185(*26

7HFKQLFDO%DVLVIRUWKe 

&RQWDLQPHQW3URWHFWLRQ

DQG5HOHDVH5HGXFWLRQ

5XOHPDNLQJIRU%RLOLQJ

:DWHU5HDFWRUVZLWK0DUN,

DQG0DUN,,&RQWDLQPHQWV

2IILFHRI1XFOHDU5HJXODWRU\5HVHDUFh
AVAILABILITY OF REFERENCE MATERIALS 

IN NRC PUBLICATIONS 

N RC Reference Material 

As of November 1999, you may electronically access 

NUREG-series publications and other NRC records at 

the NRC's Public Electronic Reading Room at 

http://www.nrc.gov/reading-rm.html. Publicly released 

records include, to name a few, NUREG-series 

publications; Federal Register notices; applicant, 

licensee, and vendor documents and correspondence; 

NRC correspondence and internal memoranda; bulletins 

and information notices; inspection and investigative 

reports; licensee event reports; and Commission papers 

and their attachments. 

NRC publications in the NUREG series, NRC 

regulations, and Title 10, "Energy," in the Code of 

Federal Regulations may also be purchased from one 

of these two sources. 

1. The Superintendent of Documents

U.S. Government Publishing Office

Mail Stop SSOP

Washington, DC 20402-0001

Internet: http://bookstore.gpo.gov

Telephone: 1-866-512-1800

Fax: (202) 512-2104

2. The National Technical Information Service

5301 Shawnee Road

Alexandria, VA 22161-0002

http://www.ntis.gov

1-800-553-6847 or, locally, (703) 605-6000

A single copy of each NRC draft report for comment is 

available free, to the extent of supply, upon written 

request as follows: 

U.S. Nuclear Regulatory Commission 

Office of Administration 

Multimedia, Graphics, Storage, and 

Distribution Branch 

Washington, DC 20555-0001 

E-mail: distribution.resource@nrc.gov

Facsimile: (301) 415-2289

Some publications in the NUREG series that are 

posted at the NRC's Web site address 

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

are updated periodically and may differ from the last 

printed version. Although references to material 

found on a Web site bear the date the material was 

accessed, the material available on the date cited 

may subsequently be removed from the site. 

SR-CR 08/2016 

Non-NRC Reference Material 

Documents available from public and special technical 

libraries include all open literature items, such as books, 

journal articles, transactions, Federal Register notices, 

Federal and State legislation, and congressional reports. 

Such documents as theses, dissertations, foreign reports 

and translations, and non-NRC conference proceedings 

may be purchased from their sponsoring organization. 

Copies of industry codes and standards used in a 

substantive manner in the NRC regulatory process are 

maintained atThe NRC Technical Library 

Two White Flint North 

11545 Rockville Pike 

Rockville, MD 20852-2738 

These standards are available in the library for reference 

use by the public. Codes and standards are usually 

copyrighted and may be purchased from the originating 

organization or, if they are American National Standards, 

fromAmerican National Standards Institute 

11 West 42nd Street 

New York, NY 10036-8002 

http://www.ansi.org 

(212) 642-4900

Legally binding regulatory requirements are stated only 

in laws; NRC regulations; licenses, including technical 

specifications; or orders, not in NUREG-series 

publications. The views expressed in contractorprepared publications in this series are not necessarily 

those of the NRC. 

The NU REG series comprises (1) technical and 

administrative reports and books prepared by the staff 

(NUREG-XXXX) or agency contractors 

(NUREG/CR-XXXX), (2) proceedings of conferences 

(NUREG/CP-XXXX), (3) reports resulting from 

international agreements (NUREG/IA-XXXX), (4) 

brochures (NUREG/BR-XXXX), and (5) compilations of 

legal decisions and orders of the Commission and 

Atomic and Safety Licensing Boards and of Directors' 

decisions under Section 2.206 of NRC's regulations 

(NUREG-0750). 

DISCLAIMER: This report was prepared as an account of 

work sponsored by an agency of the U.S. Government. 

Neither the U.S. Government nor any agency thereof, nor 

any employee, makes any warranty, expressed or 

implied, or assumes any legal liability or responsibility for 

any third party's use, or the results of such use, of any 

information, apparatus, product, or process disclosed in 

this publication, or represents that its use by such third 

party would not infringe privately owned rights. 
185(*

7HFKQLFDO%DVLVIRUWKH

&RQWDLQPHQW3URWHFWLRQ

DQG5HOHDVH5HGXFWLRQ

5XOHPDNLQJIRU%RLOLQJ

:DWHU5HDFWRUVZLWK0DUN,

DQG0DUN,,&RQWDLQPHQWV

0DQXVFULSW &RPSOHWHG 'HFHPEHU

'DWH3XEOLVKHG March8

3UHSDUHGE\

-RQDWKDQ%DUU, 6XGKDPD\%DVX,

+RVVHLQ (VPDLOL, DQG 0DUWLQ6WXW]NH

2IILFHRI1XFOHDU5HJXODWRU\5HVHDUFK

iii

ABSTRACT

The work documented in this report focused on developing the technical basis for a potential 

rulemaking action on containment protection and release reduction (CPRR) strategies for boiling 

water reactors with Mark I and Mark II containments. The work covered three areas of analyses: 

(1) accident sequence analysis (event tree development) to identify accident sequences initiated

by extended loss of ac power (ELAP) due to internal events and seismic events deemed to be the

most significant risk contributors; (2) accident progression analysis of these sequences and

assessment of radiological source terms; and (3) analysis of offsite consequences including

individual early fatality risk and latent cancer fatality risk, land contamination, and economic

consequences. The calculated offsite consequences were weighted by accident frequency to

assess relative public health risk reduction associated with various CPRR strategies. Important

findings and key insights from the work are delineated.

v 

TABLE OF CONTENTS

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

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

LIST OF TABLES....................................................................................................................xiii

EXECUTIVE SUMMARY ..........................................................................................................xv

ACKNOWLEDGMENTS........................................................................................................xxiii

ABBREVIATIONS AND ACRONYMS ....................................................................................xxv

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

1.1 Containment Venting Issue – A Historical Perspective.............................................1-1

1.2 Post-Fukushima Development .................................................................................1-3

1.3 Content and Structure of Report ..............................................................................1-6

1.4 References for Chapter 1.........................................................................................1-9

2 ACCIDENT SEQUENCE ANALYSIS................................................................................2-1

2.1 Identification of CPRR Strategies.............................................................................2-2

2.1.1 Influencing Factors.....................................................................................2-2

2.1.2 Regulatory Analysis Sub-Alternatives ........................................................2-4

2.2 Technical Approach .................................................................................................2-8

2.2.1 Rationale Used to Develop the Technical Approach ..................................2-9

2.2.2 Summary of the Technical Approach .........................................................2-9

2.2.3 Logic Model Development........................................................................ 2-10

2.2.4 Supporting Data Analysis......................................................................... 2-16

2.2.5 Risk Quantification ................................................................................... 2-19

2.3 Results................................................................................................................... 2-22

2.3.1 Results from the CDET Quantification...................................................... 2-22

2.3.2 Results from the APET Quantification ...................................................... 2-31

2.4 References for Chapter 2.......................................................................................2-32

3 ACCIDENT PROGRESSION ANALYSIS .........................................................................3-1

3.1 MELCOR BWR Models............................................................................................3-2

3.1.1 Description of Mark I MELCOR Model .......................................................3-2

3.1.2 Description of Mark II MELCOR Model .................................................... 3-13

3.1.3 Description of Mitigation Systems and Failure Modes .............................. 3-19

3.2 Scenario Development and Run Matrix.................................................................. 3-22

3.3 Accident Progression Results ................................................................................ 3-30

3.3.1 BWR Mark I Results................................................................................. 3-30

3.3.2 BWR Mark II Results................................................................................ 3-59

3.4 Consideration of Uncertainty.................................................................................. 3-84

3.5 Accident Progression Summary and Conclusions.................................................. 3-86

3.6 References for Chapter 3....................................................................................... 3-87
vi

4 OFFSITE CONSEQUENCE ANALYSIS ...........................................................................4-1

4.1 MACCS Conceptual Models ....................................................................................4-1

4.1.1 Atmospheric Transport and Dispersion (ATD)............................................4-3

4.1.2 Phases of the MACCS Conceptual Model..................................................4-3

4.1.3 Offsite Consequence Measures .................................................................4-6

4.2 Modeling Approach..................................................................................................4-8

4.2.1 Radionuclide Release ................................................................................4-9

4.2.2 Calculational Grid..................................................................................... 4-15

4.2.3 Site Data......................................................................................................4-16

4.2.4 Meteorological Data ................................................................................. 4-18

4.2.5 Atmospheric Transport and Dispersion .................................................... 4-20

4.2.6 Early Phase Exposure Pathways, Protective Actions, and Costs ............. 4-24

4.2.7 Intermediate Phase Exposure Pathways, Protective Actions, and 

Costs ....................................................................................................... 4-36

4.2.8 Long-Term Phase Exposure Pathways, Protective Actions, and Costs ....4-37

4.2.9 Dosimetry....................................................................................................4-39

4.2.10 Health Effects .......................................................................................... 4-41

4.3 Offsite Consequence Results................................................................................. 4-43

4.3.1 Individual Early Fatality Risk .................................................................... 4-45

4.3.2 Individual Latent Cancer Fatality Risk ...................................................... 4-45

4.3.3 Population Dose....................................................................................... 4-49

4.3.4 Offsite Economic Cost.............................................................................. 4-53

4.3.5 Land Contamination and Population Subject to Long-Term Protective 

Actions..................................................................................................... 4-56

4.4 Sensitivity Analyses ............................................................................................... 4-59

4.4.1 Approach....................................................................................................4-59

4.4.2 Sensitivity Calculations Supporting Regulatory Analysis: Results and 

Discussion ............................................................................................... 4-66

4.4.3 Environmental Release Timing Sensitivity................................................ 4-72

4.4.4 Evacuation Sensitivity Calculations for Fastest Release 

(Mark I Case 49)...................................................................................... 4-73

4.5 Conditional Offsite Consequences of Different CPRR Alternatives ........................ 4-75

4.6 Offsite Consequence Summary and Conclusions .................................................. 4-80

4.7 References for Chapter 4....................................................................................... 4-82

5 RISK INTEGRATION........................................................................................................5-1

5.1 Results.....................................................................................................................5-1

5.2 Sensitivity and Uncertainty Analysis.........................................................................5-5

5.2.1 Sensitivity of Core Damage Frequency to Human Error Probabilities.........5-6

5.2.2 Parametric Uncertainty Analysis ................................................................5-7

5.3 Conclusions .............................................................................................................5-9

5.4 References for Chapter 5....................................................................................... 5-10

6 TECHNICAL EVALUATION SUMMARY AND KEY INSIGHTS........................................6-1

6.1 Key Insights from Accident Sequence Analysis and Risk Integration .......................6-2

6.2 Key Insights from Accident Progression Analysis.....................................................6-3

6.3 Key Insights from Offsite Consequence Analysis .....................................................6-4

6.4 Concluding Remarks................................................................................................6-5
vii

APPENDIX A CORE DAMAGE EVENT TREES ................................................................A-1

APPENDIX B CORE DAMAGE EVENT TREE SUPPORTING INFORMATION.................B-1

APPENDIX C ACCIDENT PROGRESSION EVENT TREES..............................................C-1

APPENDIX E INTERNAL EVENT ELAP FREQUENCY SUPPORTING 

INFORMATION............................................................................................E-1

APPENDIX F SEISMIC ELAP FREQUENCY SUPPORTING INFORMATION...................F-1

APPENDIX G PLANT DAMAGE STATE FREQUENCIES .................................................G-1

APPENDIX H SOURCE TERM INFORMATION TO MAP EACH MELCOR CASE 

VARIATION TO A MACCS SOURCE TERM BIN........................................H-1

APPENDIX I

APPENDIX J

MACCS SENSITIVITY CALCULATION RESULTS ......................................I-1

OFFSITE CONSEQUENCE RESULTS FOR COMPOSITE MELCOR 

CASES.........................................................................................................J-1

APPENDIX D ACCIDENT PROGRESSION EVENT TREE SUPPORTING 

INFORMATION................................................................................................D-1

ix

LIST OF FIGURES

Figure 2-1 Modular approach to event tree development.................................................... 2-14

Figure 2-2 Core-damage event tree quantification .............................................................. 2-20

Figure 2-3 Contributions to ELAP frequency and core-damage frequency ......................... 2-25

Figure 2-4 Identification of significant events with respect to core damage frequency........ 2-27

Figure 3-1 MELCOR reactor coolant system nodalization for Mark I analysis....................... 3-3

Figure 3-2 MELCOR reactor vessel nodalization for Mark I analysis .................................... 3-4

Figure 3-3 MELCOR core nodalization for Mark I analysis.................................................... 3-7

Figure 3-4 MELCOR nodalization of the primary containment for Mark I analysis ................ 3-9

Figure 3-5 Discrete representation of debris spreading in the cavity................................... 3-11

Figure 3-6 MELCOR nodalization of the reactor building for Mark I analysis ...................... 3-12

Figure 3-7 MELCOR reactor vessel nodalization for Mark II analysis ................................. 3-14

Figure 3-8 MELCOR core nodalization for Mark II analysis................................................. 3-17

Figure 3-9 MELCOR nodalization of the primary containment for Mark II analysis ............. 3-18

Figure 3-10 Mark I RPV pressure for Case 9 (SAWM).......................................................... 3-31

Figure 3-11 Mark I RPV water level for Case 9 (SAWM)....................................................... 3-32

Figure 3-12 Mark I fuel cladding temperature in Ring 1 for Case 9 (SAWM)......................... 3-33

Figure 3-13 Mark I fuel cladding temperatures at core mid-plane for Case 9 (SAWM) ......... 3-34

Figure 3-14 Mark I core degradation and relocation for Case 9 (SAWM).............................. 3-34

