile radius and beyond. Again, 

accounting for a potential bias, the ratio may be more like a factor of unity. Therefore overall, 

when all exposed individuals are considered (i.e., regardless of their proximity to the site), little

difference in conditional latent cancer fatality risk is expected between the Peach Bottom 

STSTBO from the current work and the SST1 source term from the 1982 Siting Study. 

The decrease in the ratio from 20 to 1 occurs because the plume is expected to carry radioactive 

material farther from the site, relocation of the population beyond the 10-mile EPZ limits 

exposures during the emergency phase, and the habitability criterion limits exposures during the 

long-term phase. However, implementing the habitability requirement requires significantly 

greater decontamination and condemnation of land in the case of the 1982 Siting Study than for 

SOARCA.

The 1982 Siting Study was a consequence analysis and purposefully did not report absolute risk, 

because as the study states, “Probability times consequence is not an adequate representation of 
7-25

risk,” possibly because, “very large variations (factors of 10 to 100) in the accident probabilities 

associated with a specific design.” However, if a comparison was made using the suggested 

“representative probabilities” at the time, the factors would be much larger since the frequency 

of the Unmitigated STSBO without RCIC blackstart is about a factor of 30 lower than the 

frequency estimated for the SST1 source term. The ratios on the basis of risk (1/reactor year) are 

therefore about 600 for residents living within 10 miles of the plant and about 30 for residents 

living within 50 miles of the plant. 

Table 7-15 Mean, LNT, LCF Risk per Event (dimensionless) for Residents within the 

Specified Radii of the Peach Bottom Site for the Recreation of the Siting 

Study Using the SST1 Source Term and for the Unmitigated STSBO without 

RCIC blackstart Calculated for SOARCA. CDFs Were Estimated to be 

10-5

/yr and 3x10-7/yr for the SST1 and STSBO without RCIC blackstart 

Source Terms, Respectively

Radius of Circular Area 

(mi) SST1 PB STSBO Ratio SST1 to STSBO

10 7.4x10-3 2.1x10-4 36

20 2.1x10-3 5.7x10-4 4

30 9.1x10-4 3.9x10-4 2

40 5.3x10-4 2.4x10-4 2

50 4.2x10-4 1.9x10-4 2

Table 7-16 is similar to Table 7-15, but compares results calculated with the US background 

(620 mrem/yr) truncation level. The ratio at 10 miles is larger than the one in Table 7-15

because more of the doses exceed the truncation level in the case of the SST1 result. The ratio 

beyond 10 miles does not change significantly.

Table 7-16 Mean, LCF Risk per Event (dimensionless) Using US Background (620 

mrem/yr) Dose-Truncation for Residents within the Specified Radii of the 

Peach Bottom Site for the Recreation of the 1982 Siting Study Using the 

SST1 Source Term and for the Unmitigated STSBO without RCIC 

blackstart Scenario Calculated for SOARCA. CDFs Were Estimated to Be 

10-5

/yr and 3x10-7/yr for the SST1 and STSBO without RCIC blackstart 

Source Terms, Respectively

Radius of Circular Area 

(mi) SST1 PB STSBO Ratio SST1 to STSBO

10 6.2x10-4 1.2x10-5 50

20 8.1x10-4 3.7x10-4 2

30 4.2x10-4 2.4x10-4 2

40 2.8x10-4 1.3x10-4 2

50 2.3x10-4 9.7x10-5 2

Table 7-17 shows the conditional latent cancer fatality risks at three dose truncation levels for the 

recreation of the 1982 Siting Study results. Predicted LCF risks only vary by about an order of 

magnitude or less over the range of dose truncation levels considered. 
7-26

Table 7-17 Mean, LCF Risk per Event (dimensionless) for Three Levels of Dose 

Truncation for Residents within the Specified Radii of the Peach Bottom Site 

for the Recreation of the 1982 Siting Study Using the SST1 Source Term. The 

CDF for This Sequence of the 1982 Siting Study is 10-5/yr

Radius of Circular Area 

(mi) LNT US BGR HPS

10 7.4x10-3 6.2x10-4 7.2x10-4

20 2.1x10-3 8.1x10-4 9.3x10-4

30 9.1x10-4 4.2x10-4 5.0x10-4

40 5.3x10-4 2.8x10-4 3.5x10-4

50 4.2x10-4 2.3x10-4 3.0x10-4

7.3.7 Surface Roughness

All of the SOARCA analyses presented above use a surface roughness length that represents a 

typical value for the US, which is 10 cm. This value was used in the 1982 Siting Study and 

NUREG-1150 and has become a de facto default for most if not all license-related consequence 

analyses, (e.g., SAMA) analyses for license extension. However, this value of surface roughness 

is not necessarily the best choice for all regions of the country. In this section, we examine a 

more site-specific value of surface roughness as a sensitivity analysis to determine whether this 

parameter is significant for estimated risk. 

The effect of increased surface roughness is twofold: It increases vertical mixing of the plume 

and it increases deposition velocities for all aerosol sizes. Both effects are treated in this 

sensitivity analysis and are discussed in the subsequent paragraphs.

Peach Bottom is located on the Susquehanna River near the southern border of Pennsylvania. 

The area surrounding the Peach Bottom Atomic Power Station is characterized by a mosaic of 

forests, farmland mostly used to grow corn, suburban areas, and several stone quarries. 

Farmland and forests are the major land-use categories. Each of these land-use types correspond 

to a typical surface roughness or a range of surface roughness as shown in Table 7-18 [36]. 
7-27

Table 7-18 Surface roughness for various land-use categories for the area surrounding 

the Peach Bottom site

Land-Use Category Surface Roughness (cm)

Farmland recently plowed 1

Farmland with mature corn 10

Suburban housing 5 to 20

Suburban institutional buildings 70

Woodland forests 20 to 100

Stone quarries 50 to 200

Determining the best choice of surface roughness to represent the range of land-use categories is 

not a simple task. The value of 10 cm used in the base case is representative of the corn fields 

that make up a significant fraction of the countryside surrounding the Peach Bottom site. 

Woodland forests also make up a large fraction of the area and have a mean surface roughness of 

about 60 cm. An intermediate choice representing the average between cornfields and woodland 

forest, about 30 to 35 cm, would also be a reasonable choice for this area. For this sensitivity 

analysis, a mean value representing typical woodland forest, 60 cm, was chosen to determine 

whether the results are sensitivity to this input parameter. Results for this sensitivity analysis are 

based on the STSBO without RCIC blackstart, which is the accident scenario with the largest 

consequences of those considered in SOARCA for Peach Bottom.

The effect on vertical mixing has traditionally been modeled by means of a multiplicative factor 

on vertical dispersion. The empirical expression for this factor is the ratio of surface roughness 

at the site in question to a standard value of surface roughness to the 1/5th power. Most of the 

data upon which empirical dispersion models have been based were taken at a site characterized 

by prairie grass [33], which was estimated to have a surface roughness of 3 cm. Thus, the 

empirical equation used to scale vertical dispersion uses the actual surface roughness divided by 

3 cm to the 1/5th power. The standard multiplicative factor corresponding to a 10-cm surface 

roughness is (10 / 3)0.2 = 1.27, which is the value used in all of the calculations presented above. 

A surface roughness length of 60 cm corresponds to a multiplicative factor of 1.82, which was 

used in this sensitivity analysis.

The effect of surface roughness on deposition velocity has been characterized by 

Bixler et al. [34] based on expert elicitation data [29]. Bixler et al. provides a set of correlations 

for estimating deposition velocity as a function of aerosol diameter, wind speed, surface 

roughness, and percentile representing degree of belief by the experts. Here, we use the 

50th percentile from the experts to get a best estimate deposition velocity. The 50th percentile 

correlation is as follows:

ln(vd) = −3.112 + 0.992 ∙ ln�dp� + 0.190 ∙ [ln�dp�]

2 − 0.072 ∙ [ln�dp�]3 + 5.922 ∙ z0

− 6.314 ∙ z0

2 + 0.169 ∙ v

where

vd = deposition velocity (cm/s)

dp = aerosol diameter (μm)

z0 = surface roughness (m)

v = mean wind speed (m/s)
7-28

Table 7-19 shows the aerosol deposition velocities calculated with the above equation that were 

used in this study for each aerosol bin in the MELCOR model. A mean wind speed of 2.2 m/s 

was used to obtain the results in the table. The column of deposition velocities corresponding to 

a surface roughness of 10 cm were used for all of the results shown in the preceding subsections. 

The column of deposition velocities corresponding to a surface roughness of 60 cm were used in 

this sensitivity analysis. Increasing surface roughness from 10 to 60 cm roughly doubles the 

deposition velocity. Because the correlation only extends to aerosol diameters of about 20 μm, 

the same depositions velocities are used for the top two aerosol bins. MELCOR calculates size 

distributions for the aerosols released into the atmosphere as a function of time, and this 

information is used to determine the fraction of the aerosols that belong to each size bin for each 

chemical class. Here, only the deposition velocities are varied to account for the effect of surface 

roughness; the calculated aerosol size distribution is independent of surface roughness and so 

was not varied.

Table 7-19 Deposition Velocities Used for Base Case Calcs and for Surface Roughness 

Sensitivity Study for Each of the Ten Aerosol Bins in the MELCOR Model

Mass Median Aerosol Diameter 

(μm)

Deposition Velocity (cm/s) for Specified 

Surface Roughness

10 cm 60 cm

0.15 0.053 0.11

0.29 0.049 0.10

0.53 0.064 0.14

0.99 0.11 0.23

1.8 0.21 0.45

3.4 0.43 0.92

6.4 0.84 1.8

11.9 1.4 2.9

22.1 1.7 3.7

41.2 1.7 3.7

Table 7-20 compares the base case results presented in Subsection 7.3.3 (10-cm surface 

roughness) with the sensitivity results (60-cm surface roughness). The trends from this table are 

discussed in the following paragraphs.

Table 7-20 Mean, Individual, LCF Risk per Event (Dimensionless) for Residents within 

the Specified Radii of the Peach Bottom Site for the Unmitigated STSBO 

Scenario without RCIC blackstart, which has a Mean CDF of 3x10-7 pry. 

Risks are shown for the base case (10 cm) and the sensitivity case (60 cm)

Radius of 

Circular Area 

(mi)

LNT US BGR HPS

Base 

Case

Increased 

Roughness

Base 

Case

Increased 

Roughness

Base 

Case

Increased 

Roughness

10 2.1x10-4 2.3x10-4 1.2x10-5 1.1x10-5 1.3x10-5 1.2x10-5

20 5.7x10-4 4.6x10-4 3.7x10-4 2.7x10-4 3.7x10-4 2.4x10-4

30 3.9x10-4 3.2x10-4 2.4x10-4 1.7x10-4 2.2x10-4 1.4x10-4

40 2.4x10-4 2.0x10-4 1.3x10-4 9.4x10-5 1.1x10-4 6.4x10-5

50 1.9x10-4 1.7x10-4 9.7x10-5 6.9x10-5 7.3x10-5 4.3x10-5
7-29

A number of factors need to be considered to understand the trends contained in these results. 

For elevated releases, enhanced vertical dispersion increases the ground-level concentrations 

close to the point of release because the plume spreads down to the ground more quickly. The 

MACCS2 results show that ground-level concentrations and resulting doses are greater for the 

sensitivity case within about 4 miles of the plant; at greater distances the ground-level doses are 

smaller for the sensitivity case. At intermediate distances, enhanced vertical dispersion reduces 

ground-level concentrations because the plume spreads more rapidly in the vertical dimension 

and, as a result, becomes more dilute at ground level. At long distances, the ground-level 

concentrations are about the same because the plume becomes well mixed within the mixing 

layer where the plume is confined. 

Compounding the effect of increased surface roughness on vertical dispersion, as described 

above, is the fact that the deposition velocity is also increased. This leads to more deposition at 

shorter distances and faster depletion of the plume. As a result, doses at short distances are 

increased and doses at longer distances are diminished.

The trends for the dose-truncation results are similar to the LNT results, but the reductions in risk 

are generally greater than 20% for the sensitivity case. In contrast to the LNT risk, the risks 

within 10 miles are slightly lower for the sensitivity case than for the base case when using dose 

truncation based on either the US background or HPS Position. 

Comparing the LNT results, the results at intermediate distances are about 20% lower for the 

sensitivity case with increased surface roughness. The risk within 10 miles for the sensitivity 

case is about 10% higher because of the larger ground-level concentrations within 4 miles and 

because of increased deposition onto the ground. The results within 50 miles are reduced by 

only 10%. This is because the plume is well mixed between the ground and the top of the 

mixing layer by 50 miles under many of the prevailing weather conditions.

A general observation based on this sensitivity study is that the specific choice of surface 

roughness only has a relatively modest effect on LNT predictions of risk; it has a larger, but less 

than a factor-of-two, effect for the two dose truncation levels considered here.

7.3.8 Importance of Chemical Classes

Each isotope present in the core of a nuclear reactor contributes to the overall risk of an accident; 

however, the release of some isotopes contributes to risk much more than others. There are three 

reasons some isotopes are more important than others: 

• abundance of an isotope in the inventory in the core at the beginning of an accident,

• release fraction of an isotope into the atmosphere, and

• the dose conversion factors for an isotope, which depends strongly on the type and 

energy of the radiation produced, the half life of the isotope, and for internal pathways, 

the biokinetics of the isotope.
7-30

There are 69 isotopes in the treatment of consequences considered in the MACCS2 analysis, as 

described in Appendix A. These isotopes are grouped into a set of 9 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 both 

reasonable and useful to examine how these same chemical classes influence the evaluation of 

risk.

One approach to estimate the relative importance of each chemical class on risk is to release one 

chemical class at a time and evaluate the fraction of the overall risk that results, where overall 

risk is evaluated by releasing all chemical classes simultaneously. The problem with this 

approach is that the contributions from the individual chemical classes add up to more than the 

overall risk. The difference results from the amount of remedial action that is taken to reduce 

doses to the public. For example, much less remedial action is taken when doses are small, 

which may be the case when only one chemical class is released at a time, than when doses are 

large. Because less remedial action is taken, the contribution of an individual chemical group to 

risk is greater when it is released on its own than when it is part of a larger release. To make the 

fractional contributions from individual chemical groups add to unity, the contribution from a 

single chemical class must be normalized by the sum of the individual contributions of the 

chemical classes rather than the risk calculated for the combined effect of all chemical classes. 

This inherent nonlinearity tends to diminish the effect of the major contributors and exaggerate 

the effect of the minor contributors.

To minimize the effect of the nonlinearities described in the previous paragraph, an alternative 

approach is adopted here. That is to evaluate the contribution of a chemical class by performing 

calculations with all but that one chemical class. The effect of that chemical class is then 

calculated by taking the difference between the risk when all chemical classes are included and 

the risk for all but that one chemical class (i.e., setting the release fractions for that chemical 

class to zero). 

The relative importance of each chemical class was evaluated for all three accident sequences the 

unmitigated LTSBO, the STSBO with RCIC blackstart, and the unmitigated STSBO without 

RCIC blackstart. Results were also calculated for each dose response: LNT, US BGR, and HPS. 

The results for the unmitigated LTSBO scenario for the population within 10 miles are shown in 

Figure 7-9, Figure 7-10, and Figure 7-11. Results at longer distances are shown in subsequent 

figures.

The first of these, Figure 7-9, is for LNT. It shows the importance of each chemical group on 

total risk, on just the emergency-phase risk, and on just the long-term-phase risk. The cesium 

group dominates the total risk and the long-term phase risk, but contributes only a few percent to 

the emergency-phase risk owing to the relatively long half lives of the cesium isotopes (e.g., 

137Cs has a half life of 30 yrs). Tellurium, barium/strontium, iodine, and cerium contribute most 

of the emergency-phase risk owing to the short half lives of the isotopes represented by these 

chemical classes. However, the emergency phase contributes very little to the total risk because 

99.5% of the population within 10 miles evacuate and do not receive any dose during the 

emergency phase.
7-31

Figure 7-9 Percentage contribution to total, emergency-phase, and long-term-phase, 

mean, individual risk for the population within 10 miles by chemical class for 

the Peach Bottom unmitigated LTSBO based on the LNT hypothesis

Figure 7-10 and Figure 7-11 show the total risk contributions of each chemical class for the 

unmitigated LTSBO using dose truncation based on the US BGR and the HPS Position, 

respectively. These plots also show risk contributions to the population living within 10 miles of 

the plant. They only show the total risk contribution because annual doses in the first year are 

combinations of emergency- and long-term phase doses. As a result of the overlapping 

contributions to the first year, the individual contributions of the two phases cannot be easily 

deconvolved from the whole. These figures show that the barium/strontium, cesium, tellurium, 

and iodine chemical classes contribute most of the risk for these dose truncation criteria, 

although the order of importance is different in the two figures. Isotopes with relatively short 

half-lives tend to be more dominate than those with longer half-lives because most of the risk is 

from doses received during the first year for either of the dose-truncation criteria. Longer-term 

annual doses are limited by the habitability criterion to values below the dose-truncation levels.