Figure 3-15 Mark I suppression pool temperature for Case 9 (SAWM)................................. 3-35

Figure 3-16 Mark I containment pressure and vent flow for Case 9 (SAWM)........................ 3-36

Figure 3-17 Mark I containment water level for Case 9 (SAWM)........................................... 3-37

Figure 3-18 Mark I RPV water injection rate for Case 9 (SAWM).......................................... 3-37

Figure 3-19 Mark I containment atmosphere temperature for Case 9 (SAWM) .................... 3-38

Figure 3-20 Mark I debris spreading in DW for Case 9 (SAWM)........................................... 3-38

Figure 3-21 Mark I cesium fission product distribution for Case 9 (SAWM) .......................... 3-40

Figure 3-22 Mark I iodine fission product distribution for Case 9 (SAWM) ............................ 3-40

Figure 3-23 Mark I CsI particle size distribution for Case 9 (SAWM)..................................... 3-41

Figure 3-24 Mark I RPV pressure for Case 10 (SAWA)......................................................... 3-42

Figure 3-25 Mark I containment pressure and vent flow for Case 10 (SAWA) ...................... 3-43

Figure 3-26 Mark I containment water level for Case 10 (SAWA) ......................................... 3-43

Figure 3-27 Mark I cesium fission product distribution for Case 10 (SAWA)......................... 3-44

Figure 3-28 Mark I iodine fission product distribution for Case 10 (SAWA)........................... 3-44

Figure 3-29 Mark I containment water level for Case 1 ......................................................... 3-45

Figure 3-30 Mark I ex-vessel debris temperature for Case 1................................................. 3-46

Figure 3-31 Mark I debris spreading in DW for Case 1.......................................................... 3-47

Figure 3-32 Mark I refueling bay gas concentration for Case 1............................................. 3-48

Figure 3-33 Mark I containment atmosphere temperature for Case 1................................... 3-49

Figure 3-34 Mark I cesium fission product distribution for Case 1 ......................................... 3-49

Figure 3-35 Mark I iodine fission product distribution for Case 1........................................... 3-50
x 

Figure 3-36 Mark I cesium environmental release fraction.................................................... 3-51

Figure 3-37 Effect of water injection on selected cesium releases for selected cases .......... 3-52

Figure 3-38 Mark I containment gas and structure temperatures .......................................... 3-52

Figure 3-39 Mark I maximum containment gas and DW liner temperatures.......................... 3-53

Figure 3-40 Mark I hydrogen generation and transport for Case 9........................................ 3-54

Figure 3-41 Mark I drywell gas distribution for Case 9........................................................... 3-54

Figure 3-42 Mark I wetwell gas distribution for Case 9 .......................................................... 3-55

Figure 3-43 Mark I wetwell vent line gas distribution for Case 9............................................ 3-55

Figure 3-44 Mark I wetwell vent line gas distribution at venting for Case 9 ........................... 3-56

Figure 3-45 Mark I RPV water level for Case 9 (IVR)............................................................ 3-57

Figure 3-46 Mark I cesium release fraction from fuel for Case 9 (IVR).................................. 3-57

Figure 3-47 Mark I cesium release fraction to environment for Case 9 (IVR)........................ 3-58

Figure 3-48 Mark II RPV pressure for Case 11...................................................................... 3-61

Figure 3-49 Mark II RPV water level for Case 11 .................................................................. 3-62

Figure 3-50 Mark II fuel cladding temperature in Ring 1 for Case 11 .................................... 3-62

Figure 3-51 Mark II fuel cladding temperatures at core mid-plane for Case 11..................... 3-63

Figure 3-52 Mark II core degradation and relocation for Case 11 ......................................... 3-64

Figure 3-53 Mark II concrete ablation depth due to MCCI for Case 11 ................................. 3-65

Figure 3-54 Mark II containment pressure and vent flow for Case 11 ................................... 3-65

Figure 3-55 Mark II suppression pool temperature for Case 11 ............................................ 3-66

Figure 3-56 Mark II containment water levels for Case 11 .................................................... 3-66

Figure 3-57 Mark II containment atmosphere temperatures for Case 11 .............................. 3-67

Figure 3-58 Mark II cesium fission product distribution for Case 11 ...................................... 3-68

Figure 3-59 Mark II iodine fission product distribution for Case 11........................................ 3-68

Figure 3-60 Mark II RPV pressure for Case 1........................................................................ 3-71

Figure 3-61 Mark II RPV water level for Case 1 .................................................................... 3-71

Figure 3-62 Mark II core degradation and relocation for Case 1 ........................................... 3-72

Figure 3-63 Mark II cesium fission product distribution for Case 1 ........................................ 3-72

Figure 3-64 Mark II iodine fission product distribution for Case 1.......................................... 3-73

Figure 3-65 Mark II containment pressure and vent flow for Case 1 ..................................... 3-73

Figure 3-66 Mark II containment atmosphere temperatures for Case 1 ................................ 3-74

Figure 3-67 Mark II containment water levels for Case 1 ...................................................... 3-74

Figure 3-68 Mark II containment configurations..................................................................... 3-75

Figure 3-69 Mark II fission product transport paths during accident phases.......................... 3-76

Figure 3-70 Mark II environment release sensitivity to suppression pool bypass timing ....... 3-78

Figure 3-71 Mark II with flooded lower reactor cavity sensitivity to suppression pool 

bypass timing...................................................................................................... 3-79

Figure 3-72 Mark II hydrogen generation and transport for Case 11p1................................. 3-80

Figure 3-73 Mark II drywell gas distribution for Case 11p1.................................................... 3-80

Figure 3-74 Mark II wetwell gas distribution for Case 11p1................................................... 3-81

Figure 3-75 Mark II wetwell vent gas distribution for Case 11p1 ........................................... 3-81
xi

Figure 3-76 Mark II cesium environmental release fraction................................................... 3-82

Figure 3-77 Cesium release fractions for Mark II cases and sensitivities as well as 

corresponding Mark I cases ............................................................................... 3-83

Figure 4-1 Wind Rose for Peach Bottom 2006 and Limerick 2013...................................... 4-19

Figure 4-2 Peach Bottom emergency response timeline and travel speeds (mph) ............. 4-30

Figure 4-3 Limerick emergency response timeline and travel speeds (mph) ...................... 4-32

Figure 4-4 Mark I source terms – individual latent cancer fatality risk per event ................. 4-47

Figure 4-5 Mark II source terms – individual latent cancer fatality risk per event ................ 4-47

Figure 4-6 Mark I source terms – individual latent cancer fatality risk per event with 

breakdown by accident phase............................................................................ 4-48

Figure 4-7 Mark II source terms – individual latent cancer fatality risk per event with 

breakdown by accident phase............................................................................ 4-49

Figure 4-8 Mark I source terms – population dose per event (person-rem)......................... 4-50

Figure 4-9 Mark II source terms – population dose per event (person-rem)........................ 4-50

Figure 4-10 Mark I source terms – population dose per event (person-rem) with 

breakdown by accident phase and exposure pathway (0-50 miles)................... 4-52

Figure 4-11 Mark II source terms – population dose per event (person-rem) with 

breakdown by accident phase and exposure pathway (0-50 miles)................... 4-52

Figure 4-12 Mark I source terms – offsite economic cost ($ 2013)........................................ 4-54

Figure 4-13 Mark II source terms – offsite economic cost ($ 2013)....................................... 4-54

Figure 4-14 Mark I source terms – offsite economic cost ($ 2013) with breakdown by 

accident phase and protective action type (0-50 miles) ..................................... 4-55

Figure 4-15 Mark II source terms – offsite economic cost ($ 2013) with breakdown by 

accident phase and protective action type (0-50 miles) ..................................... 4-56

Figure 4-16 Mark I source terms – land exceeding long-term phase habitability criterion 

(square miles) per event..................................................................................... 4-57

Figure 4-17 Mark II source terms – land exceeding long-term phase habitability criterion 

(square miles) per event..................................................................................... 4-57

Figure 4-18 Mark I source terms – population displaced from interdicted and 

condemned land per event................................................................................. 4-58

Figure 4-19 Mark II source terms – population displaced from interdicted and 

condemned land per event................................................................................. 4-59

Figure 4-20 CRF for conditional ILCF risk (0-10 mi) for evacuation sensitivity 

calculations for BWR Mark I MELCOR Case 49 ................................................ 4-74

Figure 4-21 Conditional Mark I offsite consequences for each CPRR alternative as a 

percentage of the case with no external water addition. .................................... 4-78

Figure 4-22 Conditional Mark II offsite consequences for each CPRR alternative as a 

percentage of the case with no external water addition. .................................... 4-79

Figure 5-1 Comparison of regulatory analysis sub-alternatives using individual latent 

cancer fatality risk (0-10 miles)............................................................................. 5-3

Figure 5-2 Comparison of regulatory analysis sub-alternatives using population dose 

risk (0-50 miles).................................................................................................... 5-3

Figure 5-3 Comparison of regulatory analysis sub-alternatives using offsite cost risk 

(0–50 miles).......................................................................................................... 5-4
xii

Figure 5-4 Comparison of regulatory analysis sub-alternatives using area of land 

exceeding long-term habitability criterion (0-50 miles) ......................................... 5-4

Figure 5-5 Comparison of regulatory analysis sub-alternatives using population 

subject to long-term habitability criterion (0-50 miles) .......................................... 5-5

Figure 5-6 En masse sensitivity of core-damage frequency to human error probability 

for Wetwell-First venting strategy ......................................................................... 5-6

Figure 5-7 En Masse sensitivity of core-damage frequency to human error probability 

for Drywell-First (passive actuation) venting strategy........................................... 5-6

Figure 5-8 En Masse sensitivity of core-damage frequency to human error probability 

for Drywell-First (manual actuation) venting strategy ........................................... 5-7

Figure 5-9 Results of the parametric uncertainty analysis ..................................................... 5-8
xiii

LIST OF TABLES

Table 2-1 Consideration of BWR Mark I and Mark II plants in the accident sequence

analysis................................................................................................................. 2-1

Table 2-2 Regulatory analysis sub-alternatives.................................................................... 2-7

Table 2-3 Plant damage state naming scheme.................................................................. 2-15

Table 2-4 Release category naming scheme ..................................................................... 2-16

Table 2-5 Mapping release categories to MELCOR cases................................................. 2-21

Table 2-6 Core-damage frequencies by site for Wetwell-First strategy.............................. 2-23

Table 2-7 Core-damage frequencies by site for Drywell-First (passive) strategy ............... 2-23

Table 2-8 Core-damage frequencies by site for Drywell-First (manual) strategy ............... 2-24

Table 2-9 Importance measures with respect to core-damage frequency.......................... 2-26

Table 2-10 Significant plant damage states – Wetwell First strategy ................................... 2-28

Table 2-11 Significant plant damage states – Drywell-First (passive) strategy .................... 2-29

Table 2-12 Significant plant damage states – Drywell-First (manual) strategy..................... 2-30

Table 2-13 Containment failure mode probabilities .............................................................. 2-31

Table 2-14 Frequencies of significant release categories .................................................... 2-32

Table 3-1 Listing of the alternative options and actions...................................................... 3-26

Table 3-2 Calculation matrix for a representative BWR with Mark I containment............... 3-28

Table 3-3 Calculation matrix for a representative BWR with Mark II containment.............. 3-29

Table 3-4 Timing of key events for selected Mark I cases.................................................. 3-30

Table 3-5 Summary of main parameters for Mark I analysis .............................................. 3-59

Table 3-6 Timing of key events for selected Mark II cases................................................. 3-60

Table 3-7 Summary of main parameters for Mark II analysis ............................................. 3-83

Table 4-1 Phases of the MACCS conceptual model ............................................................ 4-4

Table 4-2 Binning strategy for Mark I source terms............................................................ 4-12

Table 4-3 Identification of source term bin for each Mark I source term case .................... 4-13

Table 4-4 Binning strategy for Mark II source terms........................................................... 4-14

Table 4-5 Identification of source term bin for each Mark II source term case ................... 4-14

Table 4-6 External filter effectiveness for three example BWR Mark I cases..................... 4-15

Table 4-7 Radial boundaries used in the polar calculational grid ....................................... 4-16

Table 4-8 Site data types and sources ............................................................................... 4-17

Table 4-9 Population growth multipliers for Peach Bottom and Limerick ........................... 4-17

Table 4-10 Summary of meteorological data sets used for Peach Bottom and Limerick ..... 4-18

Table 4-11 Seasonal diurnal mixing heights (m) used for Peach Bottom and Limerick ....... 4-20

Table 4-12 Surface roughness calculation for Peach Bottom and Limerick ......................... 4-22

Table 4-13 Dry deposition velocities for Peach Bottom and Limerick................................... 4-23

Table 4-14 Emergency response timeline ............................................................................ 4-25

Table 4-15 Peach Bottom evacuation cohorts and parameters............................................ 4-30

Table 4-16 Limerick evacuation cohorts and parameters..................................................... 4-31

Table 4-17 Shielding and protection factors for evacuating cohorts..................................... 4-33
xiv

Table 4-18 Shielding and protection factors for nonevacuating population .......................... 4-33

Table 4-19 Early phase relocation parameters..................................................................... 4-34

Table 4-20 Decontamination plan data for Peach Bottom and Limerick............................... 4-39

Table 4-21 Exposure pathways and population types used to compute individual and 

collective health effects ...................................................................................... 4-42

Table 4-22 MACCS results for 18 Mark I source term bins .................................................. 4-44

Table 4-23 MACCS results for 9 Mark II source term bins ................................................... 4-45

Table 4-24 Sites selected for population sensitivity calculations .......................................... 4-60

Table 4-25 Mark I site population information....................................................................... 4-61

Table 4-26 Mark II site population information...................................................................... 4-61

Table 4-27 Population growth multipliers for selected sensitivity sites ................................. 4-62

Table 4-28 Selection of source terms for Mark I sensitivity calculations............................... 4-62

Table 4-29 Selection of source terms for Mark II sensitivity calculations.............................. 4-63

Table 4-30 Intermediate and long-term phase sensitivity calculation parameters................ 4-66

Table 4-31 Ratio of consequences for evacuation delay sensitivity cases to baseline 

cases .................................................................................................................. 4-67

Table 4-32 Ratio of consequences for larger nonevacuating cohort sensitivity cases to 

baseline cases.................................................................................................... 4-68

Table 4-33 Ratio of consequences for 1-year intermediate phase duration sensitivity 

cases to baseline cases ..................................................................................... 4-69

Table 4-34 Ratio of consequences for zero intermediate phase sensitivity cases to 

baseline cases.................................................................................................... 4-70

Table 4-35 Ratio of consequences for 2 rem-per-year long-term phase habitability 

criterion sensitivity cases to baseline cases ....................................................... 4-71

Table 4-36 Results for baseline cases with different site files .............................................. 4-72

Table 4-37 Comparison of source term environmental release timing for Mark I Case 

48 and Mark I Case 50 ....................................................................................... 4-73

Table 4-38 Average Mark I conditional offsite consequences for the different MELCOR 

cases associated with the CPRR alternatives.................................................... 4-76

Table 4-39 Average Mark II conditional offsite consequences for the different MELCOR 

cases associated with the CPRR alternatives.................................................... 4-77

Table 4-40 Conditional Mark I and Mark II offsite consequences and consequence 

reduction factor (CRF) for each CPRR alternative ............................................. 4-77

Table 5-1 Risk estimates by regulatory analysis sub-alternative.......................................... 5-2

Table 5-2 Uncertainty analysis inputs................................................................................... 5-7
xv

EXECUTIVE SUMMARY

This report documents research conducted by the staff of the Office of Nuclear Regulatory 

Research (RES) to support the agency initiative to evaluate containment venting and filtration for 

boiling water reactor (BWR) Mark I and Mark II containments. Containment venting and filtration

was identified by the Near Term Task Force (NTTF), put together in the aftermath of the 

Fukushima Dai-ichi nuclear power plant accident in Japan, for all containment types but prioritized 

as a short-term action item for the two containment types mentioned above. The initiative was led 

by the Office of Nuclear Reactor Regulation (NRR) Japan Lessons Learned Directorate (JLD).

The containment venting and filtration issue has a long history behind it. Subsequent to the Three 

Mile Island Unit 2 nuclear plant core melt event in 1979, controlled (and potentially filtered) release 

was identified in NUREG-0585 as a favorable alternative to catastrophic failure of the containment.

In SECY-88-147, “Integration Plan for Closure of Severe Accident Issues,” dated May 25, 1988, the 

U.S. Nuclear Regulatory Commission (NRC) staff presented to the Commission its plan to evaluate 

potential generic severe accident containment vulnerabilities in a research effort entitled the 

“Containment Performance Improvement Program” (CPIP). All light water reactor (LWR) 

containment types were considered in the program. Potential improvements for Mark I 

containments, documented in NUREG/CR-5225, included: (1) hydrogen control; (2) alternate water 

supply for reactor vessel injection and containment drywell sprays; (3) containment pressure relief 

capability (venting); (4) enhanced reactor pressure vessel (RPV) depressurization system reliability; 

(5) core debris control; and (6) emergency procedures and training.

Potential improvements for Mark II containments, identified in NUREG/CR-5528, were largely the 

same as those for Mark I containments. However, less definitive conclusions were reached 

regarding venting of Mark II containments for a number of reasons. Potential improvements for 

other containment types were not studied in as much detail as for Mark I and Mark II 

containments.

The events at the Fukushima Dai-ichi nuclear power plant brought the containment venting and 

filtration issue to the forefront once again. In SECY-11-0137, the NRC staff described its 

proposals for regulatory actions to address containment venting related NTTF recommendations.

The insights gained from Fukushima Dai-ichi led the agency to impose additional requirements for 

reliable hardened venting systems for plants with Mark I and Mark II containments through the 

Order EA-12-050, on the basis of ensuring adequate protection of public health and safety. This 

order provided requirements to ensure reliable operation of the hardened venting system; 

however, it did not include requirements for reliable operation under severe accident conditions.

One recommendation, identified in SECY-11-0137 and further developed in SECY-12-0025, was 

consideration of additional performance requirements, including filters, for hardened containment 

vent systems for Mark I and Mark II containment designs. This would improve reliability during 

severe accident conditions and limit the release of radioactive materials if the venting systems 

were used after significant core damage had occurred.

To support formulation of possible additional regulatory actions related to the performance of Mark 

I and Mark II containments during severe accidents, RES performed a systematic analysis of 

accident source terms and consequences for a representative BWR plant with Mark I containment 

and for representative accident scenarios, predicated on an extended loss of alternating current 

power (ELAP) as in Fukushima. The technical analysis, completed by RES in 2012 and 

documented in SECY-12-0157, concluded that the plants with Mark I and Mark II containments 

would benefit from a containment venting and water addition strategy for a vast majority of severe 

accident sequences (whether the vent includes an engineered filter or not). Of particular 
xvi

importance to note in SECY-12-0157 was an accident scenario without venting which resulted in 

containment failure with significant release of radioactive materials to the environment. This lent 

further support to having a venting provision, capable of operating under severe accident 

conditions, for Mark I and Mark II containments. This also affirmed the Commission’s earlier 

action, in the immediate aftermath of Fukushima, to impose Order EA-12-050 on the basis of 

ensuring adequate protection of public health and safety. The Order EA-12-050 was subsequently 

rescinded and replaced by the Order EA-13-109 emphasizing, in particular, the functionality of the 

reliable hardened vents under severe accident conditions.

Subsequently, the RES staff initiated additional technical work to develop technical basis for 

potential rulemaking involving filtering strategies with drywell filtration and severe accident 

management of BWR Mark I and II containments. This additional technical work was an extension 

of the previous work by RES documented in SECY-12-0157, and covered three areas of 

analyses: (1) accident sequence analysis (event tree development) to identify accident sequences 

initiated by extended loss of ac power (ELAP) due to internal events and seismic events deemed 

to be the most significant risk contributors; (2) accident progression analysis of these sequences 

and assessment of radiological source terms; and (3) analysis of offsite consequences using 

NRC’s probabilistic offsite consequence computer code, MACCS (MELCOR Accident 

Consequence Code System). The calculated offsite consequences were weighted by accident 

frequency to assess relative public health risk reduction associated with various containment 

protection and release reduction measures. 

The report documents the technical work to support the regulatory basis1 for potential CPRR 

rulemaking activity (SECY-15-0085). Chapter 1 provides a historical perspective of the 

containment venting issue and a brief narrative on the resurgence of the issue in the aftermath of 

Fukushima. Chapter 2 of the report documents the staff’s accident sequence analysis that was 

used to select risk-significant accident sequences and quantify their frequency. Chapter 3 

describes in detail the MELCOR analysis, the scope of which falls broadly into two categories: (1) 

reactor systems and containment thermal-hydraulics under severe accident conditions, and (2) 

assessment of source terms (i.e., timing and magnitude of fission product releases to the 

environment). Chapter 4 describes in detail the offsite consequence analysis using MACCS, 

which evaluates health risks as well as land contamination and economic consequences. Chapter 

5 describes the results of accident sequence analysis integrated with the results of offsite 

consequence analysis, and provides a discussion of integrated risk. Key insights from the 

technical work are summarized in Chapter 6 in three major areas: risk evaluation, source term 

assessment using MELCOR, and offsite consequence assessment using MACCS. In summary, 

the report provides technical inputs to the regulatory analysis, which is documented in an 

enclosure to SECY-15-0085. 

The technical work in each of the three major disciplines was based on a number of assumptions 

or considerations that are delineated in detail in the respective chapters describing the work.

Some top-level considerations are:

1 The complete draft regulatory basis attached to SECY-15-0085 includes relevant background, a regulatory 

evaluation, a technical evaluation that is based on the technical work documented in this report, performance 

criteria information, impact analysis, and the staff’s conclusion. 
xvii

• SRM-SECY-12-0157 direction that the regulatory basis should assume the benefits of

severe accident capable hardened venting systems (EA-13-109) that would accrue

equally to engineered filters and to filtration strategies;

• Consideration of a variety of performance criteria such as a decontamination factor,

equipment and procedure availability similar to those required to implement 10 CFR

50.54(hh)(2), or other measures (e.g., quantitative health objective or QHO) that may be

developed during the stakeholder engagement;

• Consideration of requirements associated with measures to enhance the capability to

maintain containment integrity and to cool core debris; and

• SRM-SECY-12-0110 direction that economic consequences should not be treated as

equivalent in regulatory character to matters of adequate protection of public health and

safety.

The first of these considerations essentially set the boundary conditions for all technical analyses 

documented in this report. For example, the CPRR alternatives defined in the risk analysis were 

predicated on the established capability of post-core damage containment venting. In two of four 

alternatives selected in the risk analysis (see Chapter 2 for details), additional capability to inject 

water (in the core and/or the containment) was not considered, whereas, in the other two 

alternatives, water injection was probabilistically considered. Moreover, in one of the two latter 

alternatives, an external engineered filter was considered as a further accident management 

feature, and its potential benefit was investigated. The accident sequence analysis was also 

informed by: (1) the lessons learned from the events at the Fukushima Dai-ichi Nuclear Power 

Plant; (2) the accident management alternatives being contemplated by the industry; and (3) the 

current state of knowledge of severe accident progression and mitigation alternatives in a BWR. 

The accident sequence analysis involved development of a core damage event tree (CDET) and 

an accident progression event tree (APET), and binning of a rather large number of possible end 

states to a manageable number of categories with similar outcomes. In accordance with NTTF 

Recommendation 5.1, the evaluation of CPRR alternatives was focused on accidents that are 

initiated by a prolonged station blackout (SBO) event, i.e., an extended loss of alternating current 

power (ELAP) event with loss of all offsite and onsite ac power sources that lasts longer than the 

SBO coping duration specified in 10 CFR 50.63, “Loss of all alternating current power.” The 

human reliability aspect was considered in the accident sequence formulation and despite an 

initial attempt to develop a comprehensive human reliability assessment (HRA), only a bounding 

approach to incorporating HRA into accident sequence analysis was implemented at the end.

The accident sequence analysis results show a low value for core damage frequency (CDF) from 

an ELAP event, and provide insights into which initiating events (e.g. an earthquake), mitigation 

system performance (e.g., RCIC failure), or operator actions (equivalently, human error probability 

associated with such actions) contribute the most to overall CDF for the BWR plants with Mark I 

and Mark II containments. For example, the major contribution to seismically induced ELAP is 

from earthquakes that cause site peak ground accelerations in the range of 0.3 to 0.75g. Also, 

significant contributors to CDF include seismic failures of the batteries, DC switchgear, and the 

emergency diesel generators (EDGs) and their supporting equipment. Failure of the portable 

FLEX pump and failure to start of the reactor core isolation cooling (RCIC) pump is also 

significant. Over a reasonable range of values, CDF is not particularly sensitive to human error 

probabilities for in-control-room and ex-control-room operator actions. 
xviii

The scope of MELCOR analysis covered broadly two categories: (1) reactor systems and 

containment thermal-hydraulics under severe accident conditions, and (2) assessment of source 

terms i.e., timing and magnitude of fission product releases to the environment. The development 

of the MELCOR calculation matrix was based on the CPRR alternatives defined by the accident 

sequence analysis. 

The MELCOR analysis investigated detailed accident progression, source terms, and the 

containment response for representative Mark I and Mark II containment designs following an 

ELAP. The selection of accident scenarios considered for MELCOR analyses is informed by the 

recent state-of-the-art reactor consequence analysis or SOARCA, the Fukushima accident 

reconstruction study, and also by the work documented in SECY-12-0157. The representative 

Mark I containment selected was Peach Bottom Unit 2 like configuration, and the representative 

Mark II was Lasalle like configuration. The calculation matrix for Mark I included sensitivities to: (1) 

mode of venting; (2) status of RPV depressurization; (3) mode of FLEX water injection (to RPV or 

drywell); and (4) water management (i.e., water injection control by throttling flow). The matrix for 

the Mark II analysis included a subset of the Mark I matrix based on the insights from the Mark I 

MELCOR calculations. Additionally, for the Mark II analysis, sensitivities were performed to 

examine the impact of the pedestal and lower cavity designs among the fleet by modifying the 

base model.

The outcome of MELCOR analysis for the first category includes containment temperature and 

pressure signatures and hydrogen distribution in the containment, reactor building, vent line, etc. - 

all indicative of the state of containment vulnerability under severe accident conditions. These 

quantities provide needed information to assess containment integrity and also provide technical 

insights for developing staff guidance for the severe accident capable hardened vent Order EA13-109. The outcome of the MELCOR analysis for the second category includes estimates of

fission products release to the environment.

There was no fundamental shift in the scope and technical approach with regard to MELCOR 

analysis performed in support of the CPRR rulemaking (SECY-15-0085) when compared to what 

was done in SECY-12-0157. The technical approach in both cases considered best estimate 

modeling of accident progression, and incorporated both preventative and mitigative accident 

management measures including venting, water addition and/or water management, as well as 

the option of using engineered filters. However, it is important to recognize that in SECY-12-0157, 

water addition was considered in a general way as the industry’s post-Fukushima severe accident 

management strategies were still evolving and the concept of severe accident water addition 

(SAWA) and severe accident water management (SAWM) had not yet emerged. Moreover, 

industry’s approach to adapt its flexible coping strategy (FLEX), initially developed to meet 10 

CFR 50.54(hh)(2) requirements, was being formulated for severe accident mitigation applications 

at the time. In contrast, during the effort leading to SECY-15-0085, these various concepts and 

severe accident management measures became more mature. The technical analysis 

documented here was informed by this new development, and the analysis resulted in findings 

that are no longer supportive of the recommendation in SECY-12-0157 related to an external 

engineered filter. 