0%

10%

20%

30%

40%

50%

60%

70%

80%

90%

100%

Xe Cs Ba/Sr I Te Ru Mo Ce La

Percentage of Risk

Radionuclide Group

Total

Emergency Phase

Long-Term Phase
7-32

Figure 7-10 Percentage contribution to total, mean, individual risk for the population 

within 10 miles by chemical class for the Peach Bottom unmitigated LTSBO 

based on US BGR dose truncation

Figure 7-11 Percentage contribution to total, mean, individual risk for the 

population within 10 miles by chemical class for the Peach Bottom 

unmitigated LTSBO based on a truncation level reflecting the HPS 

Position for quantifying health effects

0%

10%

20%

30%

40%

50%

60%

70%

80%

90%

100%

Xe Cs Ba/Sr I Te Ru Mo Ce La

Percentage of Risk

Radionuclide Group

0%

10%

20%

30%

40%

50%

60%

70%

80%

90%

100%

Xe Cs Ba/Sr I Te Ru Mo Ce La

Percentage of Risk

Radionuclide Group
7-33

The following figures, Figures 7-12 through 7-17, are analogous to those immediately above but 

show the relative importance of the chemical classes for the population within 20 and 50 miles. 

The trends are similar, but the emergency phase plays a larger role because significant portions 

of the population do not evacuate before the plume arrives and, thereby, receive a dose during 

the emergency phase. The most important set of chemical classes using the LNT hypothesis is 

cesium, tellurium, barium/strontium, cerium, and iodine in that order. For the two dose 

truncation criteria, cesium is less important because of the relatively long half-lives of the 

dominate isotopes. Generally, the barium/strontium, cesium, tellurium, and iodine classes are 

the most important chemical classes, although the order of importance differs depending on the 

choice of radius and dose truncation criterion.

Figure 7-18 through Figure 7-26 show the importance of the chemical classes for the STSBO 

scenario with RCIC blackstart. These figures are analogous to the ones above. There are a total 

of nine figures showing the three dose truncation levels at three distances. The LNT figures 

show that, again, the cesium class is the dominate contributor for the total and long-term phase 

risk. Following the cesium class are the tellurium, barium/strontium, and cerium classes in terms 

of importance. For the non-LNT assumptions, the chemical classes that contribute most of the 

risk are the barium/strontium, tellurium, iodine, cesium, and cerium classes, although the order 

of importance varies with dose truncation level and distance. The reason that the cesium class is 

less important for either of the dose-truncation levels considered here than for the LNT is the 

same as the one given for the unmitigated LTSBO scenario above.
7-34

Figure 7-12 Percentage contribution to total, emergency-phase, and long-term-phase, 

mean, individual risk for the population within 20 miles by chemical class for 

the Peach Bottom unmitigated LTSBO based on the LNT hypothesis

Figure 7-13 Percentage contribution to total, mean, individual risk for the population 

within 20 miles by chemical class for the Peach Bottom unmitigated LTSBO 

based on US BGR dose truncation

0%

10%

20%

30%

40%

50%

60%

70%

80%

90%

100%

Xe Cs Ba/Sr I Te Ru Mo Ce La

Percentage of Risk

Radionuclide Group

Total

Emergency Phase

Long-Term Phase

0%

10%

20%

30%

40%

50%

60%

70%

80%

90%

100%

Xe Cs Ba/Sr I Te Ru Mo Ce La

Percentage of Risk

Radionuclide Group
7-35

Figure 7-14 Percentage contribution to total, mean, individual risk for the population 

within 20 miles by chemical class for the Peach Bottom unmitigated LTSBO 

based on a truncation level reflecting the HPS Position for quantifying health 

effects

Figure 7-15 Percentage contribution to total, emergency-phase, and long-term-phase, 

mean, individual risk for the population within 50 miles by chemical class for 

the Peach Bottom unmitigated LTSBO based on the LNT hypothesis

0%

10%

20%

30%

40%

50%

60%

70%

80%

90%

100%

Xe Cs Ba/Sr I Te Ru Mo Ce La

Percentage of Risk

Radionuclide Group

0%

10%

20%

30%

40%

50%

60%

70%

80%

90%

100%

Xe Cs Ba/Sr I Te Ru Mo Ce La

Percentage of Risk

Radionuclide Group

Total

Emergency Phase

Long-Term Phase
7-36

Figure 7-16 Percentage contribution to total, mean, individual risk for the population 

within 50 miles by chemical class for the Peach Bottom unmitigated LTSBO 

based on US BGR dose truncation

Figure 7-17 Percentage contribution to total, mean, individual risk for the population 

within 50 miles by chemical class for the Peach Bottom unmitigated LTSBO 

based on a truncation level reflecting the HPS Position for quantifying health 

effects

0%

10%

20%

30%

40%

50%

60%

70%

80%

90%

100%

Xe Cs Ba/Sr I Te Ru Mo Ce La

Percentage of Risk

Radionuclide Group

0%

10%

20%

30%

40%

50%

60%

70%

80%

90%

100%

Xe Cs Ba/Sr I Te Ru Mo Ce La

Percentage of Risk

Radionuclide Group
7-37

Figure 7-18 Percentage contribution to total, emergency-phase, and long-term-phase, 

mean, individual risk for the population within 10 miles by chemical class for 

the Peach Bottom unmitigated STSBO with RCIC blackstart based on the 

LNT hypothesis

Figure 7-19 Percentage contribution to total, mean, individual risk for the population 

within 10 miles by chemical class for the Peach Bottom unmitigated STSBO 

with RCIC blackstart based on US BGR dose truncation

0%

10%

20%

30%

40%

50%

60%

70%

80%

90%

100%

Xe Cs Ba/Sr I Te Ru Mo Ce La

Percentage of Risk

Radionuclide Group

Total

Emergency Phase

Long-Term Phase

0%

10%

20%

30%

40%

50%

60%

70%

80%

90%

100%

Xe Cs Ba/Sr I Te Ru Mo Ce La

Percentage of Risk

Radionuclide Group
7-38

Figure 7-20 Percentage contribution to total, mean, individual risk for the population 

within 10 miles by chemical class for the Peach Bottom unmitigated STSBO 

with RCIC blackstart based on a truncation level reflecting the HPS Position 

for quantifying health effects

Figure 7-21 Percentage contribution to total, emergency-phase, and long-term-phase, 

mean, individual risk for the population within 20 miles by chemical class for 

the Peach Bottom unmitigated STSBO with RCIC blackstart based on the 

LNT hypothesis

0%

10%

20%

30%

40%

50%

60%

70%

80%

90%

100%

Xe Cs Ba/Sr I Te Ru Mo Ce La

Percentage of Risk

Radionuclide Group

0%

10%

20%

30%

40%

50%

60%

70%

80%

90%

100%

Xe Cs Ba/Sr I Te Ru Mo Ce La

Percentage of Risk

Radionuclide Group

Total

Emergency Phase

Long-Term Phase
7-39

Figure 7-22 Percentage contribution to total, mean, individual risk for the population 

within 20 miles by chemical class for the Peach Bottom unmitigated STSBO 

with RCIC blackstart based on US BGR dose truncation

Figure 7-23 Percentage contribution to total, mean, individual risk for the population 

within 20 miles by chemical class for the Peach Bottom unmitigated STSBO 

with RCIC blackstart based on a truncation level reflecting the HPS Position 

for quantifying health effects

0%

10%

20%

30%

40%

50%

60%

70%

80%

90%

100%

Xe Cs Ba/Sr I Te Ru Mo Ce La

Percentage of Risk

Radionuclide Group

0%

10%

20%

30%

40%

50%

60%

70%

80%

90%

100%

Xe Cs Ba/Sr I Te Ru Mo Ce La

Percentage of Risk

Radionuclide Group
7-40

Figure 7-24 Percentage contribution to total, emergency-phase, and long-term-phase, 

mean, individual risk for the population within 50 miles by chemical class for 

the Peach Bottom unmitigated STSBO with RCIC blackstart based on the 

LNT hypothesis

Figure 7-25 Percentage contribution to total, mean, individual risk for the population 

within 50 miles by chemical class for the Peach Bottom unmitigated STSBO 

with RCIC blackstart based on US BGR dose truncation

0%

10%

20%

30%

40%

50%

60%

70%

80%

90%

100%

Xe Cs Ba/Sr I Te Ru Mo Ce La

Percentage of Risk

Radionuclide Group

Total

Emergency Phase

Long-Term Phase

0%

10%

20%

30%

40%

50%

60%

70%

80%

90%

100%

Xe Cs Ba/Sr I Te Ru Mo Ce La

Percentage of Risk

Radionuclide Group
7-41

Figure 7-26 Percentage contribution to total, mean, individual risk for the population 

within 50 miles by chemical class for the Peach Bottom unmitigated STSBO 

with RCIC blackstart based on a truncation level reflecting the HPS Position 

for quantifying health effects

Figure 7-27 through Figure 7-35 show the results for the unmitigated STSBO without RCIC 

blackstart scenario. Comparing these figures with the previous ones for the LTSBO and with the 

STSBO with RCIC blackstart, two features immediately stand out as distinctly different. First, 

the risk is dominated by the emergency phase rather than the long-term phase at distances 

beyond 10 miles. Second and related to the first, the cesium class plays a much smaller role at 

distances of 20 and 50 miles; the cerium and barium/strontium classes are dominant in terms of 

contribution to risk even under the LNT hypothesis. The cesium class continues to dominate the 

risk during the long-term phase under the LNT hypothesis, but the contribution of the long-term 

phase is diminished for this scenario. The diminished contribution of the cesium class in the 

unmitigated STSBO without RCIC blackstart scenario is due to the comparatively large release 

fractions of the barium/strontium and cerium classes, as shown in Table 7-1. The iodine and 

tellurium release fractions are also large (11.5% and 10.4%, respectively), but most of these

releases occur at a later time (about 25 hr for iodine and tellurium as opposed to about 8 hr for 

barium/strontium and cerium), as shown in Figure 5-43 and Figure 5-44. The late releases of 

iodine and tellurium diminish the importance of these chemical groups for this scenario, as 

shown in the following figures.

The release fraction for the cerium class appears small (only 0.7%), as shown in Table 7-1, but 

this class represents a significant fraction of the overall activity in the core at reactor shutdown. 

As a result, even this small release fraction provides about half or more of the total risk for 

distances beyond 10 miles, as shown in the figures below. Cesium remains the dominate 

contributor to the long-term phase, but both the cerium and barium/strontium classes contribute 

to doses during the long-term phase as well as during the emergency phase. 

0%

10%

20%

30%

40%

50%

60%

70%

80%

90%

100%

Xe Cs Ba/Sr I Te Ru Mo Ce La

Percentage of Risk

Radionuclide Group
7-42

The trends for risk within 10 miles are somewhat different than those for the longer distances. 

Whereas the emergency phase in this accident scenario plays a major role beyond the EPZ 

(10 miles), it plays a minor role within the EPZ. This is because nearly all of the population are 

able to evacuate before plume arrival and receive no dose during the emergency phase. Thus, 

even though the emergency phase dominates beyond 10 miles, the long-term phase dominates for 

the population within 10 miles from the plant.

Figure 7-27 Percentage contribution to total, emergency-phase, and long-term-phase, 

mean, individual risk for the population within 10 miles by chemical class for 

the Peach Bottom unmitigated STSBO without RCIC blackstart based on the 

LNT hypothesis

0%

10%

20%

30%

40%

50%

60%

70%

80%

90%

100%

Xe Cs Ba/Sr I Te Ru Mo Ce La

Percentage of Risk 

Radionuclide Group

Total

Emergency Phase

Long-Term Phase
7-43

Figure 7-28 Percentage contribution to total, mean, individual risk for the population 

within 10 miles by chemical class for the Peach Bottom unmitigated STSBO 

without RCIC blackstart based on US BGR dose truncation

Figure 7-29 Percentage contribution to total, mean, individual risk for the population 

within 10 miles by chemical class for the Peach Bottom unmitigated STSBO 

without RCIC blackstart based on a truncation level reflecting the HPS 

Position for quantifying health effects

0%

10%

20%

30%

40%

50%

60%

70%

80%

90%

100%

Xe Cs Ba/Sr I Te Ru Mo Ce La

Percentage of Risk

Radionuclide Group

0%

10%

20%

30%

40%

50%

60%

70%

80%

90%

100%

Xe Cs Ba/Sr I Te Ru Mo Ce La

Percentage of Risk

Radionuclide Group
7-44

Figure 7-30 Percentage contribution to total, emergency-phase, and long-term-phase, 

mean, individual risk for the population within 20 miles by chemical class for 

the Peach Bottom unmitigated STSBO without RCIC blackstart based on the 

LNT hypothesis

Figure 7-31 Percentage contribution to total, mean, individual risk for the population 

within 20 miles by chemical class for the Peach Bottom unmitigated STSBO 

without RCIC blackstart based on US BGR dose truncation

0%

10%

20%

30%

40%

50%

60%

70%

80%

90%

100%

Xe Cs Ba/Sr I Te Ru Mo Ce La

Percentage of Risk

Radionuclide Group

Total

Emergency Phase

Long-Term Phase

0%

10%

20%

30%

40%

50%

60%

70%

80%

90%

100%

Xe Cs Ba/Sr I Te Ru Mo Ce La

Percentage of Risk

Radionuclide Group
7-45

Figure 7-32 Percentage contribution to total, mean, individual risk for the population 

within 20 miles by chemical class for the Peach Bottom unmitigated STSBO 

without RCIC blackstart based on a truncation level reflecting the HPS 

Position for quantifying health effects

Figure 7-33 Percentage contribution to total, emergency-phase, and long-term-phase, 

mean, individual risk for population within 50 miles by chemical class for 

unmitigated STSBO w/o RCIC blackstart based on LNT hypothesis

0%

10%

20%

30%

40%

50%

60%

70%

80%

90%

100%

Xe Cs Ba/Sr I Te Ru Mo Ce La

Percentage of Risk

Radionuclide Group

0%

10%

20%

30%

40%

50%

60%

70%

80%

90%

100%

Xe Cs Ba/Sr I Te Ru Mo Ce La

Percentage of Risk

Radionuclide Group

Total

Emergency Phase

Long-Term Phase
7-46

Figure 7-34 Percentage contribution to total, mean, individual risk for the population 

within 50 miles by chemical class for the Peach Bottom unmitigated STSBO 

without RCIC blackstart based on US BGR dose truncation

Figure 7-35 Percentage contribution to total, mean, individual risk for the population 

within 50 miles by chemical class for the Peach Bottom unmitigated STSBO 

without RCIC blackstart based on a truncation level reflecting the HPS 

Position for quantifying health effects

0%

10%

20%

30%

40%

50%

60%

70%

80%

90%

100%

Xe Cs Ba/Sr I Te Ru Mo Ce La

Percentage of Risk

Radionuclide Group

0%

10%

20%

30%

40%

50%

60%

70%

80%

90%

100%

Xe Cs Ba/Sr I Te Ru Mo Ce La

Percentage of Risk

Radionuclide Group
8-1

8.0 REFERENCES

[1] Jones, J.A., “Summary Report: Peer Review of the State-of-the Art Reactor 

Consequence Analyses (SOARCA) Project. SAND2012-3714. Sandia National 

Laboratories: Albuquerque, NM. May 2012. [ML121250030].

[2] Carbajo, J.J. “Severe Accident Source Term Characteristics for Selected Peach 

Bottom Sequences Predicted by the MELCOR Code.” NUREG/CR-5942, Oak Ridge 

National Laboratory: Oak Ridge, TN. July 1993.

[3] Peach Bottom Atomic Power Station, Special Event Procedure – SE-11.

[4] Peach Bottom Atomic Power Station, Trip Procedure T-101.

[5] D.A. Powers, et al., “Accident Source Terms for Light-Water Nuclear Power Plants 

Using High-Burnup or MOX Fuel,” SAND2011-0128, Sandia National Laboratories: 

Albuquerque, NM. January 2011.

[6] Soffer, L, et al., “Accident Source Terms for Light-Water Nuclear Power Plants,” 

NUREG-1465, Nuclear Regulatory Commission: Washington, DC. February, 1995.

[7] Gauntt, R.O., et al., "MELCOR Computer Code Manuals, Vol. 1: Primer and User's 

Guide, Version 1.8.6," NUREG/CR 6119, Vol. 1, Rev. 3, Nuclear Regulatory 

Commission: Washington, DC. 2005.

[8] Laur, M.N., "Meeting with Sandia National Laboratories and an Expert Panel on 

MELCOR/MACCS Codes in Support of the State of the Art Reactor Consequence 

Analysis Project," Memo to J.T. Yerokun, Agency Document Access and 

Management System Accession Number ML062500078, Nuclear Regulatory 

Commission: Washington, DC. September, 2006.

[9] Nuclear Regulatory Commission, "Severe Accident Risks: An Assessment for Five 

U.S. Nuclear Power Plants," NUREG 1150, Nuclear Regulatory Commission: 

Washington, DC. 1990.

[10] Gauntt, R.O., “Synthesis of VERCORS and Phebus Data in Severe Accident Codes 

and Applications,” SAND2010-1633, Sandia National Laboratories: Albuquerque, 

NM. April 2010.

[11] Harper, F. T., et al., “Evaluation of Severe Accident Risks: Quantification of Major 

Input Parameters, Experts’ Determination of Containment Loads and Molten Core 

Containment Interaction Issues,” NUREG/CR-4551, Sandia National Laboratories: 

Albuquerque, NM. April 1991.
8-2

[12] Theofanous, T.G., et al., “The Probability of Liner Failure in a Mark I Containment,”

NUREG/CR-5423, Nuclear Regulatory Commission: Washington DC. 1989.

[13] Theofanous, T.G., et al., “The Probability of Mark I Failure by Melt –Attack of the 

Liner,” NUREG/CR-6025, Nuclear Regulatory Commission: Washington DC. 

November 1993.

[14] Oak Ridge National Laboratory, “SCALE: A Modular Code System for Performing 

Standardized Computer Analyses for Licensing Evaluations,” ORNL/TM-2005/39, 

Version 5.1, Vols. I-III, Oak Ridge National Laboratory: Oak Ridge, TN. 

November 2006.

[15] Philadelphia Electric Company, “Individual Plant Examination, Peach Bottom 

Atomic Power Station, Units 2 and 3,” August 1992.

[16] T.Y. Chu, et al., “Lower Head Failure Experiments and Analyses,” NUREG/CR5582, SAND98-2047, Sandia National Laboratories: Albuquerque, NM. October 

1998.

[17] J.L. Rempe, et al., “Light Water Reactor Lower Head Failure Analysis,” 

NUREG/CR-5642, EGG-2618, Idaho National Engineering Laboratory: Idaho Falls, 

ID. September 1993.

[18] Eide, S.A., et al., “Industry-Average Performance for Components and Initiating 

Events at U.S. Commercial Nuclear Power Plants,” NUREG/CR-6928, Idaho 

National Laboratory: Idaho Falls, ID. February 2007.

[19] Peach Bottom Atomic Power Station Procedure, “Reactor Pressure Vessel 

Reassembly.”

[20] Personal communication from Thomas R. Loomis of Exelon Corporation to Abdul 

Sheikh of the Nuclear Regulatory Commission about torque value for the 2 1⁄2 inch

diameter reactor head flange bolts, October 31, 2006.

[21] Nuclear Regulatory Commission, “Criteria for Preparation and Evaluation of 

Radiological Emergency Response Plans and Preparedness in Support of Nuclear 

Power Plants.” NUREG-0654/FEMA-REP-1, Rev. 1, Nuclear Regulatory 

Commission: Washington DC. November 1980.

[22] Nuclear Regulatory Commission, "Identification and Analysis of Factors Affecting 

Emergency Evacuations," NUREG/CR-6864, Nuclear Regulatory Commission:

Washington DC. January 2005. 

[23] Nuclear Regulatory Commission, "Review of NUREG-0654, Supplement 3, “Criteria 

for Protective Action Recommendations for Severe Accidents," NUREG/CR-6953, 

Vol. I, Nuclear Regulatory Commission: Washington D.C. December 2007.
8-3

[24] Nuclear Regulatory Commission, "Development of Evacuation Time Estimate 

Studies for Nuclear Power Plants," NUREG/CR-6863, Nuclear Regulatory 

Commission: Washington D.C. January 2005.

[25] Oak Ridge National Laboratory. “Oak Ridge Evacuation Modeling System 

(OREMS),” Oak Ridge National Laboratory: Oak Ridge, TN. July 2003.

[26] Rogers, G.O., et al., “Evaluating Protective Actions for Chemical Agent 

Emergencies,” ORNL-6615, Oak Ridge National Laboratory: Oak Ridge, TN. April 

1990.

[27] Chanin, D.I., and M.L. Young, “Code Manual for MACCS2: Volume 1, User’s 

Guide,” NUREG/CR-6613, Nuclear Regulatory Commission: Washington D.C. 

1998.

[28] Aldrich, D.C., et al., “Technical Guidance for Siting Criteria Development,” 

NUREG/CR-2239, Nuclear Regulatory Commission: Washington DC. 1982.

[29] Harper, F. T., et al., “Probabilistic Accident Consequence Uncertainty Analysis, 

Dispersion and Deposition Uncertainty Analysis,” NUREG/CR-6244, Nuclear 

Regulatory Commission: Washington DC. 1994.

[30] Transportation Research Board, “Highway Capacity Manual,” National Research 

Council: Washington DC. 2000.

[31] Nuclear Regulatory Commission, “Review of NUREG-0654, Supplement 3, ‘Criteria 

for Protective Action Recommendations for Severe Accidents’ – Focus Group and 

Telephone Survey,” NUREG/CR-6953, Volume 2, Nuclear Regulatory Commission: 

Washington DC. October 2008.

[32] Wheeler, T., G. Wyss, and F. Harper, “Cassini Spacecraft Uncertainty Analysis Data 

and Methodology Review and Update Volume 1: Updated Parameter Uncertainty 

Models for the Consequence Analysis,” SAND2000-2719/1, Sandia National 

Laboratories: Albuquerque, NM. November, 2000.

[33] Haugen, D. A. (Ed.), “Project Prairie Grass: A Field Program in Diffusion.” No. 59, 

Vol. III, Report AFCRC-TR-58-235, Air Force Cambridge Research Center. 1959.

[34] Bixler, N. E., E. Clauss, C.W. Morrow, and J.A. Mitchell, “Evaluation of 

Distributions Representing Important Non-Site-Specific Parameters in Offsite 

Consequence Analysis,” to be published as a NUREG Report, Nuclear Regulatory 

Commission: Washington DC. 2011.
8-4

[35] Miller, C. L., et al., “Recommendations for Enhancing Reactor Safety in the 21st

Century – The Near-Term Task Force Review of Insights From the Fukushima DaiIchi Accident,” Nuclear Regulatory Commission: Washington DC. July 2011.

[36] Whelan, G., D.L. Strenge, J.G. Droppo Jr., B.L. Steelman, and J.W. Buck, “The 

Remedial Action Priority System (RAPS): Mathematical Formulations,” 

DOE/RL-87-09, Pacific Northwest Laboratory: Richland, WA. August 1987.

[37] Eckerman, K.F. “Radiation Dose and Health Risk Estimation: Technical Basis for the 

State-of-the-Art Reactor Consequence Analysis Project.” Oak Ridge National 

Laboratory. January 30, 2012. 
APPENDIX A

PEACH BOTTOM RADIONUCLIDE INVENTORY

A-1

APPENDIX A: PEACH BOTTOM RADIONUCLIDE INVENTORY

The following tables summarize the radionuclide core inventory for the Peach Bottom plant at 

the time of shutdown. This isotopic inventory was used in each of the accident progression 

scenarios considered in this report. This inventory was computed using SCALE to model the 

specific fuel management strategy used at Peach Bottom for this unit. The SCALE model 

assumed that the accident occurs mid-cycle and that the peak fuel rod burnup is 49 MWd/kg fuel. 

Both radial and axial variations in burnup were modeled with SCALE. The inventory given 

below was calculated by integrating the isotopic inventory over the whole core.

Table A-1 Peach Bottom radionuclide core inventory and class definition

Radionuclide Class 

Name

Representative

Element

Member 

Elements

Radioactive

Mass (kg)

Noble Gas Xe He, Ne, Ar, Kr, 

Xe, Rn, H, N 531.7

Alkali Metals Cs Li, Na, K, Rb, Cs, 

Fr, Cu 323.0

Alkaline Earths Ba Be, Mg, Ca, Sr, 

Ba, Ra, Es, Fm 235.6

Halogens I F, Cl, Br, I, At 19.9

Chalcogens Te O, S, Se, Te, Po 49.1

Platinoids Ru Ru, Rh, Pd, Re, 

Os, Ir, Pt, Au, Ni 342.8

Early Transition 

Elements Mo

V, Cr, Fe, Co, 

Mn, Nb, Mo, Tc, 

Ta, W

400.2

Tetravalent Ce Ti, Zr, Hf, Ce, Th, 

Pa, Np, Pu, C 1,555.5

Trivalents La

Gd, Tb, Dy, Ho, 

Er, Tm, Yb, Lu, 

Am, Cm, Bk, Cf

1793.7

Uranium U U 132,794.0

More Volatile Main 

Group Cd Cd, Hg, Zn, As, Sb, Pb, Tl, Bi 6.6

Less Volatile Main 

Group Sn Ga, Ge, In, Sn, Ag 9.6
A-2

Table A-2 Peach Bottom noble gas radionuclide class specific isotopic activity at the

time of reactor shutdown

Isotope Activity 

(Bq)

Kr-85 3.79E+16

Kr-85m 1.03E+18

Kr-87 2.05E+18

Kr-88 2.77E+18

Xe-133 7.02E+18

Xe-135 2.58E+18

Xe-135m 1.43E+18

Table A-3 Peach Bottom alkali metals radionuclide class specific isotopic activity at the 

time of reactor shutdown

Isotope Activity 

(Bq)

Cs-134 3.61E+17

Cs-136 1.43E+17

Cs-137 3.74E+17

Rb-86 4.38E+15

Rb-88 2.80E+18

Table A-4 Peach Bottom alkali earths radionuclide class specific isotopic activity at the 

time of reactor shutdown

Isotope Activity 

(Bq)

Ba-139 6.48E+18

Ba-140 6.27E+18

Sr-89 3.79E+18

Sr-90 2.98E+17

Sr-91 4.77E+18

Sr-92 5.02E+18

Ba-137m 3.55E+17

Table A-5 Peach Bottom halogen radionuclide class specific isotopic activity at the time 

of reactor shutdown

Isotope Activity 

(Bq)

I-131 3.38E+18

I-132 4.99E+18

I-133 7.15E+18

I-134 8.14E+18

I-135 6.80E+18
A-3

Table A-6 Peach Bottom chalcogen radionuclide class specific isotopic activity at the 

time of reactor shutdown

Isotope Activity 

(Bq)

Te-127 2.71E+17

Te-127m 4.33E+16

Te-129 8.17E+17

Te-129m 1.55E+17

Te-131m 6.03E+17

Te-132 4.85E+18

Te-131 2.89E+18

Table A-7 Peach Bottom platinoid radionuclide class specific isotopic activity at the 

time of reactor shutdown

Isotope Activity 

(Bq)

Rh-105 2.77E+18

Ru-103 4.83E+18

Ru-105 3.03E+18

Ru-106 1.31E+18

Rh-103m 4.82E+18

Rh-106 1.44E+18

Table A-8 Peach Bottom early transition element radionuclide class specific isotopic 

activity at the time of reactor shutdown

Isotope Activity 

(Bq)

Nb-95 6.07E+18

Co-58 0.00E+00

Co-60 0.00E+00

Mo-99 6.52E+18

Tc-99m 5.83E+18

Nb-97 6.11E+18

Nb-97m 5.77E+18
A-4

Table A-9 Peach Bottom tetravalent radionuclide class specific isotopic activity at the 

time of reactor shutdown

Isotope Activity 

(Bq)

Ce-141 5.89E+18

Ce-143 5.64E+18

Ce-144 4.19E+18

Np-239 5.61E+19

Pu-238 6.78E+15

Pu-239 1.37E+15

Pu-240 1.13E+15

Pu-241 3.87E+17

Zr-95 6.11E+18

Zr-97 6.08E+18

Table A-10 Peach Bottom trivalent radionuclide class specific isotopic activity at the time 

of reactor shutdown

Isotope Activity 

(Bq)

Am-241 5.23E+14

Cm-242 9.57E+16

Cm-244 4.70E+15

La-140 6.48E+18

La-141 5.86E+18

La-142 5.70E+18

Nd-147 2.32E+18

Pr-143 5.55E+18

Y-90 3.03E+17

Y-91 4.82E+18

Y-92 5.05E+18

Y-93 5.58E+18

Y-91m 2.75E+18

Pr-144 4.20E+18

Pr-144m 5.85E+16
APPENDIX B

INPUT PARAMETERS FOR CONSEQUENCE 

ANALYSIS

B-1

APPENDIX B: INPUT PARAMETERS FOR CONSEQUENCE ANALYSIS

The input parameters used for the unmitigated LTSBO scenario and the STSBO with and

without RCIC blackstart are shown in this appendix in tabular form. Table B-1 contains the 

more general ATMOS input parameters used for these three scenarios. Table B-2 through 

Table B-4 contains specific inputs related to the source terms that were extracted from MELCOR 

results via the MELMACCS code. Table B-5 contains general EARLY input parameters. 

Table B-6 and Table B-7 contain parameters associated with the network evacuation model that 

was used to treat emergency response. Table B-8 contains the CHRONC input parameters.

Table B-1 ATMOS Input Parameters Used in the Peach Bottom Unmitigated LTSBO 

and STSBO with and without RCIC Blackstart scenarios

Variable Description LTSBO STSBO with 

RCIC Blackstart

STSBO without 

RCIC Blackstart

APLFRC Method of Applying Release 

Fraction PARENT PARENT PARENT

ATNAM1 Title Describing the 

ATMOS Assumptions

SOARCA PB Source 

Term Long-Term 

SBO

SOARCA PB Source 

Term Short-Term 

SBO

SOARCA PB Source 

Term Short-Term 

SBO

ATNAM2 Title Describing the Source 

Term

Peach Bottom source 

term for long term 

station blackout.

Peach Bottom source 

term for short term 

station blackout.

Peach Bottom source 

term for short term 

station blackout.

BNDMXH Boundary Weather Mixing 

Layer Height 1000 1000 1000

BNDRAN Boundary Weather Rain 

Rate 5 5 5

BNDWND Boundary Wind Speed 2.2 2.2 2.2

BRKPNT Breakpoint Time for Plume 

Meander 3600 3600 3600

BUILDH Building Height for all 

Plume Segments 50 50 50

CORINV Isotopic Inventory at Time 

of Reactor Shutdown

from MELMACCS 

(see Appendix A)

from MELMACCS 

(see Appendix A)

from MELMACCS 

(see Appendix A)

CORSCA Linear Scaling Factor on 

Core Inventory 1 1 1

CWASH1 Linear Coefficient for 

Washout 1.89E-05 1.89E-05 1.89E-05

CWASH2 Exponential Term for 

Washout 0.664 0.664 0.664

CYSIGA Linear Coefficient for 

sigma-y

Stability Class A 0.7507 0.7507 0.7507

Stability Class B 0.7507 0.7507 0.7507

Stability Class C 0.4063 0.4063 0.4063

Stability Class D 0.2779 0.2779 0.2779

Stability Class E 0.2158 0.2158 0.2158

Stability Class F 0.2158 0.2158 0.2158
B-2

Variable Description LTSBO STSBO with RCIC 

Blackstart

STSBO without 

RCIC Blackstart

CYSIGB Exponential Term

for sigma-y

Stability Class A 0.866 0.866 0.866

Stability Class B 0.866 0.866 0.866

Stability Class C 0.865 0.865 0.865

Stability Class D 0.881 0.881 0.881

Stability Class E 0.866 0.866 0.866

Stability Class F 0.866 0.866 0.866

CZSIGA Linear Coefficient for 

sigma-z

Stability Class A 0.0361 0.0361 0.0361

Stability Class B 0.0361 0.0361 0.0361

Stability Class C 0.2036 0.2036 0.2036

Stability Class D 0.2636 0.2636 0.2636

Stability Class E 0.2463 0.2463 0.2463

Stability Class F 0.2463 0.2463 0.2463

CZSIGB Exponential Term

for sigma-z

Stability Class A 1.277 1.277 1.277

Stability Class B 1.277 1.277 1.277

Stability Class C 0.859 0.859 0.859

Stability Class D 0.751 0.751 0.751

Stability Class E 0.619 0.619 0.619

Stability Class F 0.619 0.619 0.619

DISPMD Dispersion Model Flag LRDIST LRDIST LRDIST

DRYDEP Dry Deposition Flag

Xe = .FALSE.