The offsite consequences were calculated using MACCS with site-specific population, economic, 

land use, weather, and evacuation data for a reference Mark I site and a reference Mark II site.

The Peach Bottom Atomic Power Station and the Limerick Generating Station were selected as 

the site-specific reference models for offsite consequence analysis to enable greater modeling 

fidelity for the high population sites (Peach Bottom has the second highest population within a 50 

mile radius among the 15 Mark I sites and Limerick has the highest population within a 50 mile 
xix

radius among the five Mark II sites). Offsite consequence calculations were run for the source 

terms generated by MELCOR corresponding to different CPRR accident management strategies 

following an ELAP event. The results of offsite consequence analysis were used to assess relative 

public health risk reduction associated with various containment protection and release reduction 

measures in terms of a variety of consequence measures including individual early and latent 

fatality risk, population dose, land contamination, economic cost, and displaced population. Like 

the MELCOR analysis, there was no fundamental shift in the scope and technical approach with 

regard to MACCS analysis performed in support of the CPRR rulemaking (SECY-15-0085) when 

compared to what was done in SECY-12-0157. The analysis in both cases used source term 

estimates from MELCOR to calculate with MACCS atmospheric transport and dispersion, 

protective actions, exposures, and resulting offsite consequences. 

The quantitative results from the current MACCS analysis and the analysis included in SECY-12-

0157 are quite similar. However, it is important to recognize that the second top-level 

consideration concerning performance criteria has important implications on the final outcome of 

the technical analysis in SECY-15-0085. Though not explicitly stated in SECY-12-0157, the staff’s 

previous analysis effort to address containment venting and filtration implicitly assumed 

decontamination factor (DF) as a performance criterion. Specifically, a DF value of 1000 

(equivalent to one-tenth of one percent cesium release to the environment – a measure related to 

latent cancer fatality risk and land contamination) was targeted in the previous work, consistent 

with the international nuclear safety practices and guidelines. In contrast, the top-level 

performance criterion used in SECY-15-0085 is QHO – a measure related to early and latent 

fatality risk. The MACCS results for the CPRR alternatives showed that not only is there 

essentially zero individual early fatality risk for all cases analyzed in SECY-15-0085 but also the 

individual latent cancer fatality risk is orders of magnitude lower than the QHO level.

Important findings and key insights from the technical work are delineated below:

• Venting of Mark I and Mark II containments effectively prevents containment

overpressure failure. Pre-core damage anticipatory venting reduces the containment

base pressure at the time of core damage and results in a delay when post core damage

venting is required. Post-core damage containment venting is efficient in purging

hydrogen and other non-condensables from the containment.

• Venting alone, however, is not adequate, as it does not prevent other modes of

containment failure such as liner melt-through and over-temperature failure of the upper

drywell head, bypass of the suppression pool and direct release of radioactivity to the

environment. A combination of venting and water injection is required to prevent such

failures, and the current work provides a sound technical basis to that effect thus

supporting the adequate protection argument.

• Addition of water either into the RPV or the drywell has the following benefits: (1) cooling

of the core debris and containment atmosphere; (2) preventing over-temperature failure

of the upper drywell head; (3) preventing and/or delaying liner melt through in Mark I

containments; (4) maintaining a steam inerted atmosphere which can preclude an

energetic hydrogen combustion; and (5) mitigating radiological releases as it effectively

provides means for fission product scrubbing.

• The environmental releases from Mark II containments are in general comparable to or

lower than those calculated from Mark I containments. Sensitivity analysis performed to

investigate variations in lower cavity configurations of Mark II containments indicate the
xx

environmental releases for all configurations are within the range of releases predicted 

for Mark I containments.

• The major contribution to seismically induced ELAP is from earthquakes with ground

motion exceeding the plant design basis (the safe shutdown earthquake). Specifically,

earthquakes with peak ground accelerations in the range of 0.3 to 0.75g are the major

contributors.

• Significant contributors to CDF include seismic failures of the batteries, DC switchgear,

and the EDGs and their supporting equipment. Failure of the portable FLEX pump and

failure to start of the RCIC pump are also significant contributors. CDF is not particularly

sensitive to human error probabilities for in-control-room and ex-control-room operator

actions.

• The estimated mean individual latent cancer fatality risk (0-10 miles) is more than two

orders of magnitude below the NRC Safety Goal QHO. The risk is low because the coredamage frequency is low and the conditional latent cancer fatality risk is low. The range

of parametric uncertainty in the risk estimates is more than one order of magnitude, and

is largely driven by uncertainty in the seismic hazard curves.

• The estimated individual early fatality risk is essentially zero in all cases and for all

alternatives considered, consistent with the findings previously in the State-of-the-Art

Reactor Consequence Analyses (SOARCA) and spent fuel pool (SFP) consequence

studies. This risk remains unchanged for a wide range of sensitivity analysis.

• The release to the environment is delayed long enough after the accident initiation to

allow time for the emergency planning zone (EPZ) population to evacuate. Therefore the

ILCF risk is dominated by long-term phase exposures to slightly contaminated areas

(under the 500 mrem per year habitability criterion threshold). A larger release may

displace more people for more time, and therefore incur a larger societal cost, but the

health risk to the public, measured in ILCF risk, is effectively capped by the habitability

criterion.

In general, larger releases of radioactive materials to the environment displace more people for 

more time, and therefore incur larger societal costs. However, for a larger release, the cancer 

fatality risk to the public shows a nonlinear response because protective actions are in place 

primarily to reduce exposures (habitability criterion) at the tradeoff of other societal costs such as 

land contamination, displaced population, and economic losses. Releases that span a longer 

duration were often seen to result in higher societal costs because longer durations allow more 

time for the wind to shift direction and thus spread plumes in more directions.

• The potential effectiveness of an external filter on reducing the environmental release is

heavily influenced by release pathways. For accident cases in which the entire release flows

through a vent pathway, the external filter can reduce the environmental release

substantially. For accident cases resulting in a bypass, there is less benefit from an external

filter because some of the release may bypass the venting system. Generally, while an

engineered filter might accrue additional incremental benefits in terms of further reducing the

long-term public health risk, it is not warranted for adequate protection as significant margins

exist between estimated plant risks and the NRC established safety goals.
xxi

• Sensitivity calculations do not change any of the consequence analysis insights related to

the QHO metrics. Individual early fatality risk remains essentially zero for all sensitivity

calculations performed, and ILCF risk remains well below the QHO, even in calculations

assuming a larger habitability criterion (e.g., 2 rem per year instead of 500 mrem per year).

In summary, the work documented in this NUREG report provides the technical basis to address 

containment venting and filtration issue (NTTF 5.1) for BWRs with Mark I and Mark II 

containments. The results support the overall conclusion in SECY-15-0085 that no additional 

regulatory action beyond the implementation of the severe accident capable vent order EA-13-109 

is required based on adequate protection. The analysis confirms that significant margins exist 

between estimated plant risks that might be influenced by improvements in containment 

performance and the NRC established safety goals. However, these margins may be eroded 

somewhat when considering other accident scenarios and/or precursor events in any reasonable 

combination, and will likely retain the public health risks to an acceptable value from the adequate 

protection standpoint. That said, the NRC will continue to assess information emerging from 

ongoing international research activities on containment performance, and will continue to engage 

in long-term activities to enhance safety under established research programs. 

Coincident with the analyses by the NRC staff and industry related to the CPRR rulemaking and 

related Orders, licensees were developing revisions to the severe accident managmenet 

guidelines (SAMGs) to address lessons learned from the Fukushima accident. The analyses 

performed to address issues related to containment venting and severe accident water addition 

provided valuable insights and supported actual revisions to the SAMGs for plants with Mark I and 

Mark II containments. This incorporation of insights from the modelling of beyond-design-basis 

events and severe accidents into plant guidance documents provides a useful example of the 

potential benefits of the efforts by the NRC and industry to develop improved analytical 

capabilities. Another example is the use of the results and technical insights in formulating the 

regulatory basis for the mitigation of beyond design basis events rulemaking. This rulemaking, 

though not imposing any regulatory footprint on SAMG, nevertheless directs the NRC staff to 

provide periodic oversight to industry’s SAMG implementation through NRC’s updated Reactor 

Oversight Program (ROP). A third and related example is that of a recent initiative in many 

Organization for Economic Cooperation and Development (OECD) countries and a collective 

effort in the OECD program of work to develop technical insights on how the type of beyond 

design basis analysis work documented in this report can inform SAMG in a positive way and in 

so doing, enhance the safety of nuclear power plants.

xxiii

ACKNOWLEDGMENTS

The work documented in the report was done by the staff of the Office of Nuclear Regulatory 

Research (RES) in response to a Commission direction to address containment venting and 

filtration for boiling water reactor (BWR) Mark I and Mark II containments. The initiative to address 

the issue was led by the Office of Nuclear Reactor Regulation (NRR) Japan Lessons Learned 

Directorate (JLD). RES interacted closely with the JLD staff, as well as staff from technical 

divisions of NRR and also staff from other NRC program offices in this effort. The authors wish to 

acknowledge valuable suggestions and advice from these offices. In particular, the authors wish 

to acknowledge the contribution of Rajender Auluck, Robert Beall, and William Reckley of JLD 

and also Aaron Szabo (former NRR staff) in this respect.

The technical work was a collaborative effort between the Division of Risk Analysis (DRA) and the 

Division of Systems Analysis (DSA) within RES. Besides the authors, other staff from these two 

divisions contributed to this collaborative effort in various ways, most notably by providing expert 

advice and guidance to the authors and by conducting a peer review of the product. The authors 

wish to acknowledge the contributions of Edward Fuller, Tina Ghosh, John Lane, Allen 

Notafrancesco, and Nathan Siu in this regard. The authors also acknowledge William Reckley 

from NRR who reviewed a draft version of the report. The authors further wish to acknowledge 

James Chang from DRA who provided insights into probabilistic risk assessment by performing 

scoping human reliability analysis, and Jesse Phillips from Sandia National Laboratories (SNL) 

who performed accident analysis of BWR Mark II containments. The authors also wish to 

acknowledge Nathan Bixler of SNL and Joseph Jones, formerly of SNL, for their support and peer 

review of the offsite consequence analysis and for their affirmation that it is consistent with 

SOARCA best practices modeling approaches. Finally, the authors acknowledge earlier 

contribution of the SNL staff in developing MELCOR input models for BWR Mark I and Mark II 

containments.

During the conduct of the technical work and preparation of the report, the authors benefitted 

greatly from interactions with the ACRS members and other external stakeholders. Their collective 

contribution in shaping the scope and nature of the report is gratefully acknowledged.

xxv

ABBREVIATIONS AND ACRONYMS

AC alternating current

ACRS Advisory Committee on Reactor Safeguards

ANS American Nuclear Society

APET accident progression event tree

ARTIST Aerosol Trapping in a Steam Generator

ATD atmospheric transport and dispersion

atm atmosphere

AV anticipatory venting

BEIR Biological Effects of Ionizing Radiation

BWR boiling water reactor

BWROG Boiling Water Reactor Owners’ Group 

CCDP conditional core damage probability

CDET core-damage event tree

CDF core damage frequency

CPIP Containment Performance Improvement Program

CPRR containment protection and release reduction

CRAC Calculation of Reactor Accident Consequences

CRF consequence reduction factor

CST condensate storage tank

DC direct current

DCF dose conversion factor

DF decontamination factor

DW drywell

DWF drywell first strategy

DW LMT drywell liner melt through

EAS emergency alert system

ECCS emergency core cooling systems

EDG emergency diesel generator

EDMG extended damage mitigation guideline

ELAP extended loss of alternating current power

EOP emergency operating procedures

EPA Environmental Protection Agency

EPRI Electric Power Research Institute

EPS emergency power system

EPZ emergency planning zone
xxvi

ETE evacuation time estimate

FCVS filtered containment venting system

FGR Federal Guidance Report

FLEX flexible coping strategy

FP flow path

FPT fission product tests

FV Fussell-Vesely

gpm gallons per minute

Gy gray

HEPs human error probabilities

HFE human failure events

HP high pressure

HPCI high pressure cooling injection system

HPS Health Physics Society

HRA human reliability assessment

HTC heat transfer coefficient

ILCF individual latent cancer fatality

IPE individual plant examination

IPEEE individual plant examination of external events

JLD Japanese lessons learned directorate

LCF latent cancer fatality

LMT liner melt through

LNT linear no threshold

LOOP loss of offsite power events

LP low pressure

LTSBO long-term station blackout

LWR light water reactor

m meter

MAAP Modular Accident Analysis Program

MACCS MELCOR Accident Consequence Code System

MCCI molten core concrete interactions

MDBDE mitigation of beyond-design-basis event

MP medium pressure

mph miles per hour
xxvii

MSIV main steam isolation valve

MSL main steam line

MSLCR main steam line creep rupture

mSv millisievert

NCG non-condensable gas

NRC Nuclear Regulatory Commission

NRR Office of Nuclear Reactor Regulation

NSRC National SAFER Response Centers

NTTF Near Term Task Force

OECD Organization for Economic Cooperation and Development

OLO open at PCPL and leave open

ORO offsite response organization

PAG protective action guide

PAR protective action recommendation

PCPL primary containment pressure limit

PDS plant damage state

PEMA Pennsylvania Emergency Management Agency

PRA probabilistic risk assessment

PRT pressure relief tank

PWR pressurized water reactor

QHO qualitative health objective

RAW risk achievement worth

RCIC reactor core isolation cooling

RES Office of Nuclear Regulatory Research

ROP reactor oversight program

RPV reactor pressure vessel

SAE Site Area Emergency

SAGs severe accident guidelines

SAMG severe accident management guidelines

SAWA severe accident water addition

SAWM severe accident water management

SBO station blackout

scfh standard cubic feet per hour

SFP spent fuel pool

SG steam generator
xxviii

SGTR steam generator tube rupture

SNL Sandia National Laboratories

SOARCA State of the Art Reactor Consequence Analyses

SPAR standardized plant analysis risk

SRM staff requirements memorandum

SRV safety relief valve

SV safety valve

TAF top of active fuel

UA uncertainty analysis

URC MINALT upper reactor cavity ablation depth reaching minimum altitude of the 

concrete

VC vent cycling

WWF wetwell first strategy
1-1

1 INTRODUCTION

This report documents the results of research conducted by the staff of the Office of Nuclear 

Regulatory Research (RES) to support the agency initiative to address the containment venting 

issue for the boiling water reactor (BWR) Mark I and Mark II containments. The venting issue was 

identified by the Near Term Task Force (NTTF), put together in the aftermath of the Fukushima 

Dai-ichi nuclear power plant accident in Japan, for all containment types, but prioritized as a shortterm action item for the two containment types mentioned above. The initiative to address the 

venting issue was led by the Office of Nuclear Reactor Regulation (NRR) Japan Lessons Learned 

Directorate (JLD).