Other Groups = 

.TRUE.

Xe = .FALSE.

Other Groups = 

.TRUE.

Xe = .FALSE.

Other Groups = 

.TRUE.

ENDAT1 Control flag indicating only 

ATMOS is to be run .FALSE. .FALSE. .FALSE.

GRPNAM

Names of the Chemical 

Classes (Used by 

WinMACCS)

Chemical Class 1 Xe Xe Xe

Chemical Class 2 Cs Cs Cs

Chemical Class 3 Ba Ba Ba

Chemical Class 4 I I I

Chemical Class 5 Te Te Te

Chemical Class 6 Ru Ru Ru

Chemical Class 7 Mo Mo Mo

Chemical Class 8 Ce Ce Ce

Chemical Class 9 La La La

IBDSTB Boundary Weather Stability 

Class Index 4 4 4

IDEBUG Debug Switch for Extra 

Debugging Print 0 0 0
B-3

Variable Description LTSBO STSBO with RCIC 

Blackstart

STSBO without 

RCIC Blackstart

IGROUP Definition of Radionuclide 

Group Numbers 1 = Xe 1 = Xe 1 = Xe

2 = Cs 2 = Cs 2 = Cs

3 = Ba 3 = Ba 3 = Ba

4 = I 4 = I 4 = I

5 = Te 5 = Te 5 = Te

6 = Ru 6 = Ru 6 = Ru

7 = Mo 7 = Mo 7 = Mo

8 = Ce 8 = Ce 8 = Ce

9 = La 9 = La 9 = La

INWGHT

Number of Samples for Each 

Bin Used for Nonuniform 

Weather Bin Sampling

Bin 1 71 71 71

Bin 2 42 42 42

Bin 3 12 12 12

Bin 4 52 52 52

Bin 5 57 57 57

Bin 6 74 74 74

Bin 7 21 21 21

Bin 8 12 12 12

Bin 9 49 49 49

Bin 10 103 103 103

Bin 11 77 77 77

Bin 12 35 35 35

Bin 13 51 51 51

Bin 14 75 75 75

Bin 15 14 14 14

Bin 16 4 4 4

Bin 17 44 44 44

Bin 18 12 12 12

Bin 19 17 17 17

Bin 20 24 24 24

Bin 21 24 24 24

Bin 22 12 12 12

Bin 23 4 4 4

Bin 24 8 8 8

Bin 25 12 12 12

Bin 26 12 12 12

Bin 27 12 12 12

Bin 28 1 1 1

Bin 29 3 3 3

Bin 30 5 5 5

Bin 31 4 4 4

Bin 32 12 12 12

Bin 33 1 1 1

Bin 34 7 7 7

Bin 35 9 9 9

Bin 36 12 12 12
B-4

Variable Description LTSBO STSBO with RCIC 

Blackstart

STSBO without 

RCIC Blackstart

IRSEED Seed for Random Number 

Generator 79 79 79

LATITU Latitude of Power Plant 39 ̊ 45' 32" 39 ̊ 45' 32" 39 ̊ 45' 32"

LIMSPA Last Interval for Measured 

Weather 25 25 25

LONGIT Longitude of Power Plant 76 ̊ 16' 09" 76 ̊ 16' 09" 76 ̊ 16' 09"

MAXGRP Number of Radionuclide 

Groups 9 9 9

MAXHGT Flag for Mixing Height DAY_AND_NIGHT DAY_AND_NIGHT DAY_AND_NIGHT

MAXRIS Selection of Risk Dominant 

Plume 2 1 9

METCOD Meteorological Sampling 

Option Code 2 2 2

MNDMOD Plume Meander Model Flag OFF OFF OFF

NAMSTB List of Pseudostable 

Nuclides

Isotope 1 I-129 I-129 I-129

Isotope 2 Xe-131m Xe-131m Xe-131m

Isotope 3 Xe-133m Xe-133m Xe-133m

Isotope 4 Cs-135 Cs-135 Cs-135

Isotope 5 Sm-147 Sm-147 Sm-147

Isotope 6 U-234 U-234 U-234

Isotope 7 U-235 U-235 U-235

Isotope 8 U-236 U-236 U-236

Isotope 9 U-237 U-237 U-237

Isotope 10 Np-237 Np-237 Np-237

Isotope 11 Rb-87 Rb-87 Rb-87

Isotope 12 Zr-93 Zr-93 Zr-93

Isotope 13 Nb-93m Nb-93m Nb-93m

Isotope 14 Nb-95m Nb-95m Nb-95m

Isotope 15 Tc-99 Tc-99 Tc-99

Isotope 16 Pm-147 Pm-147 Pm-147

NPSGRP Number of Particle Size 

Groups 10 10 10

NRINTN Number of Rain Intensity 

Breakpoints 3 3 3

NRNINT Number of Rain Distance 

Intervals 5 5 5

NSBINS Number of Weather Bins to 

Sample 36 36 36

NUCNAM Radionuclide Names See Appendix A See Appendix A See Appendix A

NUCOUT Radionuclide Used in 

Dispersion Print Cs-137 Cs-137 Cs-137

NUMCOR Number of Compass Sectors 

in the Grid 64 64 64

NUMISO Number of Radionuclides 69 69 69

NUMRAD Number of Radial Spatial 

Intervals 26 26 26

NUMREL Number of Released Plume 

Segments 29 33 86
B-5

Variable Description LTSBO STSBO with RCIC 

Blackstart

STSBO without 

RCIC Blackstart

NUMSTB Number of Defined 

Pseudostable Radionuclides 16 16 16

OALARM Time to Reach General 

Emergency Conditions 0 0 0

PDELAY Plume Release Times MELMACCS Data 

(See Table B-2)

MELMACCS Data 

(See Table B-2)

MELMACCS Data 

(See Table B-2)

PLHEAT Plume Heat Contents MELMACCS Data 

(See Table B-2)

MELMACCS Data 

(See Table B-2)

MELMACCS Data 

(See Table B-2)

PLHITE Plume Release Heights MELMACCS Data 

(See Table B-2)

MELMACCS Data 

(See Table B-2)

MELMACCS Data 

(See Table B-2)

PLMDEN Plume Mass Density MELMACCS Data 

(See Table A.2-2)

MELMACCS Data 

(See Table B-2)

MELMACCS Data 

(See Table B-2)

PLMFLA Plume Mass Flow Rate MELMACCS Data 

(See Table B-2)

MELMACCS Data 

(See Table B-2)

MELMACCS Data 

(See Table B-2)

PLMMOD Flag for Plume Rise Input 

Option DENSITY DENSITY DENSITY

PLUDUR Plume Segment Durations MELMACCS Data 

(See Table B-2)

MELMACCS Data 

(See Table B-2)

MELMACCS Data 

(See Table B-2)

PSDIST Particle Size Distribution by 

Group

MELMACCS Data 

(See Table B-3)

MELMACCS Data 

(See Table B-3)

MELMACCS Data 

(See Table B-3)

REFTIM Plume Reference Time Point 0. for first

0.5 for subsequent

0. for first

0.5 for subsequent

0. for first

0.5 for subsequent

RELFRC Release Fractions of the 

Source Term

MELMACCS Data 

(See Table B-4)

MELMACCS Data 

(See Table B-4)

MELMACCS Data 

(See Table B-4)

RNDSTS Endpoints of Rain Distance 

Intervals

Interval 1 3.22 3.22 3.22

Interval 2 5.63 5.63 5.63

Interval 3 11.27 11.27 11.27

Interval 4 20.92 20.92 20.92

Interval 5 32.19 32.19 32.19

RNRATE Rain Intensity Breakpoints 

for Weather Binning

Intensity 1 2 2 2

Intensity 2 4 4 4

Intensity 3 6 6 6

SCLADP Scaling Factor for A-D 

Plume Rise 1.0 1.0 1.0

SCLCRW Scaling Factor for Critical 

Wind Speed 1.0 1.0 1.0

SCLEFP Scaling Factor for E-F 

Plume Rise 1.0 1.0 1.0

SIGYINIT Initial Sigma-y for All 

Plume Segments 11.6 11.6 11.6

SIGYINIT Initial Sigma-z for All 

Plume Segments 23.3 23.3 23.3

SPAEND Radial distances for grid 

boundaries

Ring 1 0.16 0.16 0.16

Ring 2 0.52 0.52 0.52
B-6

Variable Description LTSBO STSBO with RCIC 

Blackstart

STSBO without 

RCIC Blackstart

SPAEND Ring 3 1.21 1.21 1.21

Ring 4 1.61 1.61 1.61

Ring 5 2.13 2.13 2.13

Ring 6 3.22 3.22 3.22

Ring 7 4.02 4.02 4.02

Ring 8 4.83 4.83 4.83

Ring 9 5.63 5.63 5.63

Ring 10 8.05 8.05 8.05

Ring 11 11.27 11.27 11.27

Ring 12 16.09 16.09 16.09

Ring 13 20.92 20.92 20.92

Ring 14 25.75 25.75 25.75

Ring 15 32.19 32.19 32.19

Ring 16 40.23 40.23 40.23

Ring 17 48.28 48.28 48.28

Ring 18 64.37 64.37 64.37

Ring 19 80.47 80.47 80.47

Ring 20 112.65 112.65 112.65

Ring 21 160.93 160.93 160.93

Ring 22 241.14 241.14 241.14

Ring 23 321.87 321.87 321.87

Ring 24 563.27 563.27 563.27

Ring 25 804.67 804.67 804.67

Ring 26 1609.34 1609.34 1609.34

TIMBAS Time Base for Plume 

Expansion Factor 36000 36000 36000

VDEPOS Dry Deposition Velocities

Aerosol Bin 1 5.35E-04 5.35E-04 5.35E-04

Aerosol Bin 2 4.91E-04 4.91E-04 4.91E-04

Aerosol Bin 3 6.43E-04 6.43E-04 6.43E-04

Aerosol Bin 4 1.08E-03 1.08E-03 1.08E-03

Aerosol Bin 5 2.12E-03 2.12E-03 2.12E-03

Aerosol Bin 6 4.34E-03 4.34E-03 4.34E-03

Aerosol Bin 7 8.37E-03 8.37E-03 8.37E-03

Aerosol Bin 8 1.37E-02 1.37E-02 1.37E-02

Aerosol Bin 9 1.70E-02 1.70E-02 1.70E-02

Aerosol Bin 10 1.70E-02 1.70E-02 1.70E-02

WETDEP Wet Deposition Flag

Xe = .FALSE.

Other groups = 

.TRUE.

Xe = .FALSE.

Other groups = 

.TRUE.

Xe = .FALSE.

Other groups = 

.TRUE.

XPFAC1 Base Time for Meander 

Expansion Factor 0.01 0.01 0.01

XPFAC2 Breakpoint for Expansion 

Factor Model 0.01 0.01 0.01

YSCALE Scale Factor for Horizontal 

Dispersion 1 1 1

ZSCALE Scale Factor for Vertical 

Dispersion 1.27 1.27 1.27
B-7

Table B-2 Plume Parameters Used in the Peach Bottom Unmitigated LTSBO and 

STSBO with and without RCIC Blackstart scenarios

Peach Bottom LTSBO

Plume 

Segment PDELAY PLHEAT PLHITE PLMDEN PLMFLA PLUDUR

1 7.20E+04 9.18E+06 3.96E+01 9.28E-01 1.46E+02 3.60E+03

2 7.22E+04 6.96E+05 4.00E+00 5.44E-01 6.87E+00 3.66E+03

3 7.56E+04 4.87E+06 3.96E+01 9.84E-01 1.40E+02 3.60E+03

4 7.92E+04 4.37E+06 3.96E+01 9.96E-01 1.35E+02 3.60E+03

5 8.28E+04 3.68E+06 3.96E+01 1.01E+00 1.29E+02 3.74E+03

6 8.66E+04 3.08E+06 3.96E+01 1.03E+00 1.20E+02 3.60E+03

7 9.02E+04 2.66E+06 3.96E+01 1.04E+00 1.14E+02 3.60E+03

8 9.38E+04 2.21E+06 3.96E+01 1.06E+00 1.08E+02 3.60E+03

9 9.74E+04 1.94E+06 3.96E+01 1.06E+00 1.03E+02 3.60E+03

10 1.01E+05 1.75E+06 3.96E+01 1.07E+00 9.96E+01 3.60E+03

11 1.05E+05 1.58E+06 3.96E+01 1.08E+00 9.63E+01 3.60E+03

12 1.08E+05 1.48E+06 3.96E+01 1.08E+00 9.42E+01 3.60E+03

13 1.12E+05 1.45E+06 3.96E+01 1.08E+00 9.32E+01 3.60E+03

14 1.15E+05 1.40E+06 3.96E+01 1.08E+00 9.20E+01 3.60E+03

15 1.19E+05 1.36E+06 3.96E+01 1.09E+00 9.10E+01 3.60E+03

16 1.23E+05 1.33E+06 3.96E+01 1.09E+00 9.01E+01 3.60E+03

17 1.26E+05 1.30E+06 3.96E+01 1.09E+00 8.95E+01 3.60E+03

18 1.30E+05 1.28E+06 3.96E+01 1.09E+00 8.90E+01 3.60E+03

19 1.33E+05 1.27E+06 3.96E+01 1.09E+00 8.86E+01 3.60E+03

20 1.37E+05 1.26E+06 3.96E+01 1.09E+00 8.83E+01 3.60E+03

21 1.41E+05 1.25E+06 3.96E+01 1.09E+00 8.81E+01 3.60E+03

22 1.44E+05 1.25E+06 3.96E+01 1.09E+00 8.79E+01 3.60E+03

23 1.48E+05 1.25E+06 3.96E+01 1.09E+00 8.79E+01 3.60E+03

24 1.51E+05 1.24E+06 3.96E+01 1.09E+00 8.78E+01 3.60E+03

25 1.55E+05 1.24E+06 3.96E+01 1.09E+00 8.77E+01 3.60E+03

26 1.59E+05 1.24E+06 3.96E+01 1.09E+00 8.76E+01 3.60E+03

27 1.62E+05 1.24E+06 3.96E+01 1.09E+00 8.75E+01 3.60E+03

28 1.66E+05 1.24E+06 3.96E+01 1.09E+00 8.74E+01 3.60E+03

29 1.69E+05 1.24E+06 3.96E+01 1.09E+00 8.73E+01 3.42E+03
B-8

Peach Bottom STSBO with RCIC Blackstart

Plume 

Segment PDELAY PLHEAT PLHITE PLMDEN PLMFLA PLUDUR

1 6.09E+04 8.80E+06 3.96E+01 9.42E-01 1.40E+02 3.60E+03

2 6.09E+04 9.39E+05 4.00E+00 6.17E-01 6.61E+00 3.60E+03

3 6.45E+04 3.69E+06 3.96E+01 1.01E+00 1.29E+02 3.60E+03

4 6.81E+04 3.63E+06 3.96E+01 1.02E+00 1.27E+02 3.60E+03

5 7.17E+04 3.18E+06 3.96E+01 1.03E+00 1.22E+02 3.63E+03

6 7.54E+04 2.84E+06 3.96E+01 1.04E+00 1.17E+02 3.60E+03

7 7.90E+04 2.45E+06 3.96E+01 1.05E+00 1.11E+02 3.60E+03

8 8.26E+04 2.11E+06 3.96E+01 1.06E+00 1.06E+02 3.60E+03

9 8.62E+04 1.91E+06 3.96E+01 1.07E+00 1.03E+02 3.60E+03

10 8.98E+04 1.73E+06 3.96E+01 1.07E+00 9.92E+01 3.60E+03

11 9.34E+04 1.58E+06 3.96E+01 1.08E+00 9.64E+01 3.60E+03

12 9.70E+04 1.48E+06 3.96E+01 1.08E+00 9.43E+01 3.60E+03

13 1.01E+05 1.41E+06 3.96E+01 1.08E+00 9.29E+01 3.60E+03

14 1.04E+05 1.36E+06 3.96E+01 1.08E+00 9.18E+01 3.60E+03

15 1.08E+05 1.33E+06 3.96E+01 1.09E+00 9.09E+01 3.60E+03

16 1.11E+05 1.30E+06 3.96E+01 1.09E+00 9.02E+01 3.60E+03

17 1.15E+05 1.34E+06 3.96E+01 1.09E+00 9.09E+01 3.60E+03

18 1.19E+05 1.28E+06 3.96E+01 1.09E+00 9.11E+01 3.60E+03

19 1.22E+05 1.27E+06 3.96E+01 1.09E+00 8.92E+01 3.60E+03

20 1.26E+05 1.27E+06 3.96E+01 1.09E+00 8.90E+01 3.60E+03

21 1.29E+05 1.26E+06 3.96E+01 1.09E+00 8.87E+01 3.60E+03

22 1.33E+05 1.26E+06 3.96E+01 1.09E+00 8.85E+01 3.60E+03

23 1.37E+05 1.26E+06 3.96E+01 1.09E+00 8.84E+01 3.60E+03

24 1.40E+05 1.25E+06 3.96E+01 1.09E+00 8.83E+01 3.60E+03

25 1.44E+05 1.25E+06 3.96E+01 1.09E+00 8.82E+01 3.60E+03

26 1.47E+05 1.25E+06 3.96E+01 1.09E+00 8.81E+01 3.60E+03

27 1.51E+05 1.25E+06 3.96E+01 1.09E+00 8.80E+01 3.60E+03

28 1.55E+05 1.25E+06 3.96E+01 1.09E+00 8.80E+01 3.60E+03

29 1.58E+05 1.25E+06 3.96E+01 1.09E+00 8.78E+01 3.60E+03

30 1.62E+05 1.24E+06 3.96E+01 1.09E+00 8.78E+01 3.60E+03

31 1.65E+05 1.24E+06 3.96E+01 1.09E+00 8.77E+01 3.60E+03

32 1.69E+05 1.24E+06 3.96E+01 1.09E+00 8.76E+01 3.60E+03

33 1.73E+05 1.24E+06 3.96E+01 1.09E+00 6.56E+01 2.40E+02
B-9

Peach Bottom STSBO without RCIC Blackstart

Plume 

Segment PDELAY PLHEAT PLHITE PLMDEN PLMFLA PLUDUR

1 2.93E+04 1.94E+00 1.75E+01 1.11E+00 1.94E+00 3.60E+03

2 3.05E+04 4.02E+01 3.96E+01 6.64E-01 4.02E+01 3.60E+03

3 3.05E+04 2.17E+01 4.00E+00 6.43E-01 2.17E+01 3.60E+03

4 3.12E+04 1.44E+02 3.96E+01 9.81E-01 1.44E+02 3.66E+03

5 3.41E+04 8.34E+00 4.00E+00 5.24E-01 8.34E+00 3.62E+03

6 3.48E+04 1.18E+02 3.96E+01 1.05E+00 1.18E+02 3.56E+03

7 3.77E+04 3.85E+00 4.00E+00 5.49E-01 3.85E+00 3.60E+03

8 3.84E+04 1.04E+02 3.96E+01 1.07E+00 1.04E+02 3.60E+03

9 4.01E+04 1.75E+00 1.75E+01 1.11E+00 1.75E+00 3.60E+03

10 4.13E+04 2.65E+00 4.00E+00 4.72E-01 2.65E+00 3.60E+03

11 4.20E+04 9.13E+01 3.96E+01 1.09E+00 9.13E+01 3.60E+03

12 4.37E+04 3.34E+00 1.75E+01 1.11E+00 3.34E+00 3.60E+03

13 4.56E+04 8.39E+01 3.96E+01 1.11E+00 8.39E+01 3.60E+03

14 4.73E+04 3.59E+00 1.75E+01 1.11E+00 3.59E+00 3.60E+03

15 4.92E+04 8.17E+01 3.96E+01 1.11E+00 8.17E+01 3.60E+03

16 5.09E+04 3.58E+00 1.75E+01 1.11E+00 3.58E+00 3.60E+03

17 5.28E+04 8.04E+01 3.96E+01 1.11E+00 8.04E+01 3.60E+03

18 5.45E+04 3.51E+00 1.75E+01 1.11E+00 3.51E+00 3.60E+03

19 5.64E+04 7.92E+01 3.96E+01 1.12E+00 7.92E+01 3.60E+03

20 5.81E+04 3.41E+00 1.75E+01 1.11E+00 3.41E+00 3.60E+03

21 6.00E+04 7.82E+01 3.96E+01 1.12E+00 7.82E+01 3.60E+03

22 6.17E+04 3.31E+00 1.75E+01 1.12E+00 3.31E+00 3.60E+03

23 6.36E+04 7.75E+01 3.96E+01 1.12E+00 7.75E+01 3.60E+03

24 6.53E+04 3.21E+00 1.75E+01 1.12E+00 3.21E+00 3.60E+03

25 6.72E+04 7.73E+01 3.96E+01 1.12E+00 7.73E+01 3.60E+03

26 6.89E+04 3.14E+00 1.75E+01 1.12E+00 3.14E+00 3.63E+03

27 7.08E+04 7.77E+01 3.96E+01 1.12E+00 7.77E+01 3.60E+03

28 7.25E+04 3.16E+00 1.75E+01 1.12E+00 3.16E+00 3.60E+03

29 7.44E+04 7.77E+01 3.96E+01 1.12E+00 7.77E+01 3.60E+03

30 7.61E+04 3.22E+00 1.75E+01 1.12E+00 3.22E+00 3.60E+03

31 7.80E+04 7.90E+01 3.96E+01 1.12E+00 7.90E+01 3.60E+03

32 7.97E+04 3.35E+00 1.75E+01 1.11E+00 3.35E+00 3.60E+03

33 8.16E+04 8.60E+01 3.96E+01 1.11E+00 8.60E+01 3.60E+03

34 8.33E+04 3.81E+00 1.75E+01 1.11E+00 3.81E+00 3.60E+03

35 8.52E+04 8.83E+01 3.96E+01 1.11E+00 8.83E+01 3.72E+03

36 8.69E+04 4.19E+00 1.75E+01 1.10E+00 4.19E+00 3.60E+03

37 8.82E+04 5.96E-01 3.96E+01 1.11E+00 5.96E-01 3.60E+03

38 8.89E+04 8.60E+01 3.96E+01 1.11E+00 8.60E+01 3.60E+03

39 9.05E+04 4.47E+00 1.75E+01 1.09E+00 4.47E+00 3.60E+03

40 9.18E+04 5.99E-01 3.96E+01 1.11E+00 5.99E-01 3.60E+03

41 9.25E+04 8.72E+01 3.96E+01 1.11E+00 8.72E+01 3.60E+03

42 9.41E+04 4.68E+00 1.75E+01 1.09E+00 4.68E+00 3.60E+03

43 9.54E+04 6.12E-01 3.96E+01 1.11E+00 6.12E-01 3.60E+03
B-10

Peach Bottom STSBO without RCIC Blackstart

Plume 

Segment PDELAY PLHEAT PLHITE PLMDEN PLMFLA PLUDUR

44 9.61E+04 8.79E+01 3.96E+01 1.11E+00 8.79E+01 3.60E+03

45 9.77E+04 4.89E+00 1.75E+01 1.08E+00 4.89E+00 3.60E+03

46 9.97E+04 8.85E+01 3.96E+01 1.11E+00 8.85E+01 3.60E+03

47 1.01E+05 5.05E+00 1.75E+01 1.08E+00 5.05E+00 3.60E+03

48 1.03E+05 8.89E+01 3.96E+01 1.10E+00 8.89E+01 3.60E+03

49 1.05E+05 5.18E+00 1.75E+01 1.08E+00 5.18E+00 3.60E+03

50 1.07E+05 8.93E+01 3.96E+01 1.10E+00 8.93E+01 3.60E+03

51 1.09E+05 5.28E+00 1.75E+01 1.07E+00 5.28E+00 3.60E+03

52 1.11E+05 8.95E+01 3.96E+01 1.10E+00 8.95E+01 3.60E+03

53 1.12E+05 5.34E+00 1.75E+01 1.07E+00 5.34E+00 3.60E+03

54 1.14E+05 8.97E+01 3.96E+01 1.10E+00 8.97E+01 3.60E+03

55 1.16E+05 5.39E+00 1.75E+01 1.07E+00 5.39E+00 3.60E+03

56 1.18E+05 8.99E+01 3.96E+01 1.10E+00 8.99E+01 3.60E+03

57 1.19E+05 5.44E+00 1.75E+01 1.07E+00 5.44E+00 3.60E+03

58 1.21E+05 9.00E+01 3.96E+01 1.10E+00 9.00E+01 3.60E+03

59 1.23E+05 5.47E+00 1.75E+01 1.07E+00 5.47E+00 3.60E+03

60 1.25E+05 9.01E+01 3.96E+01 1.10E+00 9.01E+01 3.60E+03

61 1.27E+05 5.50E+00 1.75E+01 1.07E+00 5.50E+00 3.60E+03

62 1.29E+05 9.02E+01 3.96E+01 1.10E+00 9.02E+01 3.60E+03

63 1.30E+05 5.53E+00 1.75E+01 1.07E+00 5.53E+00 3.60E+03

64 1.32E+05 9.01E+01 3.96E+01 1.10E+00 9.01E+01 3.60E+03

65 1.34E+05 5.56E+00 1.75E+01 1.06E+00 5.56E+00 3.60E+03

66 1.36E+05 9.02E+01 3.96E+01 1.10E+00 9.02E+01 3.60E+03

67 1.37E+05 5.57E+00 1.75E+01 1.06E+00 5.57E+00 3.60E+03

68 1.39E+05 9.03E+01 3.96E+01 1.10E+00 9.03E+01 3.60E+03

69 1.41E+05 5.59E+00 1.75E+01 1.06E+00 5.59E+00 3.60E+03

70 1.43E+05 9.03E+01 3.96E+01 1.10E+00 9.03E+01 3.60E+03

71 1.45E+05 5.60E+00 1.75E+01 1.06E+00 5.60E+00 3.60E+03

72 1.47E+05 9.03E+01 3.96E+01 1.10E+00 9.03E+01 3.60E+03

73 1.48E+05 5.62E+00 1.75E+01 1.06E+00 5.62E+00 3.60E+03

74 1.50E+05 9.03E+01 3.96E+01 1.10E+00 9.03E+01 3.60E+03

75 1.52E+05 5.63E+00 1.75E+01 1.06E+00 5.63E+00 3.60E+03

76 1.54E+05 9.03E+01 3.96E+01 1.10E+00 9.03E+01 3.60E+03

77 1.55E+05 5.64E+00 1.75E+01 1.06E+00 5.64E+00 3.60E+03

78 1.57E+05 9.02E+01 3.96E+01 1.10E+00 9.02E+01 3.60E+03

79 1.59E+05 5.65E+00 1.75E+01 1.06E+00 5.65E+00 3.60E+03

80 1.61E+05 9.02E+01 3.96E+01 1.10E+00 9.02E+01 3.60E+03

81 1.63E+05 5.66E+00 1.75E+01 1.06E+00 5.66E+00 3.60E+03

82 1.65E+05 9.01E+01 3.96E+01 1.10E+00 9.01E+01 3.60E+03

83 1.66E+05 5.68E+00 1.75E+01 1.06E+00 5.68E+00 3.60E+03

84 1.68E+05 9.00E+01 3.96E+01 1.10E+00 9.00E+01 3.60E+03

85 1.70E+05 5.68E+00 1.75E+01 1.06E+00 5.68E+00 3.06E+03

86 1.72E+05 8.98E+01 3.96E+01 1.10E+00 8.98E+01 1.08E+03
B-11

Table B-3 Plume Parameters Used in the Peach Bottom Unmitigated LTSBO and 

STSBO with and without RCIC Blackstart scenarios

Peach Bottom LTSBO

Class Bin 1 Bin 2 Bin 3 Bin 4 Bin 5 Bin 6 Bin 7 Bin 8 Bin 9 Bin 10

Xe 0.1000 0.1000 0.1000 0.1000 0.1000 0.1000 0.1000 0.1000 0.1000 0.1000

Cs 0.1171 0.1602 0.3303 0.3061 0.0575 0.0180 0.0071 0.0025 0.0009 0.0004

Ba 0.0412 0.2230 0.4450 0.2275 0.0486 0.0111 0.0030 0.0005 0.0001 0.0001

I 0.0231 0.1503 0.5219 0.2579 0.0333 0.0102 0.0027 0.0005 0.0001 0.0000

Te 0.0286 0.1370 0.4603 0.2779 0.0643 0.0249 0.0059 0.0009 0.0002 0.0001

Ru 0.0393 0.1512 0.3597 0.2682 0.1214 0.0484 0.0104 0.0011 0.0002 0.0001

Mo 0.1553 0.1638 0.2563 0.3259 0.0649 0.0202 0.0087 0.0033 0.0012 0.0005

Ce 0.0306 0.1930 0.4357 0.2427 0.0708 0.0218 0.0047 0.0005 0.0001 0.0001

La 0.0235 0.1473 0.3990 0.2664 0.1109 0.0433 0.0089 0.0007 0.0000 0.0000