The RES work focused on developing the technical basis for a potential rulemaking action on 

containment protection and release reduction (CPRR), previously also known as filtered 

containment venting system (FCVS). The RES work covered three areas of analyses: (1) accident 

sequence analysis (event tree development) to identify accident sequences initiated by extended 

loss of ac power (ELAP) due to internal events and seismic events deemed to be the most 

significant risk contributors; (2) accident analysis of these sequences and assessment of 

radiological source terms; and (3) analysis of consequences with particular emphasis on health 

effects, both short term and long term. The results of consequence analysis were used to assess 

relative public health risks (more appropriately, health risk reduction) associated with various 

containment protection and release reduction measures. The relative risk measures were used by 

NRR to perform regulatory analysis for various mitigation options considered.

1.1 Containment Venting Issue – A Historical Perspective

The containment venting issue is not new and has a long history behind it. The U.S. Nuclear 

Regulatory Commission (NRC) and nuclear industry have recognized the potential need to vent 

Mark I and Mark II containment designs to cope with severe accident conditions since at least the 

early 1980s. These containment designs as well as other pressure suppression containments 

have been shown to be capable of addressing the requirements related to the design-basis 

accidents. However, various studies (e.g., NUREG-1150, “Severe Accident Risks: An Assessment 

for Five U.S. Nuclear Power Plants” [1]) have shown that the Mark I and Mark II containments do 

not have the same margins of safety that other containments (e.g., large dry ones) have during 

accidents that exceed the conditions established by design basis events. In 1983, the NRC 

approved Revision 2 to the Boiling Water Reactor Owners’ Group (BWROG) Emergency 

Procedure Guidelines (EPG), which included guidance for operators to vent Mark I and Mark II 

containments in response to containment overpressure conditions. Though emergency 

procedures have existed since the 1980s for Mark I and Mark II containment venting systems for 

beyond-design-basis accidents, the NRC’s actions to date for operating reactors have not 

required containment venting systems for Mark I and Mark II containments be designed for severe 

accident conditions. 

The NRC evaluated the possible imposition of additional containment functional requirements for 

operating reactors in other previous studies as well. Subsequent to the Three Mile Island Unit 2 

nuclear plant core melt event in 1979, NUREG-0585, “TMI-2 Lessons Learned Task Force Final 

Report,” October 1979, [2] stated: 

Available studies indicate that controlled venting of the containment to prevent failure due 

to overpressure could be an effective means of delaying ultimate containment failure by 

melting through. If appropriately filtered to partially decontaminate the gases that would be 
1-2

released in order to avoid overpressurization, such venting may significantly reduce the 

consequences and risk from core-melt accidents.

A controlled (and potentially filtered) release was identified as a favorable alternative to 

catastrophic failure of the containment. 

In SECY-88-147, “Integration Plan for Closure of Severe Accident Issues,” dated May 25, 

1988, [3] the U.S. Nuclear Regulatory Commission (NRC) staff presented to the Commission its 

plan to evaluate potential generic severe accident containment vulnerabilities in a research effort 

entitled the “Containment Performance Improvement Program” (CPIP). This effort was predicated 

on the presumption that there are generic severe accident challenges to each light water reactor 

(LWR) containment type that should be assessed to determine whether additional regulatory 

guidance or requirements concerning needed containment features were warranted, and to 

confirm the adequacy of the existing Commission policy. These assessments were needed 

because of the uncertainty in the ability of LWR containments to successfully survive some severe 

accident challenges, as indicated by the results documented in NUREG-1150. 

All LWR containment types were considered in the CPI program, beginning with the BWR Mark I 

containments. The potential improvements for Mark I containments were documented in 

NUREG/CR-5225 (including Addendum 1), “An Overview of BWR Mark-I Containment Venting 

Risk Implications,” [4] and SECY-89-017, “Mark I Containment Performance Improvement 

Program,” dated January 23, 1989 [5]. In the latter document, the staff described its findings 

associated with six areas of potential improvement for Mark I containments. These were: (1) 

hydrogen control, (2) alternate water supply for reactor vessel injection and containment drywell 

sprays, (3) containment pressure relief capability (venting), (4) enhanced reactor pressure vessel 

(RPV) depressurization system reliability, (5) core debris control, and (6) emergency procedures 

and training. Each area was evaluated to determine the potential benefits in terms of reducing the 

core melt frequency, containment failure probability, and offsite consequences. The staff provided 

cost-justification for, and recommended implementation of, all the aforementioned improvements

with two exceptions: hydrogen control (beyond then the existing rule) and core debris control (i.e., 

feasibility of confining core debris through design of curbs in the drywell and curbs or weir walls in 

the torus room below the wetwell).

In the subsequent staff requirements memorandum (SRM), however, the Commission concluded 

that the majority of the staff’s recommended safety improvements would be evaluated by 

licensees as part of the Individual Plant Examination (IPE) Program [6]. The only exception was 

the hardened vent capability recommendation. The Commission directed the staff to approve 

installation of hardened vents under the provisions of 10 CFR 50.59, “Changes, Tests, and 

Experiments,” [7] for licensees that would voluntarily implement this improvement and perform a 

back-fit analysis for requiring a hard vent installation at those plants who declined voluntary 

installation. Thus, NRC issued Generic Letter 89-16, “Installation of a Hardened Wetwell Vent,” [8] 

to all licensees of BWRs with Mark I containments to encourage licensees to voluntarily install a 

hardened wetwell vent in September 1989 providing an example of an acceptable design that 

used the suppression pool to achieve as much reduction in effluent radioactivity as possible. 

In response to the issuance of the generic letter, all Mark I licensees installed a version of a 

hardened vent under 10 CFR 50.59. Some licensees also installed a hardened vent branch line 

from the drywell. The Boiling Water Reactor Owners’ Group (BWROG) developed a general 

design criteria document that was subsequently approved by the staff (with clarifications). The 

hardened vent was specifically to provide an exhaust line from the wetwell air space to a suitable 

release point (e.g., stack, reactor building or turbine building roof). The basic design objective of 
1-3

the hardened vent was to mitigate the loss of decay heat removal accident sequence, and not for 

operation during a severe accident. Because the modifications to the plant were performed in 

accordance with the Code of Federal Regulations, 10 CFR 50.59, “Changes, tests and 

experiments,” detailed information regarding individual plant configurations was not submitted to 

the NRC staff for review.

In concluding the CPIP effort, the NRC determined that the low probability of severe accidents 

resulted in the costs of plant modifications beyond the installation of a hardened vent exceeded 

the calculated benefits and, as such, were not cost-justified for Mark I and Mark II containment 

designs. Legislators and regulators in other countries did impose additional requirements in the 

aftermath of the accidents at Three Mile Island and Chernobyl. In effect, those other regulatory 

authorities assessed filtered vents and other severe accident management strategies with an 

emphasis on the defense-in-depth argument and with less or no consideration of cost/benefit 

analyses. 

The potential improvements for Mark II containments were published in NUREG/CR-5528, “An 

Assessment of BWR Mark-II Containment Challenges, Failure Modes, and Potential 

Improvements.” [9] Mark II containment vulnerabilities and potential improvements identified in 

this document were largely the same as those for Mark I containments. However, less definitive 

conclusions were reached regarding venting of Mark II containments for a number of reasons.

The findings and recommendations for Mark II containments as well as other containment types 

were documented in SECY-90-120, “Recommendations of Containment Performance 

Improvement Program for Plants with Mark II, Mark III, Ice Condenser, and Dry Containment." [10] 

However, unlike for Mark I containments, no generic letter requiring containment improvement 

was issued for these other containment types.

1.2 Post-Fukushima Development

The events at the Fukushima Dai-ichi nuclear power plant brought the containment venting issue 

once again to the forefront. The accidents involved an extended loss of electrical power and heatremoval systems, resulting in containment pressures that exceeded the containment design 

pressure. Plant conditions at Fukushima Dai-ichi (e.g., loss of all electrical power or station 

blackout) hampered the efforts of operators to address the containment overpressure conditions 

using the installed venting systems, which ultimately contributed to the compromise of all fission 

product barriers and significant releases of radioactive material. The events highlighted the need 

for safety improvements for nuclear power plants related to beyond-design-basis natural hazards, 

and the resulting effects on plant systems and barriers from an extended loss of electrical power 

and access to heat removal systems. 

In SECY-11-0137, “Prioritization of Recommended Actions To Be Taken in Response to 

Fukushima Lessons Learned,” dated October 3, 2011 [11], the NRC staff described its proposals 

for the regulatory actions to address containment venting related NTTF recommendations [12]. 

Venting containment can help prevent or delay the loss of, or facilitate recovery of, important 

safety functions such as reactor core cooling, reactor coolant inventory control, containment 

cooling, and containment pressure control. The insights gained from Fukushima Dai-ichi led the 

agency to impose additional requirements for reliable hardened venting systems for plants with 

Mark I and Mark II containments. As such, on March 11, 2012, the NRC issued an order (EA-12-

050) [13] to all licensees of BWR facilities with Mark I and Mark II containment designs to require

a reliable hardened vent.
1-4

The EA-12-050 order provided requirements to ensure reliable operation of the hardened venting 

system in support of strategies relating to the prevention of core damage. However, EA-12-050 

did not include requirements for reliable operation under severe accident conditions; rather, it 

focused on requirements prior to the onset of core damage. As such, EA-12-050 did not prescribe 

the venting location (drywell or wetwell) as essentially all vent flow prior to RPV breach would 

pass through the suppression pool. Nevertheless, the existing emergency operating procedures 

(EOPs), severe accident management guidelines (SAMGs), and extended damage mitigation 

guidelines (EDMGs) for BWRs with Mark I and Mark II containments contain provisions for venting 

containment following core damage.

One of the six additional recommendations identified in SECY-11-0137, and further developed in 

SECY-12-0025 [14], was consideration of additional performance requirements, including filters 

for hardened containment vent systems for Mark I and Mark II containment designs. This would

improve reliability during severe accident conditions and limit the release of radioactive materials if

the venting systems were used after significant core damage had occurred. The NRC staff 

identified an additional issue in SECY-11-0137 related to possible modifications of the 

containment vents, including the addition of engineered filters. In the SRM for SECY-11-0137, 

dated December 15, 2011 [15], the Commission directed the NRC staff as follows:

The staff should quickly shift the issue of “Filtration of Containment Vents” from the 

“additional issues” category and merge it with the Tier 1 issue of hardened vents for Mark I 

and Mark II containments such that the analysis and interaction with stakeholders needed 

to inform a decision on whether filtered vents should be required can be performed 

concurrently with the development of the technical bases, acceptance criteria, and design 

expectations for reliable hardened vents.

In accordance with the direction in SRM-SECY-11-0137, the additional issue of filtration of 

containment vents was merged with the Tier 1 issue of hardened vents for Mark I and Mark II 

containments to facilitate further analysis and interaction with stakeholders so as to inform the 

need and benefit of filtered vents. In SECY-12-0025, the staff explained that it needed to resolve 

technical and policy issues before regulatory action could be proposed that would require 

licensees to install filters, or change any other performance requirement, for hardened 

containment vent systems. 

To support additional regulatory actions related to the performance of Mark I and Mark II 

containments during severe accidents, RES performed a systematic analysis of accident source 

terms and consequences in late 2011 and mid 2012. This analysis was conducted for a 

representative BWR plant with Mark I containment and for representative accident scenarios, 

predicated on an extended loss of alternating current power (ELAP) as in Fukushima. The 

analysis used NRC severe accident code MELCOR and the consequence code MACCS 

(MELCOR Accident Consequence Code System) for source term and consequence 

assessments, respectively, and was informed by lessons-learned and best practices from the 

State-of-the-Art Reactor Consequence Analysis (SOARCA) project [16]. The analysis provided an 

assessment of the sensitivity of the plant risks to selected accident management strategies, 

keeping in mind that such strategies were still evolving at the time. The MELCOR and MACCS 

simulations were used along with insights from previous studies (e.g., individual plant 

examinations, NUREG-1150, CPIP, severe accident mitigation alternatives) to evaluate the 

potential benefits of features in Mark I and Mark II containment designs. These designs included 

containment venting systems with and without engineered filters, and provision for water addition 

into the pressure vessel and/or the containment. 
1-5

The technical analysis, documented in SECY-12-0157 [17], concluded that the plants with Mark I 

containments (and by extrapolation Mark II containments as well) would benefit from a 

containment venting and water addition strategy for a vast majority of severe accident sequences 

(whether the vent includes an engineered filter or not). In addition to its own assessments and 

analyses, the staff relied on information gained through interactions with various stakeholders

including the nuclear industry, which provided insights to the NRC staff during several public 

meetings, and also through a report the Electric Power Research Institute published. 