Peach Bottom STSBO with RCIC Blackstart

Class Bin 1 Bin 2 Bin 3 Bin 4 Bin 5 Bin 6 Bin 7 Bin 8 Bin 9 Bin 10

Xe 0.1000 0.1000 0.1000 0.1000 0.1000 0.1000 0.1000 0.1000 0.1000 0.1000

Cs 0.1704 0.1914 0.2388 0.2874 0.0796 0.0169 0.0094 0.0039 0.0017 0.0005

Ba 0.0303 0.1751 0.4211 0.3035 0.0581 0.0084 0.0023 0.0008 0.0003 0.0001

I 0.0322 0.1958 0.4343 0.2527 0.0634 0.0164 0.0040 0.0008 0.0003 0.0001

Te 0.0499 0.1770 0.3605 0.2675 0.0987 0.0353 0.0088 0.0016 0.0005 0.0001

Ru 0.0501 0.1443 0.2997 0.3035 0.1406 0.0480 0.0114 0.0017 0.0005 0.0001

Mo 0.2081 0.1901 0.1871 0.2966 0.0834 0.0167 0.0108 0.0047 0.0021 0.0006

Ce 0.0188 0.1393 0.4164 0.3264 0.0786 0.0164 0.0034 0.0005 0.0001 0.0001

La 0.0218 0.1333 0.3404 0.3148 0.1352 0.0442 0.0093 0.0008 0.0001 0.0000

Peach Bottom STSBO without RCIC Blackstart

Class Bin 1 Bin 2 Bin 3 Bin 4 Bin 5 Bin 6 Bin 7 Bin 8 Bin 9 Bin 10

Xe 0.1000 0.1000 0.1000 0.1000 0.1000 0.1000 0.1000 0.1000 0.1000 0.1000

Cs 0.0288 0.2922 0.2076 0.2174 0.1847 0.0587 0.0089 0.0012 0.0002 0.0003

Ba 0.0178 0.0546 0.1120 0.3364 0.3641 0.0994 0.0135 0.0018 0.0002 0.0001

I 0.0282 0.5160 0.2181 0.1261 0.0864 0.0217 0.0023 0.0003 0.0003 0.0006

Te 0.0393 0.3360 0.2596 0.2010 0.1225 0.0376 0.0034 0.0004 0.0001 0.0001

Ru 0.0323 0.0946 0.1503 0.2915 0.2935 0.1127 0.0209 0.0036 0.0005 0.0003

Mo 0.0300 0.1217 0.1995 0.2859 0.2584 0.0880 0.0141 0.0020 0.0002 0.0002

Ce 0.0219 0.0558 0.1024 0.3281 0.3666 0.1076 0.0151 0.0021 0.0002 0.0002

La 0.0203 0.0724 0.1218 0.3231 0.3501 0.0976 0.0127 0.0016 0.0002 0.0002

.
B-12

Table B-4 Release Fraction Parameters Used in the Peach Bottom Unmitigated 

LTSBO and STSBO with and without RCIC Blackstart Scenarios

Peach Bottom LTSBO

Plume 

Segment Xe Cs Ba I Te Ru Mo Ce La

1 8.75E-01 2.32E-03 1.17E-03 1.65E-03 1.73E-03 1.31E-06 6.39E-04 1.32E-05 1.91E-06

2 7.93E-02 3.29E-04 2.01E-05 4.99E-05 1.27E-04 9.46E-08 9.48E-05 1.85E-07 2.79E-08

3 1.71E-02 2.47E-04 3.24E-03 2.27E-03 4.00E-04 7.92E-07 2.52E-05 4.90E-05 3.49E-06

4 3.23E-03 1.39E-04 5.75E-04 1.10E-03 3.01E-04 7.10E-07 1.65E-05 1.01E-05 2.83E-06

5 7.40E-04 4.24E-04 1.48E-04 6.08E-03 7.48E-03 4.11E-07 7.67E-06 3.23E-06 2.40E-06

6 9.39E-04 2.61E-04 1.15E-04 3.64E-03 1.76E-03 4.13E-07 6.80E-06 2.46E-06 2.05E-06

7 6.99E-04 1.06E-04 7.24E-05 1.22E-03 2.99E-03 4.31E-07 6.98E-06 2.13E-06 1.95E-06

8 2.97E-04 6.06E-05 6.33E-05 5.46E-04 7.42E-04 4.12E-07 6.36E-06 1.74E-06 1.64E-06

9 1.39E-04 8.02E-05 2.80E-05 8.82E-04 3.63E-04 4.13E-07 5.89E-06 1.23E-06 1.18E-06

10 8.46E-05 7.28E-05 2.15E-05 8.05E-04 4.03E-04 4.21E-07 5.29E-06 1.71E-06 1.69E-06

11 5.97E-05 5.33E-05 1.02E-05 5.08E-04 2.82E-04 4.19E-07 5.07E-06 1.65E-06 1.63E-06

12 4.96E-05 3.33E-05 4.59E-06 1.56E-04 2.06E-04 4.11E-07 5.63E-06 1.57E-06 1.55E-06

13 4.42E-05 3.02E-05 5.34E-06 9.41E-05 2.47E-04 4.01E-07 6.04E-06 1.45E-06 1.44E-06

14 3.96E-05 2.74E-05 1.36E-05 7.53E-05 3.13E-04 3.94E-07 5.77E-06 1.35E-06 1.34E-06

15 3.59E-05 2.63E-05 2.23E-05 6.53E-05 4.07E-04 3.84E-07 5.62E-06 1.26E-06 1.25E-06

16 3.34E-05 2.66E-05 5.12E-06 6.08E-05 3.28E-04 3.78E-07 5.74E-06 1.19E-06 1.18E-06

17 3.06E-05 2.76E-05 4.04E-06 6.02E-05 3.34E-04 3.70E-07 5.98E-06 1.12E-06 1.11E-06

18 2.93E-05 2.85E-05 4.33E-06 5.94E-05 3.93E-04 3.62E-07 6.36E-06 1.05E-06 1.04E-06

19 2.71E-05 3.00E-05 4.34E-06 6.02E-05 3.89E-04 3.53E-07 6.84E-06 9.87E-07 9.83E-07

20 2.55E-05 3.23E-05 8.15E-06 6.31E-05 2.78E-04 3.44E-07 7.41E-06 9.35E-07 9.31E-07

21 2.27E-05 3.47E-05 1.38E-05 6.71E-05 2.68E-04 3.35E-07 8.10E-06 8.90E-07 8.86E-07

22 2.11E-05 3.65E-05 5.32E-06 7.33E-05 2.96E-04 3.07E-07 8.54E-06 8.51E-07 8.47E-07

23 2.05E-05 4.15E-05 5.60E-06 8.51E-05 3.42E-04 3.09E-07 9.73E-06 8.15E-07 8.12E-07

24 1.95E-05 4.67E-05 5.86E-06 1.02E-04 3.42E-04 3.06E-07 1.09E-05 7.82E-07 7.78E-07

25 1.86E-05 5.16E-05 6.03E-06 1.27E-04 3.43E-04 2.99E-07 1.18E-05 7.53E-07 7.50E-07

26 1.82E-05 5.72E-05 6.11E-06 1.61E-04 3.53E-04 2.92E-07 1.28E-05 7.26E-07 7.23E-07

27 1.74E-05 6.10E-05 6.32E-06 2.05E-04 3.79E-04 2.86E-07 1.31E-05 6.99E-07 6.96E-07

28 1.67E-05 3.25E-05 6.38E-06 1.11E-04 3.51E-04 2.79E-07 6.72E-06 6.75E-07 6.72E-07

29 1.54E-05 2.55E-05 6.04E-06 7.60E-05 1.98E-04 2.59E-07 5.37E-06 6.19E-07 6.16E-07
B-13

Peach Bottom STSBO with RCIC Blackstart

Plume 

Segment Xe Cs Ba I Te Ru Mo Ce La

1 8.60E-01 2.42E-03 1.66E-03 1.90E-03 2.29E-03 1.20E-06 6.56E-04 1.68E-05 2.13E-06

2 8.85E-02 4.22E-04 3.99E-05 7.02E-05 2.09E-04 1.32E-07 1.20E-04 2.40E-07 3.26E-08

3 2.02E-02 2.15E-04 4.28E-03 2.29E-03 8.33E-04 4.01E-07 1.76E-05 7.56E-05 3.61E-06

4 5.32E-03 1.02E-04 5.37E-04 1.07E-03 3.48E-04 3.81E-07 7.97E-06 1.00E-05 2.41E-06

5 1.51E-03 6.39E-05 8.13E-05 6.60E-04 2.14E-04 3.61E-07 5.06E-06 3.33E-06 2.45E-06

6 8.70E-04 1.26E-04 4.44E-05 1.75E-03 1.75E-03 2.84E-07 3.25E-06 2.42E-06 2.15E-06

7 8.32E-04 7.58E-05 5.18E-05 1.05E-03 7.84E-04 2.15E-07 2.03E-06 2.14E-06 1.90E-06

8 4.41E-04 1.04E-04 4.80E-05 1.50E-03 6.51E-04 2.40E-07 1.70E-06 1.28E-06 1.16E-06

9 2.13E-04 4.35E-05 5.05E-05 5.72E-04 1.71E-03 2.57E-07 1.61E-06 1.80E-06 1.75E-06

10 1.20E-04 2.69E-05 5.92E-05 3.06E-04 7.49E-04 2.65E-07 1.71E-06 1.71E-06 1.68E-06

11 7.74E-05 1.51E-05 2.86E-05 9.63E-05 2.02E-04 2.67E-07 2.06E-06 1.61E-06 1.60E-06

12 4.05E-05 1.61E-05 3.02E-05 7.74E-05 1.96E-04 2.66E-07 2.56E-06 1.51E-06 1.50E-06

13 4.17E-05 1.40E-05 2.40E-05 6.22E-05 2.37E-04 2.62E-07 2.53E-06 1.40E-06 1.39E-06

14 5.76E-05 1.33E-05 3.86E-05 5.33E-05 3.01E-04 2.59E-07 2.49E-06 1.31E-06 1.30E-06

15 5.39E-05 1.29E-05 5.58E-05 4.97E-05 3.54E-04 2.56E-07 2.42E-06 1.23E-06 1.22E-06

16 4.91E-05 1.31E-05 1.47E-05 4.81E-05 2.81E-04 2.49E-07 2.49E-06 1.16E-06 1.15E-06

17 4.75E-05 4.58E-05 5.74E-05 3.28E-04 9.72E-04 3.29E-07 6.42E-06 1.09E-06 1.08E-06

18 4.05E-05 2.05E-05 1.58E-05 1.52E-04 5.36E-04 2.12E-07 2.83E-06 1.02E-06 1.01E-06

19 3.79E-05 1.84E-05 6.90E-06 1.27E-04 2.59E-04 2.10E-07 2.71E-06 9.75E-07 9.68E-07

20 3.53E-05 1.56E-05 7.82E-06 6.08E-05 2.72E-04 2.05E-07 3.13E-06 9.28E-07 9.22E-07

21 3.25E-05 1.68E-05 8.97E-06 5.73E-05 1.65E-04 1.88E-07 3.55E-06 8.86E-07 8.80E-07

22 2.99E-05 1.87E-05 7.47E-06 6.39E-05 1.64E-04 1.71E-07 3.99E-06 8.47E-07 8.41E-07

23 2.83E-05 2.15E-05 4.55E-06 7.57E-05 1.81E-04 1.69E-07 4.60E-06 8.11E-07 8.06E-07

24 2.66E-05 2.48E-05 4.35E-06 9.32E-05 2.02E-04 1.65E-07 5.26E-06 7.80E-07 7.75E-07

25 2.49E-05 2.88E-05 4.28E-06 1.17E-04 2.28E-04 1.61E-07 5.96E-06 7.50E-07 7.46E-07

26 2.35E-05 3.33E-05 4.23E-06 1.50E-04 2.59E-04 1.57E-07 6.64E-06 7.22E-07 7.18E-07

27 2.22E-05 3.17E-05 4.17E-06 8.83E-05 2.95E-04 1.54E-07 7.34E-06 6.96E-07 6.92E-07

28 2.09E-05 3.06E-05 4.11E-06 6.29E-05 1.87E-04 1.50E-07 7.47E-06 6.70E-07 6.66E-07

29 1.96E-05 2.42E-05 4.06E-06 7.08E-05 1.19E-04 1.45E-07 5.43E-06 6.45E-07 6.42E-07

30 1.85E-05 2.15E-05 4.01E-06 8.06E-05 1.19E-04 1.42E-07 4.46E-06 6.24E-07 6.21E-07

31 1.76E-05 1.40E-05 3.96E-06 9.33E-05 1.21E-04 1.39E-07 2.02E-06 6.03E-07 6.00E-07

32 1.66E-05 1.50E-05 3.92E-06 1.09E-04 1.22E-04 1.36E-07 2.04E-06 5.83E-07 5.80E-07

33 1.01E-06 1.06E-06 2.59E-07 7.85E-06 8.25E-06 9.00E-09 1.39E-07 3.82E-08 3.80E-08
B-14

Peach Bottom STSBO without RCIC Blackstart

Plume 

Segment Xe Cs Ba I Te Ru Mo Ce La

1 1.13E-03 5.67E-06 9.10E-05 1.11E-05 9.18E-06 4.79E-09 1.41E-06 8.74E-06 1.16E-07

2 5.21E-01 2.40E-03 2.98E-02 3.75E-03 3.58E-03 2.58E-06 6.15E-04 2.95E-03 3.72E-05

3 7.48E-02 3.61E-04 1.09E-02 1.29E-03 9.89E-04 2.26E-07 7.37E-05 8.41E-04 1.41E-05

4 1.98E-01 8.81E-04 3.80E-02 4.19E-03 3.15E-03 3.02E-07 1.49E-04 2.66E-03 4.90E-05

5 3.70E-03 8.13E-05 1.69E-04 2.12E-04 2.11E-04 8.78E-09 1.58E-05 7.26E-06 4.70E-07

6 2.01E-02 3.61E-04 1.18E-03 8.66E-04 8.98E-04 5.36E-08 7.27E-05 6.02E-05 2.34E-06

7 1.16E-03 5.96E-05 5.39E-04 1.03E-04 7.69E-05 3.28E-09 1.33E-05 2.25E-06 1.84E-07

8 1.08E-02 4.65E-04 3.59E-03 6.22E-04 3.91E-04 3.09E-08 1.09E-04 3.11E-05 1.48E-06

9 1.12E-03 7.34E-06 1.01E-04 1.63E-05 1.36E-05 3.48E-09 1.66E-06 7.42E-06 1.20E-07

10 5.67E-04 3.23E-05 6.83E-05 1.24E-05 3.72E-05 1.82E-09 8.03E-06 1.37E-06 1.05E-07

11 6.67E-03 2.46E-04 5.78E-04 1.27E-04 4.14E-04 1.88E-08 5.93E-05 1.84E-05 9.53E-07

12 1.95E-03 1.55E-05 1.70E-04 2.80E-05 2.78E-05 5.78E-09 3.57E-06 1.20E-05 2.05E-07

13 5.32E-03 2.57E-04 8.56E-04 2.24E-04 1.99E-04 1.34E-08 5.84E-05 1.05E-05 8.08E-07

14 1.95E-03 1.91E-05 1.71E-04 3.02E-05 2.88E-05 5.45E-09 4.37E-06 1.10E-05 2.00E-07

15 5.01E-03 2.81E-04 1.58E-03 4.31E-04 1.25E-04 1.07E-08 6.18E-05 8.28E-06 8.70E-07

16 1.81E-03 2.17E-05 1.76E-04 3.38E-05 2.65E-05 4.75E-09 4.92E-06 9.33E-06 1.85E-07

17 4.78E-03 2.98E-04 1.18E-03 4.10E-04 1.11E-04 9.07E-09 6.94E-05 7.01E-06 1.12E-06

18 1.66E-03 2.39E-05 1.65E-04 3.66E-05 2.39E-05 4.07E-09 5.45E-06 7.79E-06 1.75E-07

19 4.60E-03 3.20E-04 4.99E-04 5.76E-04 1.05E-04 7.47E-09 7.33E-05 5.77E-06 1.05E-06

20 1.51E-03 2.55E-05 1.45E-04 4.10E-05 2.16E-05 3.46E-09 5.82E-06 6.46E-06 1.63E-07

21 4.40E-03 2.53E-04 7.84E-04 5.51E-04 1.00E-04 6.58E-09 5.32E-05 4.84E-06 9.73E-07

22 1.38E-03 2.56E-05 1.34E-04 4.37E-05 1.95E-05 2.95E-09 5.76E-06 5.36E-06 1.51E-07

23 4.23E-03 2.36E-04 1.13E-03 3.10E-04 9.64E-05 5.95E-09 5.24E-05 4.13E-06 9.16E-07

24 1.26E-03 2.51E-05 1.30E-04 4.04E-05 1.77E-05 2.51E-09 5.66E-06 4.45E-06 1.40E-07

25 4.11E-03 2.21E-04 6.95E-04 4.65E-05 1.03E-04 5.46E-09 5.58E-05 3.58E-06 8.67E-07

26 1.17E-03 2.47E-05 1.17E-04 3.42E-05 1.64E-05 2.17E-09 5.66E-06 3.74E-06 1.31E-07

27 3.96E-03 2.34E-04 8.36E-05 4.40E-05 1.19E-04 5.02E-09 5.98E-05 3.10E-06 8.24E-07

28 1.11E-03 2.13E-05 8.54E-05 2.53E-05 1.37E-05 1.63E-09 4.98E-06 2.73E-06 1.06E-07

29 3.89E-03 2.44E-04 5.21E-05 5.61E-05 1.44E-04 4.51E-09 6.33E-05 2.47E-06 7.83E-07

30 1.07E-03 1.91E-05 6.14E-05 1.89E-05 1.21E-05 1.23E-09 4.56E-06 1.97E-06 8.78E-08

31 4.12E-03 2.36E-04 3.97E-05 8.28E-05 1.83E-04 4.41E-09 6.07E-05 2.03E-06 7.60E-07

32 1.07E-03 1.91E-05 4.95E-05 1.80E-05 1.30E-05 1.05E-09 4.62E-06 1.60E-06 8.31E-08

33 9.23E-03 1.43E-04 3.99E-05 5.78E-04 2.49E-04 3.51E-09 2.73E-05 2.15E-06 8.15E-07

34 1.24E-03 2.04E-05 4.70E-05 2.90E-05 1.71E-05 1.04E-09 4.77E-06 1.54E-06 9.33E-08

35 7.72E-03 3.45E-04 7.37E-05 4.52E-03 4.74E-04 3.31E-09 1.42E-05 3.19E-06 7.50E-07

36 1.38E-03 3.17E-05 4.54E-05 2.02E-04 3.07E-05 1.03E-09 4.80E-06 1.48E-06 1.04E-07

37 3.22E-05 9.69E-06 6.78E-07 1.42E-04 9.97E-06 2.30E-11 1.35E-07 2.05E-08 4.80E-09

38 4.20E-03 1.54E-03 8.71E-05 2.28E-02 1.69E-03 3.07E-09 1.89E-05 2.77E-06 6.85E-07

39 1.38E-03 6.87E-05 4.21E-05 7.82E-04 7.56E-05 9.79E-10 4.74E-06 1.36E-06 1.10E-07

40 2.06E-05 1.04E-05 3.48E-07 1.56E-04 1.99E-05 1.55E-11 1.08E-07 1.30E-08 4.52E-09

41 2.80E-03 1.48E-03 4.78E-05 2.21E-02 3.45E-03 2.18E-09 1.55E-05 1.70E-06 6.50E-07

42 1.32E-03 1.02E-04 3.79E-05 1.31E-03 1.79E-04 9.09E-10 4.56E-06 1.22E-06 1.14E-07

43 1.50E-05 8.34E-06 3.61E-07 1.24E-04 4.99E-05 1.48E-11 1.14E-07 9.31E-09 4.34E-09

44 2.02E-03 1.09E-03 5.50E-05 1.61E-02 8.02E-03 2.14E-09 1.68E-05 1.30E-06 6.17E-07

45 1.24E-03 1.20E-04 3.44E-05 1.60E-03 4.07E-04 8.48E-10 4.46E-06 1.09E-06 1.18E-07

46 1.47E-03 6.02E-04 8.16E-05 8.38E-03 9.10E-03 2.10E-09 1.91E-05 1.15E-06 5.85E-07

47 1.14E-03 1.21E-04 3.21E-05 1.63E-03 6.55E-04 7.90E-10 4.44E-06 9.64E-07 1.19E-07
B-15

Peach Bottom STSBO without RCIC Blackstart

Plume 

Segment Xe Cs Ba I Te Ru Mo Ce La

48 1.09E-03 3.24E-04 1.06E-04 4.00E-03 1.11E-02 1.94E-09 1.95E-05 1.03E-06 5.55E-07

49 1.02E-03 1.13E-04 3.06E-05 1.51E-03 9.56E-04 7.32E-10 4.43E-06 8.44E-07 1.20E-07

50 8.15E-04 1.99E-04 9.58E-05 1.94E-03 1.45E-02 1.92E-09 2.07E-05 9.29E-07 5.30E-07

51 9.09E-04 1.01E-04 2.89E-05 1.32E-03 1.29E-03 6.76E-10 4.43E-06 7.34E-07 1.19E-07

52 6.24E-04 1.37E-04 9.13E-05 1.02E-03 7.55E-03 1.77E-09 1.95E-05 8.48E-07 5.07E-07

53 7.96E-04 8.88E-05 2.73E-05 1.13E-03 1.33E-03 6.23E-10 4.39E-06 6.34E-07 1.17E-07

54 4.85E-04 1.21E-04 1.08E-04 7.21E-04 3.31E-03 1.77E-09 2.02E-05 7.73E-07 4.86E-07

55 6.92E-04 7.77E-05 2.66E-05 9.60E-04 1.22E-03 5.78E-10 4.37E-06 5.48E-07 1.14E-07

56 3.85E-04 1.17E-04 1.29E-04 6.32E-04 1.97E-03 1.74E-09 2.06E-05 7.12E-07 4.67E-07

57 6.00E-04 6.84E-05 2.66E-05 8.14E-04 1.08E-03 5.39E-10 4.37E-06 4.74E-07 1.12E-07

58 3.12E-04 1.16E-04 9.57E-05 6.42E-04 1.57E-03 1.66E-09 1.99E-05 6.61E-07 4.50E-07

59 5.16E-04 6.05E-05 2.53E-05 6.92E-04 9.41E-04 5.04E-10 4.35E-06 4.11E-07 1.09E-07

60 2.58E-04 1.20E-04 4.82E-05 6.56E-04 1.55E-03 1.64E-09 2.08E-05 5.93E-07 4.34E-07

61 4.43E-04 5.39E-05 2.24E-05 5.88E-04 8.29E-04 4.73E-10 4.36E-06 3.56E-07 1.06E-07

62 2.17E-04 8.83E-05 2.79E-05 2.00E-04 1.60E-03 1.59E-09 2.01E-05 5.49E-07 4.19E-07

63 3.80E-04 4.74E-05 1.93E-05 4.89E-04 7.40E-04 4.47E-10 4.34E-06 3.09E-07 1.03E-07

64 1.86E-04 8.44E-05 2.10E-05 1.31E-04 1.72E-03 1.58E-09 2.01E-05 5.17E-07 4.04E-07

65 3.25E-04 4.20E-05 1.65E-05 4.04E-04 6.72E-04 4.25E-10 4.34E-06 2.70E-07 9.95E-08

66 1.62E-04 8.79E-05 1.92E-05 1.06E-04 1.92E-03 1.60E-09 2.15E-05 4.88E-07 3.91E-07

67 2.77E-04 3.76E-05 1.42E-05 3.33E-04 6.24E-04 4.08E-10 4.38E-06 2.37E-07 9.64E-08

68 1.41E-04 8.52E-05 1.77E-05 8.70E-05 1.86E-03 1.51E-09 2.10E-05 4.63E-07 3.79E-07

69 2.36E-04 3.41E-05 1.22E-05 2.74E-04 5.78E-04 3.92E-10 4.42E-06 2.09E-07 9.35E-08

70 1.24E-04 8.90E-05 1.77E-05 7.19E-05 1.22E-03 1.54E-09 2.23E-05 4.38E-07 3.66E-07

71 2.01E-04 3.12E-05 1.06E-05 2.25E-04 5.18E-04 3.79E-10 4.48E-06 1.86E-07 9.06E-08

72 1.11E-04 9.11E-05 1.76E-05 5.95E-05 1.07E-03 1.53E-09 2.30E-05 4.18E-07 3.55E-07

73 1.71E-04 2.90E-05 9.30E-06 1.85E-04 4.61E-04 3.68E-10 4.56E-06 1.66E-07 8.78E-08

74 9.89E-05 9.35E-05 1.73E-05 4.94E-05 6.69E-04 1.55E-09 2.38E-05 4.00E-07 3.46E-07

75 1.46E-04 2.73E-05 8.22E-06 1.52E-04 4.02E-04 3.61E-10 4.67E-06 1.50E-07 8.52E-08

76 8.89E-05 9.70E-05 1.59E-05 4.12E-05 6.33E-04 1.56E-09 2.49E-05 3.85E-07 3.38E-07

77 1.24E-04 2.61E-05 7.29E-06 1.25E-04 3.52E-04 3.55E-10 4.80E-06 1.36E-07 8.28E-08

78 7.95E-05 9.99E-05 1.46E-05 3.43E-05 6.44E-04 1.57E-09 2.58E-05 3.71E-07 3.30E-07

79 1.05E-04 2.52E-05 6.49E-06 1.03E-04 3.12E-04 3.51E-10 4.94E-06 1.24E-07 8.06E-08

80 7.25E-05 9.60E-05 1.36E-05 2.87E-05 6.22E-04 1.50E-09 2.49E-05 3.61E-07 3.26E-07

81 8.96E-05 2.44E-05 5.81E-06 8.46E-05 2.78E-04 3.47E-10 5.04E-06 1.15E-07 7.87E-08

82 6.69E-05 9.50E-05 1.31E-05 2.42E-05 6.00E-04 1.56E-09 2.49E-05 3.55E-07 3.23E-07

83 7.62E-05 2.37E-05 5.24E-06 6.95E-05 2.50E-04 3.45E-10 5.12E-06 1.06E-07 7.71E-08

84 6.12E-05 9.83E-05 1.35E-05 2.04E-05 6.25E-04 1.62E-09 2.67E-05 3.49E-07 3.22E-07

85 5.57E-05 1.98E-05 4.09E-06 4.92E-05 1.96E-04 2.94E-10 4.46E-06 8.52E-08 6.45E-08

86 1.73E-05 3.07E-05 4.19E-06 5.48E-06 1.96E-04 5.06E-10 8.72E-06 1.03E-07 9.65E-08
B-16

Table B-5 EARLY Parameters Used in the Peach Bottom Unmitigated LTSBO and 

STSBO with and without RCIC Blackstart Scenarios

Variable Description LTSBO STSBO with 

RCIC Blackstart

STSBO without 

RCIC 

Blackstart

ACNAME Latent Cancer Effect

Cancer Type 1 LEUKEMIA LEUKEMIA LEUKEMIA

Cancer Type 2 BONE BONE BONE

Cancer Type 3 BREAST BREAST BREAST

Cancer Type 4 LUNG LUNG LUNG

Cancer Type 5 THYROID THYROID THYROID

Cancer Type 6 LIVER LIVER LIVER

Cancer Type 7 COLON COLON COLON

Cancer Type 8 RESIDUAL RESIDUAL RESIDUAL

ACSUSC Population Susceptible to 

Cancer 1.0 for all cancers 1.0 for all cancers 1.0 for all cancers

ACTHRE Linear Dose-Response 

Threshold 0 0 0

BRRATE Breathing Rate (for all 

activity types) 0.000266 0.000266 0.000266

CFRISK Lifetime Cancer Fatality Risk 

Factors

Cancer Type 1 0.0111 0.0111 0.0111

Cancer Type 2 0.00019 0.00019 0.00019

Cancer Type 3 0.00506 0.00506 0.00506

Cancer Type 4 0.0198 0.0198 0.0198

Cancer Type 5 0.000648 0.000648 0.000648

Cancer Type 6 0.003 0.003 0.003

Cancer Type 7 0.0208 0.0208 0.0208

Cancer Type 8 0.0493 0.0493 0.0493

CIRISK Lifetime Cancer Injury Risk 

Factors

Cancer Type 1 0.0113 0.0113 0.0113

Cancer Type 2 0.000271 0.000271 0.000271

Cancer Type 3 0.0101 0.0101 0.0101

Cancer Type 4 0.0208 0.0208 0.0208

Cancer Type 5 0.00648 0.00648 0.00648

Cancer Type 6 0.00316 0.00316 0.00316

Cancer Type 7 0.0378 0.0378 0.0378

Cancer Type 8 0.169 0.169 0.169

CRIORG Critical Organ for EARLY 

Phase L-ICRP60ED L-ICRP60ED L-ICRP60ED

CSFACT Cloudshine Shielding Factors

Evacuation Shielding 

Factor for All but Cohort 4 1 1 1

Normal Activity 

Shielding Factor for All 

but Cohort 4

0.6 0.6 0.6

Sheltering Shielding 

Factor for All but Cohort 4 0.5 0.5 0.5
B-17

Variable Description LTSBO STSBO with 

RCIC Blackstart

STSBO without 

RCIC Blackstart

CSFACT Evacuation Shielding 

Factor for Cohort 4 1 1 1

Normal Activity 

Shielding Factor for 

Cohort 4

0.31 0.31 0.31

Sheltering Shielding 

Factor for Cohort 4 0.31 0.31 0.31

DCF_FILE Name of Dose Conversion 

Factor File

FGR13GyEquivDC

F.INP

FGR13GyEquivDCF

.INP

FGR13GyEquivDCF

.INP

DDREFA Dose-Dependent Reduction 

Factor

Cancer Type 1 2 2 2

Cancer Type 2 2 2 2

Cancer Type 3 1 1 1

Cancer Type 4 2 2 2

Cancer Type 5 2 2 2

Cancer Type 6 2 2 2

Cancer Type 7 2 2 2

Cancer Type 8 2 2 2

DDTHRE

Threshold for Applying 

Dose-Dependent Reduction 

Factor

0.2 0.2 0.2

DLTSHL Delay from Alarm Time to 

Shelter

Cohort 1 5400 3600 3600

Cohort 2 5400 3600 3600

Cohort 3 900 900 900

Cohort 4 5400 3600 3600

Cohort 5 5400 3600 3600

Cohort 6 N/A N/A N/A

DLTEVA Delay from Beginning of 

Shelter to Evacuation

Cohort 1 3600 3600 3600

Cohort 2 3600 3600 3600

Cohort 3 2700 2700 2700

Cohort 4 15300 15300 15300

Cohort 5 15300 15300 15300

Cohort 6 N/A N/A N/A

DOSEFA Cancer Dose-Response 

Linear Factors 1 for all organs 1 for all organs 1 for all organs

DOSEFB Cancer Dose-Response 

Quadratic Factors 0 for all organs 0 for all organs 0 for all organs

DOSHOT Hot-Spot Relocation Dose 

Threshold 0.05 0.05 0.05

DOSMOD Dose-Response Model Flag LNT, AT LNT, AT LNT, AT

DOSNRM Normal Relocation Dose 

Threshold 0.005 0.005 0.005

DURBEG Duration of Beginning of 

Evacuation Phase

Cohort 1 900 900 900
B-18

Variable Description LTSBO STSBO with RCIC 

Blackstart

STSBO without 

RCIC Blackstart

DURBEG Cohort 2 900 900 900

Cohort 3 3600 3600 3600

Cohort 4 1800 1800 1800

Cohort 5 1800 1800 1800

Cohort 6 N/A N/A N/A

DURMID Duration of Middle of 

Evacuation Phase 10800

Cohort 1 10800 10800 10800

Cohort 2 10800 10800 10800

Cohort 3 1800 1800 1800

Cohort 4 1800 1800 1800

Cohort 5 1800 1800 1800

Cohort 6 N/A N/A N/A

EANAM1 Text Describing the EARLY 

Assumptions

SOARCA 

calculation for 

Peach Bottom 

LTSBO, EARLY 

input

SOARCA calculation 

for Peach Bottom 

STSBO, EARLY 

input

SOARCA 

calculation for Peach 

Bottom STSBO, 

EARLY input

EANAM2 Text Describing the 

Emergency Response

Cohort 1 Group 1 Group 1 Group 1

Cohort 2 Group 2 Group 2 Group 2

Cohort 3 Group 3 Group 3 Group 3

Cohort 4 Group 4 Group 4 Group 4

Cohort 5 Group 5 Group 5 Group 5

Cohort 6 Group 6 Group 6 Group 6

EFFACA LD50 for Early Fatality 

Types

A-RED MARR 5.6 5.6 5.6

A-LUNGS 23.5 23.5 23.5

A-STOMACH 12.1 12.1 12.1

EFFACB Shape Factor for Early 

Fatality Types

A-RED MARR 6.1 6.1 6.1

A-LUNGS 9.6 9.6 9.6

A-STOMACH 9.3 9.3 9.3

EFFACY Efficacy of the KI Ingestion 0.7 0.7 0.7

EFFTHR Threshold Dose to Target 

Organ

A-RED MARR 2.32 2.32 2.32

A-LUNGS 13.6 13.6 13.6

A-STOMACH 6.5 6.5 6.5

EIFACA D50 For Early Injuries

PRODROMAL VOMIT 2 2 2

DIARRHEA 3 3 3

PNEUMONITIS 16.6 16.6 16.6

SKIN ERYTHRMA 6 6 6

TRANSEPIDERMAL 20 20 20

THYROIDITIS 240 240 240
B-19

Variable Description LTSBO STSBO with RCIC 

Blackstart

STSBO without 

RCIC Blackstart

EIFACA HYPOTHYROIDISM 60 60 60

EIFACB Shape Factor for Early 

Injuries

PRODROMAL VOMIT 3 3 3

DIARRHEA 2.5 2.5 2.5

PNEUMONITIS 7.3 7.3 7.3

SKIN ERYTHRMA 5 5 5

TRANSEPIDERMAL 5 5 5

THYROIDITIS 2 2 2

HYPOTHYROIDISM 1.3 1.3 1.3

EINAME Early Injury Effect Names 

and Corresponding Organ

PRODROMAL VOMIT A-STOMACH A-STOMACH A-STOMACH

DIARRHEA A-STOMACH A-STOMACH A-STOMACH

PNEUMONITIS A-LUNGS A-LUNGS A-LUNGS

SKIN ERYTHRMA A-SKIN A-SKIN A-SKIN

TRANSEPIDERMAL A-SKIN A-SKIN A-SKIN

THYROIDITIS A-THYROID A-THYROID A-THYROID

HYPOTHYROIDISM A-THYROID A-THYROID A-THYROID

EISUSC Susceptible Population 

Fraction

1. for all health 

effects

1. for all health 

effects

1. for all health 

effects

EITHRE Early Injury Dose Threshold

PRODROMAL VOMIT 0.5 0.5 0.5

DIARRHEA 1 1 1

PNEUMONITIS 9.2 9.2 9.2

SKIN ERYTHRMA 3 3 3

TRANSEPIDERMAL 10 10 10

THYROIDITIS 40 40 40

HYPOTHYROIDISM 2 2 2

ENDAT2

Control flag indicating only 

ATMOS and EARLY are to 

be run

.FALSE. .FALSE. .FALSE.