From a regulatory perspective, the staff presented four options in SECY-12-0157 for Commission 

consideration to address the containment venting issue. The options presented included: (1) 

maintaining the requirements established in Order EA-12-50 for reliable hardened vents and do 

nothing else; (2) issuing a new order requiring containment venting systems to be capable of 

operating under severe accident conditions; (3) issuing an order requiring containment venting 

systems capable of operating under severe accident conditions with additional external filtering 

feature to reduce release of radioactivity to environment through controlled release pathways; and 

(4) developing a performance-based severe accident management strategy for BWRs with Mark I

and Mark II containments. The staff’s regulatory analysis focused on Option 2 (severe accident

capable vent order) and Option 3 (filtered vent order) as those options involved potential near

term regulatory action, and recommended that the Commission approve Option 3. This

recommendation was not based solely on quantitative analysis; rather, a combination of

quantitative analysis and qualitative arguments invoking the long established policy statements by

the Commission on severe accident, defense-in-depth, and other related topics.

The Commission directed the staff in SRM-SECY-12-0157 [18] to: (1) issue a modification to 

Order EA-12-050 requiring licensees with Mark I and Mark II containments to ”upgrade or replace 

the reliable hardened vents required by Order EA-12-050 with a containment venting system 

designed and installed to remain functional during severe accident conditions,” and (2) develop 

technical basis and rulemaking for filtering strategies with drywell filtration and severe accident 

management of BWR Mark I and II containments. The NRC staff subsequently issued Order EA13-109, “Issuance of Order Modifying Licenses with Regard to Reliable Hardened Containment

Vents Capable of Operation Under Severe Accident Conditions” (ADAMS Accession No.

ML13143A321) [19], to ensure that vents on BWR Mark I and II containments will remain

functional in the conditions following a reactor core melt accident.

In parallel, the staff initiated additional technical work to develop technical basis for potential 

rulemaking involving filtering strategies with drywell filtration and severe accident management of 

BWR Mark I and II containments. As mentioned previously, this additional technical work by RES 

focused on developing the technical basis for a potential rulemaking action on containment 

protection and release reduction (CPRR) supporting SECY-15-0085, “Evaluation of the 

Containment Protection & Release Reduction for Mark I and Mark II Boiling Water Reactors 

Rulemaking Activities (10 CFR Part 50) (ADAMS Accession No. ML15005A1079) [20]. The 

technical work covered three areas of analyses: (1) accident sequence analysis (event tree 

development) to identify accident sequences initiated by extended loss of ac power (ELAP) due to 

internal events and seismic events deemed to be the most significant risk contributors; (2) accident 

analysis of these sequences and assessment of radiological source terms; and (3) analysis of 

consequences with particular emphasis on health effects, both short term and long term. 
1-6

1.3 Content and Structure of Report

The report documents the outcome of the technical work to support regulatory basis of the 

potential CPRR rulemaking activity. The regulatory analysis for rulemaking activity is documented 

in an enclosure to SECY-15-0085. 

Chapter 2 of the report documents the staff’s accident sequence analysis that was used to select 

risk-significant accident sequences. This work involved development of a core damage event tree 

(CDET) and an accident progression event tree (APET), and binning of a rather large number of 

possible end states to a manageable number of categories with similar outcomes. Another 

important aspect of the accident sequence analysis is the development of CPRR alternatives 

considered in the regulatory basis. In accordance with NTTF Recommendation 5.1, the evaluation 

of CPRR alternatives was focused on accidents that are initiated by a prolonged station blackout 

(SBO) event, i.e., an extended loss of alternating current power (ELAP) event with loss of all 

offsite and onsite ac power sources that lasts longer than the SBO coping duration specified in 10 

CFR 50.63, “Loss of all alternating current power.” [21]

Chapter 3 describes in detail the MELCOR analysis, the scope of which falls broadly into two 

categories: (1) reactor systems and containment thermal-hydraulics under severe accident 

conditions, and (2) assessment of source terms (i.e., timing and magnitude of fission product 

releases to the environment). The development of the MELCOR calculation matrix was tied to the 

regulatory basis alternatives mentioned in the previous paragraph. 

The outcome of MELCOR analysis for the first category includes containment temperature and

pressure signatures and hydrogen distribution in the containment, reactor building, vent line, etc., 

all indicative of the state of containment vulnerability under severe accident conditions. These 

quantities provided needed information to assess containment integrity and also provided the 

technical basis for developing staff guidance, for example, the severe accident capable hardened 

vent Order EA-13-109. The outcome of the MELCOR analysis for the second category is 

environmental source term release estimates, which were used to calculate offsite consequences.

Chapter 4 describes in detail the consequence analysis using MELCOR Accident Consequence 

Code System (MACCS). The code calculates offsite consequences for the source terms 

generated by MELCOR corresponding to different CPRR accident management strategies 

following an ELAP event. The results of consequence analysis were used to assess relative public 

health risks (more appropriately, health risk reduction) associated with various containment 

protection and release reduction measures. The staff evaluated the conditional offsite 

consequences associated with the CPRR alternatives discussed in the accident sequence 

analysis and the results are described in this chapter.

Chapter 5 provides the main results of the risk integration, i.e., risk estimates corresponding to 

regulatory analysis alternatives and sub-alternatives considered in the study and described in 

Chapter 2. The integration process involved combining the outcome of MELCOR analysis 

(Chapter 3) and MACCS analysis (Chapter 4). 

Key insights from the technical work are summarized in Chapter 6 in three major areas: accident 

sequence analysis, source term assessment using MELCOR, and consequence assessment 

using MACCS. It is important to note that the accident sequence analysis results show a low core 

damage frequency from an ELAP event for BWR plants with Mark I and Mark II containments, and 

provide insights into the relative contributions of various factors (e.g., external hazards, equipment 

failures, human errors, etc.) to overall core damage frequency. The MELCOR results show both
1-7

water addition and venting are required to maintain containment integrity and reduce source 

terms, and finally, MACCS results show essentially zero early fatality risk and sufficiently low 

individual latent cancer fatality (ILCF) risk with two order of magnitude or more safety margin 

relative to quantitative health objective (QHO). 

As with any complex analysis work involving multiple technical disciplines, integration of the work 

requires a logical structure. To that end, the CPRR technical basis work adopted a technical 

approach that consists of the following steps: 

• Accident sequence analysis that includes quantitative risk estimates and qualitative risk

insights for various CPRR strategies

• Development of an accident progression calculation matrix, consistent with the risk

evaluation, for analyzing the range of different accident management strategies

• Accident progression and source term calculations using MELCOR

• Offsite consequence calculations using MACCS for selected MELCOR source terms

• Integration of the results of the steps above to generate frequency-weighted offsite

consequences corresponding to each of the different CPRR strategies

The technical work in each of the three major disciplines was based on a number of assumptions 

or considerations, which are delineated in detail in the respective chapters describing the work. It 

is, however, important to highlight some top-level considerations. These are:

• SRM-SECY-12-0157 direction that the regulatory basis should assume the benefits of

severe accident capable hardened venting systems (EA-13-109) that would accrue

equally to engineered filters and to filtration strategies

• Consideration of a variety of performance criteria such as a decontamination factor,

equipment and procedure availability similar to those required to implement 10 CFR

50.54(hh) [22], or other measures that may be developed during the stakeholder

engagement

• Consideration of requirements associated with measures to enhance the capability to

maintain containment integrity and to cool core debris

The first of these considerations essentially set the boundary conditions for all technical analyses.

For example, the CPRR alternatives defined in the risk analysis were predicated on the 

established capability of post-core damage containment venting. In two of four alternatives 

selected in the risk analysis (see Chapter 2 for details), additional capability to inject water (in the 

core and/or the containment) was not considered whereas, in the other two alternatives, water 

injection was probabilistically considered. Moreover, in one of the two latter alternatives, an 

external engineered filter was considered as a further accident management feature, and its 

potential benefit was investigated. To that end, the MELCOR matrix for source term calculations 

was mapped to these four CPRR alternatives, and subsequent MACCS consequence calculations 

were mapped to these alternatives as well.
1-8

The second consideration concerning performance criteria has important implications on the final 

outcome of the technical analysis. Though not explicitly stated in SECY-12-0157, the staff’s 

previous analysis effort to address the containment venting issue implicitly assumed 

decontamination factor (DF) as a performance criterion. Specifically, a DF value of 1000 

(equivalent to one-tenth of one percent cesium release to the environment – a measure related to 

latent cancer fatality risk and land contamination) was targeted in the previous work, consistent 

with the international nuclear safety practices and guidelines. In contrast, the top-level 

performance criterion used in SECY-15-0085 is QHO – a measure exclusively related to public 

health risk. As will be seen later in the document, for all accident scenarios considered in the 

analysis, the risk was assessed to be well below the QHO limit. This obviated the need for 

otherwise meeting a more stringent target of a DF value of 1000. 

The third consideration associated with measures to enhance the capability to maintain 

containment integrity and to cool core debris, likewise, has important implications. In SECY-12-

0157, water addition was considered, albeit in a general way, as the industry’s post-Fukushima 

severe accident management strategies were still evolving and the concept of severe accident 

water addition (SAWA) and severe accident water management (SAWM) did not yet emerge.

Moreover, industry’s approach to adapt its flexible coping strategy (FLEX), initially developed to 

meet 10 CFR 50.54(hh) requirements, was being formulated for severe accident mitigation 

application at the time. In contrast, during the effort leading to SECY-15-0085, these various 

concepts and severe accident management measures became more mature. The technical 

analysis documented here was informed by this new development, and the analysis resulted in 

findings that are far less supportive of the recommendation in SECY-12-0157 related to an 

external engineered filter. It was recognized that while an engineered filter might accrue additional 

incremental benefits in terms of further reducing the long-term public health risk, its 

implementation could not be justified as a cost-beneficial safety enhancement measure. 

A few other high level assumptions, specific to each of the three areas of analysis (accident 

sequence analysis, MELCOR, and MACCS) are worth noting here. The accident sequence analysis 

was informed by: (1) the lessons learned from the events at the Fukushima Dai-ichi Nuclear Power 

Plant; (2) the accident management alternatives being contemplated by the industry; (3) the current 

state of knowledge of severe accident progression and mitigation alternatives in a BWR; and (4) the 

experience gained from the previous effort documented in SECY-12-0157. The human reliability 

aspect was considered in the event tree formulation and despite an initial attempt to develop a 

comprehensive human reliability assessment (HRA), only a bounding approach to incorporating 

HRA into the accident sequences was implemented at the end. 

For source term assessment, the version of MELCOR used is consistent with other recent 

MELCOR applications such as the State-of-the-Art Reactor Consequence Analysis (SOARCA) 

and the spent fuel pool (SFP) study [23]. MELCOR embodies the current state of knowledge of 

severe accident progression and mitigation; however, as with any other complex analysis tools, 

the models in MELCOR are based, in part, on phenomenological studies and, to a degree, on 

physical abstraction. As such, modeling and parametric uncertainties in MELCOR are recognized 

and addressed through uncertainty analysis and/or sensitivity studies.

For consequence analysis, likewise, the version of MACCS used is consistent with other 

applications such as SOARCA and SFP. None of these applications put economic consequence 

on the same footing as the public health consequence, consistent with a recent commission 

deliberation on the subject (see SRM-SECY-12-0110) [24]. On a specific technical note, aqueous 

release paths are currently not modeled in consequence analysis. 
1-9

1.4 References for Chapter 1

1. U.S. Nuclear Regulatory Commission, “Severe Accident Risks: An Assessment for Five

U.S. Nuclear Power Plants,” NUREG-1150, December 1990.

2. U.S. Nuclear Regulatory Commission, “TMI-2 Lessons Learned Task Force Final Report,”

NUREG-0585, October 1979.

3. U.S. Nuclear Regulatory Commission, “Integration Plan for Closure of Severe Accident

Issues,” SECY-88-147, May 25, 1988.

4. U.S. Nuclear Regulatory Commission, “An Overview of BWR Mark-I Containment Venting

Risk Implications,” NUREG/CR-5225.

5. U.S. Nuclear Regulatory Commission, “Mark I Containment Performance Improvement

Program,” SECY-89-017, January 23, 1989.

6. U.S. Nuclear Regulatory Commission, “Individual Plant Examination Program:

Perspectives on Reactor Safety and Plant Performance,” NUREG-1560, November 1996.

7. Code of Federal Regulations, Title 10, Part 50, “Changes, Tests, and Experiments,” 10

CFR 50.59.

8. U.S. Nuclear Regulatory Commission, “Installation of a Hardened Wetwell Vent,” Generic

Letter 89-16, September 1989.

9. Kelly, D.L., et al., “An Assessment of BWR Mark-II Containment Challenges, Failure

Modes, and Potential Improvements in Performance,” NUREG/CR-5528, Nuclear

Regulatory Commission, Washington DC, July 1990.

10. U.S. Nuclear Regulatory Commission, “Recommendations for Containment Performance

Improvement Program for Plants with Mark II, Mark III, Ice Condenser, and Dry

Containment,” SECY-90-120, 1990.

11. U.S. Nuclear Regulatory Commission, “Prioritization of Recommended Actions to be

Taken in Response to Fukushima Lessons Learned,” SECY-11-0137, October 3, 2011.

12. U.S. Nuclear Regulatory Commission, “Recommendations for Enhancing Reactor Safety

in the 21st Century: The Near-Term Task Force Review of Insights from the Fukushima

Dai-ichi Accident,” July 11, 2011, ADAMS Accession No. ML111861807.

13. U.S. Nuclear Regulatory Commission, “Issuance of Order to Modify Licenses with Regard

to Reliable Hardened Containment Vents,” Order EA-12-050, March 12, 2012, ADAMS

Accession No. ML12054A694.

14. U.S. Nuclear Regulatory Commission, “Proposed Orders and Requests for Information in

Response to Lessons Learned from Japan's March 11, 2011, Great Tohoku Earthquake

and Tsunami,” SECY-12-0025, February 17, 2012.
1-10

15. U.S. Nuclear Regulatory Commission, “Prioritization of Recommended Actions to be

Taken in Response to Fukushima Lessons Learned,” SRM-SECY-11-0137, December 15,

2011.

16. Bixler N., et al. “State-of-the-Art Reactor Consequence Analyses (SOARCA) Project

Volume 1: Peach Bottom Integrated Analysis,” NUREG/CR-7110 Vol. 1, Rev. 1, Nuclear

Regulatory Commission, Washington DC, May 2013.

17. U.S. Nuclear Regulatory Commission, “Consideration of Additional Requirements for

Containment Venting Systems for Boiling Water Reactors with Mark I and Mark II

Containments (REDACTED VERSION),” SECY-12-0157, November 26, 2012, ADAMS

Accession No. ML12345A030.

18. U.S. Nuclear Regulatory Commission, “Consideration Of Additional Requirements For

Containment Venting Systems For Boiling Water Reactors With Mark I and Mark II

Containments,” SRM-SECY-12-0157, March 2013.

19. U.S. Nuclear Regulatory Commission, “Issuance of Order to Modify Licenses with Regard

to Reliable Hardened Containment Vents Capable of Operation Under Severe Accident

Conditions,” Order EA-13-109, June 6, 2013, ADAMS Accession No. ML13143A334.

20. U.S. Nuclear Regulatory Commission, “Evaluation of the Containment Protection &

Release Reduction for Mark I and Mark II Boiling Water Reactors Rulemaking Activities

(10 CFR Part 50) (RIN-3150-AJ26),” SECY-15-0085, June 18, 2015, ADAMS Accession

No. ML15022A218.

21. Code of Federal Regulations, Title 10, Part 50, “Loss of All Alternating Current Power,” 10

CFR 50.63.

22. Code of Federal Regulations, Title 10, Part 50, “Conditions of Licenses,” 10 CFR

50.54(hh).

23. U.S. Nuclear Regulatory Commission, “Consequence Study of a Beyond-Design-Basis

Earthquake Affecting the Spent Fuel Pool for a U.S. Mark I Boiling Water Reactor,”

NUREG-2161, September 2014.

24. U.S. Nuclear Regulatory Commission, “Consideration of Economic Consequences within

the U.S. Nuclear Regulatory Commission’s Regulatory Framework” SECY-12-0110,

August 14, 2012.
2-1

2 ACCIDENT SEQUENCE ANALYSIS

This section details the accident sequence analysis that was performed to support the draft 

regulatory basis attached to SECY-15-0085 [1]. Using a semi-plant-specific approach, the risk 

evaluation assessed the risk impacts of potential containment protection and release reduction 

(CPRR) strategies for severe accidents initiated by extended loss of ac power (ELAP) events due 

to internal events and seismic events occurring at operating boiling water reactors (BWRs) with 

Mark I containment designs and reactor core isolation cooling (RCIC) systems. The focus on 

ELAP events is consistent with the wording of NTTF Recommendation 5.1, which alludes to 

“prolonged SBOs.” In addition to internal events and seismic events, ELAPs may also be caused 

by other types of external events (e.g., fires, floods, high winds); however, these were not 

included due to the amount of time and effort need to the collect the site-specific information 

needed to develop an appropriate logic model. It is also recognized that CPRR strategies may be 

beneficial in mitigating severe accidents that do not involve ELAP. As a result, the accident 

sequence analysis underestimates the potential benefits of CPRR strategies. NTTF 

Recommendation 5.1 applies to all BWRs with Mark I and Mark II containment designs. However, 

the accident sequence analysis excluded BWRs with Mark I containment designs and isolation 

condensers and BWRs with Mark II containments because the staff determined that it could 

develop sufficient information to inform the regulatory analysis by evaluating only the BWRs with 

Mark I containment designs and RCIC systems. This reduced the time and effort needed to 

complete the draft regulatory evaluation attached to SECY-15-0085. Table 2-1 lists all BWRs with 

Mark I and Mark II containment designs, and indicates their disposition in the accident sequence 

analysis.

Table 2-1 Consideration of BWR Mark I and Mark II plants in the accident sequence 

analysis

1 Shortly after the accident sequence analysis was commenced, the owner of the Vermont Yankee plant announced 

that it would cease operations in the fourth quarter of 2014; as a result, Vermont Yankee was excluded from the 

scope of the accident sequence analysis. 

2 Shortly after SECY-15-0085 was issued, the owners of the FitzPatrick and Pilgrim plants announced that these 

plants would cease operations in 2016 and 2019, respectively.

Included in the Accident Sequence 

Analysis 

Excluded from the Accident Sequence 

Analysis 

BWR Mark I and RCIC: 

Browns Ferry, Units 1, 2 and 3 

Brunswick, Units 1 and 2 

Cooper 

Duane Arnold 

Fermi, Unit 2 

FitzPatrick2

Hatch, Units 1 and 2 

Hope Creek, Unit 1 

Monticello 

Peach Bottom, Units 2 and 3 

Pilgrim, Unit 12

Quad Cities, Units 1 and 2 

BWR Mark I and RCIC: 

Vermont Yankee1

BWR Mark I and Isolation Condensers: 

Dresden, Units 2 and 3 

Nine Mile Point, Unit 1 

Oyster Creek 

BWR Mark II: 

Columbia 

La Salle, Units 1 and 2 

Limerick, Units 1 and 2 

Nine Mile Point, Unit 2 

Susquehanna, Units 1 and 2 
2-2

During the risk integration effort, the following risk metrics for each of the 20 regulatory analysis 

sub-alternatives defined in SECY-15-0085:

• Individual early fatality risk (0-1.3 miles and beyond)

• Individual latent cancer fatality risk (0-10 miles, 0-50 miles, and 0-100 miles)

• Population dose risk (0-50 miles and 0-100 miles)

• Offsite costs (0-50 miles and 0-100 miles)

• Land exceeding long-term habitability criterion (0-50 miles and 0-100 miles)

• Population subject to long-term protective actions (0-50 miles and 0-100 miles)

The following sections describe the rationale used to identify potential CPRR strategies (including 

their relationship to the options provided in SECY-12-0157 [2] and the alternatives provided in 

SECY-15-0085), explain the technical approach used to estimate the risk metrics, present the 

results obtained, discuss the sensitivity and parametric uncertainty analyses that were conducted, 

and provide the conclusions of the accident sequence analysis. 

2.1 Identification of CPRR Strategies

The CPRR accident sequence analysis began in March 2013 with the issuance of the staff 

requirements memorandum on SECY-12-0157 and concluded in June 2015 with the issuance of 

SECY-15-0085. During this period, the scope of the accident sequence analysis evolved as the 

staff identified various CPRR strategies, conducted analysis, solicited stakeholder input during 

public meetings, and incorporated the impacts of Fukushima-related regulatory actions. The 

following sections describe the factors that influenced the selection of CPRR strategies addressed 

by the accident sequence analysis, and identify the specific combinations of CPRR strategies that 

were addressed. 

2.1.1 Influencing Factors

A CPRR strategy is an action taken prior to or during the course of a severe accident to protect 

the containment’s structural integrity or to reduce the amount of radioactive material released to 

the environment. Examples include containment venting following core damage (a containment 

protection strategy) and the installation of engineered filters on the containment vent lines (a 

release reduction strategy). High-level strategies (e.g., containment venting) may be divided into 

more specific categories according to how they are implemented (e.g., wetwell venting or drywell 

venting). In order to conduct the accident sequence analysis, it is essential to define a set of 

possible CPRR strategies and to specify their implementation details so that a probabilistic logic 

model can be developed and quantified. The major factors that influenced the set of CPRR 

strategies addressed in the accident sequence analysis are summarized below.

2.1.1.1 Commission Direction

SECY-12-0157 identified four options to address the issue of containment venting for BWRs with 

Mark I and Mark II containments, which are reproduced below verbatim: 

1. Reliable hardened vents (Status Quo): Continue with the implementation of Order EA12-050 [3] for reliable hardened vents to reduce the likelihood of core damage and

failure of BWR Mark I and Mark II containments and take no additional regulatory action

to improve their ability to operate under severe accident conditions or to require the

installation of an engineered filtered vent system.
2-3

2. Severe accident capable vents order: Upgrade or replace the reliable hardened vents

required by EA-12-050 with a containment venting system designed and installed to

remain functional during severe accident conditions.

3. Filtered vents order: Design and install an engineered filtered containment venting

system that is intended to prevent the release of significant amounts of radioactive

material following the dominant severe accident sequences at BWRs with Mark I and

Mark II containments.

4. Severe accident confinement strategy: Pursue development of requirements and

technical acceptance criteria for confinement strategies and require licensees to justify

operator actions and systems or combinations of systems, such as suppression pools,

containment sprays, and separate filters to accomplish the function and meet the

requirements.

In response to SECY-12-0157, the Commission:

1. Approved Option 2 to issue a modification to Order EA-12-050 to require licensees of

Boiling Water Reactors (BWRs) with Mark I and Mark II containments to upgrade or

replace the reliable hardened vents required by Order EA-12-050 with a containment

venting system designed and installed to remain functional during severe accident

conditions,

2. Approved the development of technical bases and rulemaking for filtering strategies with

drywell filtration and severe accident management of BWR Mark I and II containments to

further consider Option 3 and Option 4, and

3. Directed that the technical bases should assume the installation of severe accident

capable hardened venting systems ordered under Option 2 and, as a consequence of

that action, should assume that the benefits of these vents accrue equally to engineered

filters and to filtration strategies.

In order to respond to the Commission’s direction, three high-level CPRR strategies were 

identified for subsequent analysis:

1. Severe accident containment venting (containment protection),

2. Installation of engineered filters on the containment vent lines (release reduction), and

3. Severe accident mitigation strategies, including termination of core melt progression and

core debris cooling (containment protection and release reduction).

2.1.1.2 Post-Accident Water Injection

The technical analysis performed to support SECY-12-0157 identified the need for post-accident 

water injection, which provides the following benefits:

1. Water injection to the reactor pressure vessel (RPV) can arrest a severe accident before

vessel breach.
2-4

2. Water injection to the containment, either directly into the drywell (DW) or indirectly into

the RPV (which subsequently flows through the vessel breach to the drywell), can

prevent loss of containment integrity due to liner melt-through.

3. Water injection reduces the temperature of the drywell atmosphere, which simplifies the

design of a severe accident capable drywell venting system.

The staff identified post-accident water injection as a potential CPRR strategy early in the accident 

sequence analysis. 

2.1.1.3 Actions Taken Prior to Core Damage

During a public meeting held in the spring of 2013, industry described an early (i.e., prior to core 

damage) venting strategy termed “anticipatory venting,” which was developed as part of the 

mitigating strategies required by Order EA-12-049 [4]. Prior to this meeting, it was assumed that 

containment venting would be initiated when the containment pressure approached the primary 

containment pressure limit (PCPL), approximately 60 psig. Anticipatory venting initiated prior to 

core damage at a pressure substantially lower than the PCPL would help to cool the wetwell 

inventory and, thus, prolong the operation of the RCIC pump. As a result of this meeting, the staff 

identified anticipatory venting as a potential CPRR strategy.

2.1.1.4 Order EA-13-109

As directed by the Commission in the staff requirements memorandum on SECY-12-0157, the 

staff rescinded Order EA-12-050 and issued Order EA-13-109 [5] to implement requirements for 

reliable hardened containment vents capable of operation under severe accident conditions. A 

phased approach to implementation was used to minimize delays in implementing the 

requirements originally imposed by EA-12-050. Phase 1 involves upgrading the venting 

capabilities from the containment wetwell to provide reliable, severe accident capable hardened 

vents to assist in preventing core damage and, if necessary, to provide venting capability during 

severe accident conditions. Phase 2 involves providing additional protections for severe accident 

conditions through installation of a reliable, severe accident capable drywell vent system or the 

development of a reliable containment venting strategy that makes it unlikely that a licensee would 

need to vent from the containment drywell during severe accident conditions.

If the post-accident water injection flow rate exceeds what is needed to replenish the wetwell 

inventory lost due to venting, then wetwell venting capability will eventually be lost because the 

wetwell vent connection will become submerged. Thus, the operating mode of post-accident water 

injection was identified as a potential CPRR strategy having two alternatives:

1. Severe accident water addition (SAWA): no effort is taken to prevent submergence of

the wetwell vent

2. Severe accident water management (SAWM): the post-accident water injection flow is

throttled as needed to prevent submergence of the wetwell vent.

2.1.2 Regulatory Analysis Sub-Alternatives

As the analysis progressed, specific combinations of CPRR strategies were termed “options” or 

“alternatives,” and were identified by a scheme proposed by industry. This approach was not 

entirely satisfactory because it caused confusion between the CPRR strategy options/alternatives 
2-5

being considered in the accident sequence analysis and the options/alternatives presented for the 

Commission’s consideration in SECY papers. Moreover, in August 2014, the staff determined that 

the original rulemaking name (filtering strategies) no longer matched the purpose of the activity. 

The staff felt it was more logical to have the rulemaking reflect the two issues being analyzed – 

enhanced containment protection and release reduction. Accordingly, the various combinations of 

CPRR strategies were organized into a set of regulatory analysis alternatives and sub-alternatives 

in preparation for developing a SECY paper to respond to the staff requirements memorandum on 

SECY-12-0157, as indicated in the following outline:

1. No Action/Base Case (leave Order EA-13-109 in place)

2. Overpressurization Measures

A. Make Order EA-13-109 generically applicable

3. Containment Failure Prevention Measures

A. Water addition via RPV

B. Water addition via DW

4. Release Reduction and Containment Failure Prevention Measures

A. Filtration Strategies

i. Vent Cycling

(1) Water addition via RPV

(2) Water addition via DW

ii. Water Management

(1) Water addition via RPV

(2) Water addition via DW

iii. Vent Cycling and Water Management

(1) Water addition via RPV

(2) Water addition via DW

B. Small Engineered Filter

i. Manual WWF before core damage, and manual WWF after core damage

(1) Water addition via RPV

(2) Water addition via DW

ii. Manual DWF before core damage and manual DWF after core damage

iii. Manual DWF before core damage and passive DWF after core damage

iv. Passive DWF before core damage and passive DWF after core damage

C. Large Engineered Filter

i. Manual WWF before core damage, and manual WWF after core damage

(1) Water addition via RPV

(2) Water addition via DW

ii. Manual DWF before core damage and manual DWF after core damage

iii. Manual DWF before core damage and passive DWF after core damage

iv. Passive DWF before core damage and passive DWF after core damage

Several features of this outline are noteworthy:

• The second regulatory analysis alternative was defined in order to consider the costs

and benefits associated with making Order EA-13-109 generically applicable, i.e., the

second regulatory analysis alternative explores a change in the “regulatory footprint”

concerning severe accident containment venting. There is no technical difference

between the first and second regulatory analysis alternatives.