ENDEMP Time Duration for the 

Emergency Phase 604800 604800 604800

ESPEED Evaluation Speed

Initial Evacuation Phase, 

Cohort 1 2.235 2.235 2.235

Middle Evacuation 

Phase, Cohort 1 1.341 1.341 1.341

Late Evacuation Phase, 

Cohort 1 8.941 8.941 8.941

Initial Evacuation Phase, 

Cohort 2 2.235 2.235 2.235

Middle Evacuation 

Phase, Cohort 2 1.341 1.341 1.341

Late Evacuation Phase, 

Cohort 2 8.941 8.941 8.941

Initial Evacuation Phase, 

Cohort 3 8.941 8.941 8.941

Middle Evacuation 

Phase, Cohort 3 8.941 8.941 8.941
B-20

Variable Description LTSBO STSBO with RCIC 

Blackstart

STSBO without 

RCIC Blackstart

ESPEED Late Evacuation Phase, 

Cohort 3 8.941 8.941 8.941

Initial Evacuation Phase, 

Cohort 4 1.341 1.341 1.341

Middle Evacuation 

Phase, Cohort 4 8.941 8.941 8.941

Late Evacuation Phase, 

Cohort 4 8.941 8.941 8.941

Initial Evacuation Phase, 

Cohort 5 1.341 1.341 1.341

Middle Evacuation 

Phase, Cohort 5 8.941 8.941 8.941

Late Evacuation Phase, 

Cohort 5 8.941 8.941 8.941

Cohort 6 N/A N/A N/A

ESPGRD

Speed Multiplier to Account 

for Grid-Level Variations in 

Road Network

Table B-6 Table B-6 Table B-6

ESPMUL Speed Multiplier Employed 

During Precipitation 0.7 0.7 0.7

EVATYP Evacuation Type NETWORK NETWORK NETWORK

GSHFAC Groundshine Shielding 

Factors

Evacuation Shielding 

Factor for All but Cohort 4 0.5 0.5 0.5

Normal Activity

Shielding Factor for All 

but Cohort 4

0.18 0.18 0.18

Sheltering Shielding 

Factor for All but Cohort 4 0.1 0.1 0.1

Evacuation Shielding 

Factor for Cohort 4 0.5 0.5 0.5

Normal Activity 

Shielding Factor for 

Cohort 4

0.05 0.05 0.05

Sheltering Shielding 

Factor for Cohort 4 0.05 0.05 0.05

IDIREC Direction in Network 

Evacuation Model Table B-7 Table B-7 Table B-7

IPLUME Plume Model Dispersion 

Code 3 3 3

KIMODL Model Flag for KI Ingestion KI KI KI

LASMOV Last Ring in Movement Zone 17 17 17

NUMACA Number of Latent Cancer 

Health Effects 8 8 8

NUMEFA Number of Early Fatality 

Effects 3 3 3

NUMEIN Number of Early Injury 

Effects 7 7 7

NUMEVA Outer Boundary of 

Evacuation/Shelter Region

Cohort 1 - 15

Cohorts 2 to 6 - 12

Cohort 1 - 15

Cohorts 2 to 6 - 12

Cohort 1 - 15

Cohorts 2 to 6 - 12

NUMFIN Number of Fine Grid 

Subdivisions 7 7 7
B-21

Variable Description LTSBO STSBO with RCIC 

Blackstart

STSBO without 

RCIC Blackstart

ORGFLG Doses to be Calculated for 

Specified Organ

All TRUE for FGR-13

All TRUE for 

DOSFAC2 except for 

A-Lower LI and 

L-Liver, which are 

FALSE

All TRUE for FGR-13

All TRUE for 

DOSFAC2 except for 

A-Lower LI and 

L-Liver, which are 

FALSE

All TRUE for FGR-13

All TRUE for 

DOSFAC2 except for 

A-Lower LI and 

L-Liver, which are 

FALSE

OVRRID Wind Rose Probability 

Override .FALSE. .FALSE. .FALSE.

POPFLG Population Distribution Flag FILE FILE FILE

POPFRAC Population Fraction Ingesting 

KI

Cohort 1 and 2 - 1.0

Cohort 3 to 6 - 0.0

Cohort 1 and 2 - 1.0

Cohort 3 to 6 - 0.1

Cohort 1 and 2 - 1.0

Cohort 3 to 6 - 0.1

PROTIN [E] Inhalation Protection Factor -

evacuation 0.98 0.98 0.98

PROTIN 

[N]

Inhalation Protection Factor -

normal activity for all cohorts 

except Cohort 4

0.46 0.46 0.46

PROTIN 

[N]

Inhalation Protection Factor –

normal activity for Cohort 4 0.33 0.33 0.33

PROTIN [S] Inhalation Protection Factor -

sheltering

0.33 0.33 0.33

REFPNT Reference Time Point 

(ARRIVAL or SCRAM) ALARM ALARM ALARM

RESCON Emergency phase 

resuspension coefficient 0.0001 0.0001 0.0001

RESHAF Resuspension Concentration 

Half-Life 182000 182000 182000

RISCAT Risk by Weather-Category 

Flag .FALSE. .FALSE. .FALSE.

RISTHR Risk Threshold for Fatality 

Radius 0 0 0

SKPFAC 

[E]

Skin Protection Factors -

evacuation 0.98 0.98 0.98

SKPFAC 

[N]

Skin Protection Factors -

normal activity 0.46 0.46 0.46

SKPFAC 

[S]

Skin Protection Factors -

sheltering 0.33 0.33 0.33

TIMHOT Hot Spot Relocation Time 43200 43200 43200

TIMNRM Normal Relocation Time 86400 86400 86400

TRAVELP

OINT Evacuee Movement Option CENTERPOINT CENTERPOINT CENTERPOINT

WTFRAC Weighting Fraction 

Applicable to this Scenario

Cohort 1 0.2 0.2 0.2

Cohort 2 0.355 0.355 0.355

Cohort 3 0.372 0.372 0.372

Cohort 4 0.006 0.006 0.006

Cohort 5 0.062 0.062 0.062

Cohort 6 0.005 0.005 0.005

WTNAME Type of Weighting for 

Cohorts PEOPLE PEOPLE PEOPLE
B-22

Table B-6 Grid-Level Evacuation Speed Multipliers Used in the Peach Bottom 

Unmitigated LTSBO and STSBO with and without RCIC Blackstart 

Scenarios

Compass Sector

Radial Ring 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16

1 0.75 0.75 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

2 0.75 0.75 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

3 0.75 0.75 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

4 0.75 0.75 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

5 0.75 0.75 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

6 0.75 0.75 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

7 0.75 0.75 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

8 0.75 0.75 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

9 0.75 0.75 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

10 0.75 0.75 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

11 0.75 0.75 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

12 0.75 0.75 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

13 0.75 0.75 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

14 0.75 0.75 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

15 0.75 0.75 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

16 0.75 0.75 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

17 0.75 0.75 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

Compass Sector

Radial Ring 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 32

1 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

2 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

3 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

4 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

5 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

6 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

7 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

8 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

9 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

10 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

11 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

12 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

13 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

14 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.50 1.50 1.50 1.50 1.50 1.50 1.00 1.00

15 1.00 1.00 1.00 1.00 1.00 1.50 1.50 1.50 1.00 1.00 1.00 1.00 1.00 1.50 1.50 1.50

16 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

17 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00
B-23

Compass Sector

Radial Ring 33 34 35 36 37 38 39 40 41 42 43 44 45 46 47 48

1 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

2 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

3 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

4 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

5 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

6 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

7 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

8 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

9 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

10 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

11 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

12 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

13 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

14 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

15 1.50 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

16 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

17 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

Compass Sector

Radial Ring 49 50 51 52 53 54 55 56 57 58 59 60 61 62 63 64

1 100 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 0.75 0.75 0.75

2 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 0.75 0.75 0.75

3 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 0.75 0.75 0.75

4 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 0.75 0.75 0.75

5 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 0.75 0.75 0.75

6 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 0.75 0.75 0.75

7 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 0.75 0.75 0.75

8 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 0.75 0.75 0.75

9 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 0.75 0.75 0.75

10 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 0.75 0.75 0.75

11 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 0.75 0.75 0.75

12 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 0.75 0.75 0.75

13 1.00 1.00 1.00 1.00 1.00 1.00 0.75 1.00 1.00 1.00 1.00 1.00 1.00 0.75 0.75 0.75

14 1.00 1.00 1.00 1.00 1.00 1.00 0.75 1.00 1.00 1.00 1.00 1.00 1.00 0.75 0.75 0.75

15 1.00 1.00 1.00 1.00 1.00 1.00 0.75 1.00 1.00 1.00 1.00 1.00 1.00 0.75 0.75 0.75

16 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 0.75 0.75 0.75

17 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 0.75 0.75 0.75
B-24

Table B-7 Evacuation Direction Parameters Used in the Peach Bottom Unmitigated 

LTSBO and STSBO with and without RCIC Blackstart Scenarios

Compass Sector

Radial Ring 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16

1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1

2 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1

3 1 1 1 1 1 1 1 1 1 1 1 1 1 2 1 1

4 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1

5 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 4

6 1 1 1 1 1 1 1 1 1 1 1 1 1 1 4 1

7 2 2 1 2 2 1 2 2 1 4 2 1 4 2 2 1

8 1 4 1 1 4 2 1 4 2 1 1 4 2 2 1 1

9 1 1 4 2 1 1 2 1 1 4 1 4 1 4 1 1

10 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1

11 1 1 4 2 1 4 2 1 4 4 2 1 4 2 1 4

12 2 1 1 4 1 1 4 1 4 4 2 1 4 2 1 1

13 1 1 4 1 4 2 1 1 2 1 4 4 2 2 1 1

14 1 1 4 1 1 1 2 1 2 1 2 1 1 2 1 1

15 1 1 4 2 2 2 1 1 1 1 1 1 1 1 1 1

16 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 4

17 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1

Compass Sector

Radial Ring 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 32

1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1

2 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1

3 1 1 1 1 1 1 1 1 2 1 4 4 4 4 2 2

4 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1

5 1 1 1 1 1 1 1 1 1 2 2 2 2 2 2 1

6 1 1 1 1 1 1 1 1 1 1 2 2 2 2 2 2

7 1 1 1 1 1 1 1 1 1 1 2 2 1 4 4 2

8 2 1 1 1 4 4 4 1 1 1 1 1 1 4 1 4

9 1 2 1 1 1 4 4 4 1 1 4 1 2 2 2 1

10 1 1 1 1 1 1 1 1 1 1 1 1 1 4 4 4

11 4 2 2 2 2 2 2 1 4 4 1 1 4 2 2 1

12 4 4 2 2 2 1 4 1 2 1 2 2 1 4 4 2

13 1 1 1 1 1 2 1 2 2 1 1 2 1 4 2 2

14 1 1 1 1 1 4 1 1 1 1 2 1 4 1 1 1

15 1 1 1 1 1 1 1 4 4 4 2 2 2 1 1 1

16 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1

17 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1
B-25

Compass Sector

Radial Ring 33 34 35 36 37 38 39 40 41 42 43 44 45 46 47 48

1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1

2 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1

3 2 2 2 1 1 1 4 4 4 4 2 2 2 1 1 1

4 4 2 2 1 1 1 1 4 4 4 2 2 1 4 4 2

5 1 1 1 1 2 1 4 4 2 2 1 4 4 2 2 1

6 2 2 1 1 1 1 2 1 1 4 2 2 1 2 1 4

7 1 1 1 1 4 4 2 2 1 1 4 4 1 1 4 1

8 2 1 4 4 2 1 2 1 2 1 4 2 1 4 2 1

9 1 2 1 4 2 1 2 1 2 1 1 4 1 4 1 4

10 2 2 2 2 1 2 1 4 2 1 1 1 4 2 1 4

11 1 1 4 2 1 4 1 4 2 1 4 1 1 2 1 1

12 2 1 1 4 4 2 1 1 1 1 4 1 1 1 1 4

13 1 4 2 1 4 4 1 1 4 2 1 1 2 1 2 1

14 1 1 1 1 1 1 1 1 1 1 1 1 1 1 2 1

15 1 1 1 1 1 1 1 1 1 1 1 4 1 2 1 1

16 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1

17 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1

Compass Sector

Radial Ring 49 50 51 52 53 54 55 56 57 58 59 60 61 62 63 64

1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1

2 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1

3 1 1 1 1 1 1 1 1 1 1 1 1 1 4 1 1

4 1 4 4 2 1 1 4 2 1 4 4 4 4 4 1 1

5 4 4 2 2 1 4 4 2 1 1 1 1 1 1 1 1

6 4 2 2 1 4 4 4 4 4 4 4 1 1 1 1 1

7 4 2 2 1 4 4 4 4 4 4 1 1 1 1 1 1

8 1 4 2 1 4 4 4 2 2 1 1 1 1 1 2 2

9 4 2 2 1 4 2 1 4 1 4 1 1 1 2 2 2

10 4 2 2 1 1 1 1 1 4 4 1 1 1 1 1 1

11 4 4 4 2 2 2 1 4 4 4 2 2 1 4 2 2

12 4 4 2 2 2 2 1 4 2 1 1 2 2 1 4 2

13 4 4 2 2 2 1 1 4 2 1 4 2 1 4 1 1

14 1 4 4 2 2 1 1 4 1 4 1 1 1 1 1 1

15 1 4 2 1 2 1 1 2 1 1 2 2 2 1 2 1

16 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1

17 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1
B-26

Table B-8 CHRONC Input Parameters Used in the Peach Bottom Unmitigated LTSBO 

and STSBO with and without RCIC Blackstart Scenarios

Variable Description LTSBO STSBO with 

RCIC Blackstart

STSBO 

without RCIC 

Blackstart

CHNAME CHRONC Problem Identification

Peach Bottom 

with no FoodChain 

Modeling

Peach Bottom 

with no FoodChain Modeling

Peach Bottom 

with no FoodChain 

Modeling

CDFRM Farmland Decontamination Cost

Level 1 1330 1330 1330

Level 2 2960 2960 2960

CDNFRM Nonfarmland Decontamination Cost

Level 1 7110 7110 7110

Level 2 19000 19000 19000

CRTOCR Critical Organ for CHRONC Phase L-ICRP60ED L-ICRP60ED L-ICRP60ED

DPRATE Property Depreciation Rate 0.2 0.2 0.2

DLBCST Hourly Labor Cost for Decontamination 

Worker 84000 84000 84000

DPFRCT Farm Production Dairy Fraction PB Site File PB Site File PB Site File

DSCRLT Long-Term Phase Dose Criterion 0.005 0.005 0.005

DSCRTI Intermediate-Phase Dose Criterion 100000 100000 100000

DSRATE Societal Discount Rate for Property 0.12 0.12 0.12

DSRFCT Decontamination Factors

Level 1 3 3 3

Level 2 15 15 15

DUR_INTPH

AS Duration of the Intermediate Phase 0 0 0

EVACST Emergency Phase Cost of Evac./Reloc. 172 172 172

EXPTIM Maximum Exposure Time 1580000000 1580000000 1580000000

FDPATH COMIDA2 vs. MACCS Food Model 

Switch OFF OFF OFF

FRACLD Fraction of Area that is Land PB Site File PB Site File PB Site File

FRCFRM Fraction of Area Used for Farming PB Site File PB Site File PB Site File

FRFDL Fraction of Decontamination Cost for 

Labor

Level 1 0.3 0.3 0.3

Level 2 0.35 0.35 0.35

FRFIM Farm Wealth Improvements Fraction 0.25 0.25 0.25

FRMPRD Average Annual Farm Production PB Site File PB Site File PB Site File

FRNFIM Nonfarm Wealth Improvements Fraction 0.8 0.8 0.8

FRNFDL Nonfarm Labor Cost Fraction

Level 1 0.7 0.7 0.7

Level 2 0.5 0.5 0.5

GWCOEF Long-Term Groundshine Coefficients

Term 1 0.5 0.5 0.5

Term 2 0.5 0.5 0.5

KSWTCH Diagnostic Output Option Switch 0 0 0

LBRRATE Long-Term Breathing Rate 0.000266 0.000266 0.000266

LGSHFAC Long-Term Groundshine Protection 

Factor 0.18 0.18 0.18

LPROTIN Long-Term Inhalation Protection Factor 0.46 0.46 0.46
B-27

Variable Description LTSBO STSBO with 

RCIC Blackstart

STSBO 

without RCIC 

Blackstart

LVLDEC Number of Decontamination Levels 2 2 2

NGWTRM Number of Terms in Groundshine 

Weathering Equation 2 2 2

NRWTRM Number of Terms in Resuspension 

Weathering Equation 3 3 3

POPCST Per Capita Cost of Long-Term 

Relocation 12000 12000 12000

RELCST Relocation Cost per Person-Day 172 172 172

RWCOEF Long-Term Resuspension Factor 

Coefficients

Term 1 0.00001 0.00001 0.00001

Term 2 0.0000001 0.0000001 0.0000001

Term 3 0.00000001 0.00000001 0.00000001

TFWKF Fraction Farmland Worker Time in 

Contaminated Zone

Level 1 0.1 0.1 0.1

Level 2 0.33 0.33 0.33

TFWKNF Fraction Nonfarmland Worker Time in 

Contaminated Zone

Level 1 0.33 0.33 0.33

Level 2 0.33 0.33 0.33

TGWHLF Groundshine Weathering Half-Lives

Term 1 16000000 16000000 16000000

Term 2 2800000000 2800000000 2800000000

TIMDEC Decontamination Times

Level 1 5184000 5184000 5184000

Level 2 10368000 10368000 10368000

TMPACT Time Action Period Ends 31600000 31600000 31600000

TRWHLF Resuspension Weathering Half-Lives

Term 1 16000000 16000000 16000000

Term 2 160000000 160000000 160000000

Term 3 1600000000 1600000000 1600000000

VALWF Value of Farm Wealth 9040 9040 9040

VALWNF Value of Nonfarm Wealth 210000 210000 210000




UNITED STATES 

NUCLEAR REGULATORY COMMISSION

WASHINGTON, DC 20555-0001

--------------------

OFFICIAL BUSINESS
NUREG/CR

-7110,

Vol. 1, Rev. 1

State-of-the-Art Reactor Consequence Analyses Project:

Peach Bottom Integrated Analysis