• A single sub-alternative (2A) was identified for the second regulatory analysis

alternative; no other sub-alternatives for the second regulatory analysis alternative were

defined (i.e., the “A” is superfluous).
2-6

• The outline contained ambiguities about some alternatives. Specifically, no post-accident

water injection location was specified for sub-alternatives that utilize DWF venting and

engineered filters.

Table 2-2 lists the complete and fully specified regulatory analysis sub-alternatives that were 

considered in the accident sequence. It also traces each regulatory analysis sub-alternative back 

to the options provided in SECY-12-0157.

Additionally, Table 2-2 maps each regulatory analysis sub-alternative onto the alternatives 

provided in SECY-15-0085:

1. Take no action (Order EA-13-109 implemented without additional regulatory actions)

2. Pursue rulemaking to make Order 13-109 generically applicable

3. Pursue rulemaking to address containment protection against multiple failure modes by

making Order EA-13-109 generically applicable and requiring external water addition

points (external to the reactor building) that would allow for post-accident water injection

into the RPV or DW.

4. Pursue rulemaking to address both containment protection against multiple failure

modes and release reduction measures for controlling releases through the containment

venting systems. This alternative included making Order EA-13-109 generically

applicable, requiring post-accident water injection, and reducing the fission products

released from the containment by either implementing filtering strategies or installing

engineered filters.

As the regulatory alternatives for CPRR were being developed, it became apparent that all 

licensees intended to comply with Phase 2 of Order EA-13-109 by implementing a post-accident 

water injection strategy. Thus, there is no technical difference among the first three alternatives 

presented in SECY-15-0085; rather, the first three alternatives in SECY-15-0085 explore 

differences in the “regulatory footprint” for various CPRR strategies. 
2-7

Table 2-2 Regulatory analysis sub-alternatives Index Regulatory Analysis Sub-Alternative SECY-12-0157 Option SECY-15-0085 Alternative(s)

Before Core Damage After Core Damage Vent Priority Venting Actuation Venting Operation Mode Reclose Vent if Core Damage is Imminent Post-Accident Water Injection Location Post-Accident Water Injection Operating Mode Vent Priority Venting Actuation 

Venting Operation Mode Filter Size 

1 1 2 n/a WWF M AV yes no n/a WWF M OLO no

2 2A 2 n/a WWF M AV yes no n/a WWF M OLO no

3 3A 2 1,2,3 WWF M AV yes RPV SAWA WWF M OLO no

4 3B 2 1,2,3 WWF M AV yes DW SAWA WWF M OLO no

5 4Ai(1) 4 4 WWF M AV yes RPV SAWA WWF M VC no

6 4Ai(2) 4 4 WWF M AV yes DW SAWA WWF M VC no

7 4Aii(1) 4 4 WWF M AV yes RPV SAWM WWF M OLO no

8 4Aii(2) 4 4 WWF M AV yes DW SAWM WWF M OLO no

9 4Aiii(1) 4 4 WWF M AV yes RPV SAWM WWF M VC no

10 4Aiii(2) 4 4 WWF M AV yes DW SAWM WWF M VC no

11 4Bi(1) 3 4 WWF M AV yes RPV SAWA WWF M OLO S 

12 4Bi(2) 3 4 WWF M AV yes DW SAWA WWF M OLO S 

13 4Bii 3 4 DWF M AV yes DW SAWA DWF M OLO S 

14 4Biii 3 4 DWF M AV yes DW SAWA DWF P OLO S 

15 4Biv 3 4 DWF P OLO no DW SAWA DWF P OLO S 

16 4Ci(1) 3 4 WWF M AV yes RPV SAWA WWF M OLO L 

17 4Ci(2) 3 4 WWF M AV yes DW SAWA WWF M OLO L 

18 4Cii 3 4 DWF M AV yes DW SAWA DWF M OLO L 

19 4Ciii 3 4 DWF M AV yes DW SAWA DWF P OLO L 

20 4Civ 3 4 DWF P OLO no DW SAWA DWF P OLO L 

Venting Priority 

DWF drywell first strategy 

WWF wetwell first strategy 

Venting Actuation 

M 

P 

manual 

passive (rupture disk) 

Venting Operation Mode 

AV

OLO 

VC

anticipatory venting; open at 15 psig and 

leave open 

open at PCPL and leave open 

venting cycling at PCPL with 10 psi band 

Post-accident Water Injection Location 

DW

RPV 

drywell via external connection 

reactor pressure vessel via external 

connection 

Post-accident Water Injection Operating Mode 

SAWA severe accident water addition 

SAWM severe accident water management 

Filter Type 

L 

S 

large (DF=1000) 

small (DF=10) 
2-8

The staff considered a variety of implementation details associated with the CPRR strategies 

identified in the previous sections:

1. Venting priority

a. Wetwell-first venting (WWF): the wetwell vent is preferential opened, with the drywell

providing redundancy

b. Drywell-first venting (DWF): the drywell vent is preferentially opened, with the wetwell

vent providing redundancy

2. Venting actuation

a. Manual: the plant operators open the vents

b. Passive: the vents are provided with rupture discs, which improves their reliability

since operator action is not required

3. Venting operation mode

a. Before core damage

i) Anticipatory venting (AV) at 15 psig or less

ii) Open-and-leave-open (OLO) venting at the PCPL or lower

b. After core damage

i) Vent cycling (VC) at the PCPL within a 10 psi band

ii) Open-and-leave-open (OLO) venting at the PCPL

4. Vent reclosure if core damage is imminent: yes or no

5. Post-accident water injection location: RPV or DW

2.2 Technical Approach

The technical approach used to develop the accident sequence analysis of potential CPRR 

strategies was based on simplified probabilistic risk assessment (PRA) methods. An overview of 

NRC policy regarding the level of detail to be provided in regulatory analyses is provided in 

Chapter 4 of the NUREG/BR-0058 [6]. As discussed in NUREG/BR-0184 [7], the emphasis in 

implementation of the NRC regulatory analysis guidelines should be on simplicity, flexibility, and 

commonsense, both in terms of the type of information supplied and in the level of detail provided. 

The level of treatment given to a particular issue in a regulatory analysis should reflect how crucial 

that issue is to the bottom line recommendation of the regulatory analysis.

The following sections discuss the rationale used to develop the technical approach, summarize 

the technical approach used, detail the logic model (event tree) development, describe supporting 

data analyses, and explain how the logic models were quantified and combined with results from 

the consequence analysis to produce risk metrics for each regulatory analysis sub-alternative.
2-9

2.2.1 Rationale Used to Develop the Technical Approach 

The following factors were considered during the development of the technical approach for the 

CPRR accident sequence analysis:

1. The risk integration should provide risk metrics for each of the 20 CPRR regulatory

analysis sub-alternatives according to the schedule established by the Commission and

the resources allotted by NRC management.

2. As discussed in NUREG/BR-0058 and NUREG/BR-0184, the risk integration should

provide fleet-average risk estimates. As a result, the technical approach should consider

the impacts of plant-to-plant variability.

3. Consistent with Recommendation 5.1 in the Fukushima Near-Term Task Force report

[8], the accident sequence analysis should focus on accidents initiated by ELAP events.

4. The generic estimates of release sequence frequencies and conditional consequences

provided in NUREG/BR-0184 were developed from previous probabilistic risk

assessments that did not consider CPRR strategies and, therefore, cannot be used to

provide an adequate technical basis for the CPRR risk integration.

5. CDETs should be developed in order to:

a. Model the impact of equipment failures and operator actions occurring prior to core

damage that affect severe accident progression and the probability that CPRR

strategies are successfully implemented,

b. Match the initial and boundary conditions used in the thermal-hydraulic simulation of

severe accidents (MELCOR calculations),

c. Probabilistically consider mitigating strategies for beyond design basis external

events required by Order EA-12-049 (e.g., the FLEX strategies, including anticipatory

venting).

6. The CPRR strategies addressed in the set of 20 regulatory analysis sub-alternatives are

specified at a conceptual level. As a result, it is acceptable to develop high-level generic

APETs to model the CPRR strategies because no information is available about their

specific design details.

2.2.2 Summary of the Technical Approach 

Consideration of the factors listed above resulted in development of the following technical

approach to conducting the accident sequence analysis: 

1. Accident sequences are initiated by ELAP events due to internal events and seismic

events. An ELAP is defined as a station blackout (SBO) that lasts longer than the SBO

coping duration specified in 10 CFR 50.63. ELAP frequencies are semi-plant-specific

since they are based on the plant’s emergency power system (EPS) class, SBO coping

time, and site-specific seismic hazard.
2-10

2. The core-damage event trees (CDETs) and accident progression event trees (APETs)

model a stylized (representative) BWR plant having a Mark I containment design and

RCIC system.

3. The CDETs and APETs were developed using a modular approach that allows them to

be combined and configured as needed to model each regulatory analysis

sub-alternative.

4. The CDETs and APETs are quantified with industry-average reliability parameters

(failure rates and failure-on-demand probabilities) and seismic fragilities developed from

individual plant examination of external events (IPEEE) information.

5. The CDETs and APETs are quantified with scoping estimates of human error

probabilities. Sensitivity studies confirmed that altering the human error probabilities

(HEPs) values did not impact the results in any significant manner.

6. Similar core-damage sequences are grouped together using plant damage states

(PDSs), which provide the input to the APETs.

7. The CDETs are solved for each plant within the scope of the analysis, then used to

determine fleet-average PDS frequencies.

8. Similar APET sequences are grouped together using release categories (RCs).

9. Mean risk estimates are developed by multiplying the frequencies of significant RCs by

the RC conditional consequences, then summing over all RCs.

The risk integration is not considered to be a PRA because it does not include all of the technical

elements specified in Regulatory Position 1.2 of RG 1.200 [9]. 

2.2.3 Logic Model Development 

As shown in Table 2-2, each regulatory analysis subalternative is defined by a specific

combination of CPRR strategies that are intended to prevent the occurance of a severe accident

or to mitigate its consequences should it occur. A probabilistic perspective recognizes that one or

more of the CPRR strategies may not be successfully implemented. An accident sequence 

consists of an initiating event (the ELAP event), followed by a unique combination of CRPP

strategy successes and failures that results in core damage and the subsequent release of

radioactive materials to the environment. Logic model development uses a systematic process

(event tree analysis) to identify the set of possible accident sequences associated with a specific

regulatory analysis subalternative that might occur, and to estimate their frequency of occurrence.

The logic model provides the fundamental probabilistic framework for assessing the risk

associated with a specific regulatory analysis subalternative.

Logic model development relies on the results of the accident progression analysis (MELCOR

analysis). An accident progression analysis is a simulation of a specific accident sequence that is

conducted in order to (a) understand how the specific combination of CPRR strategy successes

and failures affects the plant, and (b) estimate the fission product release (the source term)

resulting from the accident sequence. The nomenclature used to identify MELCOR calculations

somewhat overlaps with the nomenclature used to identify the regulatory analysis subalternatives.

The reader is cautioned to remember that, in the context of the accident sequence analysis, a 
2-11

specific regulatory analysis subalternative refers to a set of CPRR strategies and accident 

sequences, and that each accident sequence is linked to a specific MELCOR calculation.

The following sections provide the assumptions and ground rules used to develop the CDETs and 

APETs, describe the modular approach used for their development, and provide additional

supporting information. 

2.2.3.1 Assumptions and Ground Rules 

During development of the accident sequence analysis, the set of strategies used to comply

with the requirements of Order EA-12-049 were referred to as “SBO mitigating strategies,”

“FLEX strategies,” and “mitigation of beyond-design-basis events (MDBDE) strategies.” The

term “FLEX strategies” is a reference to the set of diverse and flexible coping strategies

(FLEX) developed by the Nuclear Energy Institute, as defined in NEI-12-06 [10], to provide 

guidance to its members about complying with Order EA-12-049. The accident sequence 

analysis nomenclature uses the modifier “FLEX” in CDET and APET headings to refer to

portable equipment and associated operator actions related to the operation of portable 

equipment.

A review of licensee approaches to implementing the requirements of Order EA-12-049 

showed a wide variation of implementation details. To make the accident sequence analysis

and risk integration tractable within the allotted schedule and budget, the CDETS and

APETs were developed for a generic BWR plant having a reactor core isolation cooling

(RCIC) system and a Mark I containment using the following assumptions and ground rules:

1. The CDETs model the first phase (use of installed plant equipment) and the second

phase (use of onsite portable equipment) of the SBO mitigation strategy. The third

phase (use of offsite portable equipment) was not considered due to lack of a

suitable probabilistic approach for modeling the interactions between the plant and

the National Response Centers established by industry’s Strategic Alliance for FLEX

Emergency Response (SAFER).

2. In the CDETs, the first phase of the SBO mitigation strategy was assumed to last

four hours. During this first phase, core cooling is provided by the RCIC system

drawing from the suppression pool:

a. No probabilistic consideration was taken for supplying the RCIC pump from

the condensate storage tank (assumed to be non-seismically qualified).

b. Based on the MELCOR analysis, the RCIC pump only needs to operate for

two hours during the first phase. If the RCIC pump fails after two hours, then

core damage will occur at about four hours after the ELAP occurs.

c. No probabilistic consideration was taken for using the high-pressure coolent

injection (HPCI) system. HPCI is a relatively high flowrate system

(approximately 10 times higher than the portable FLEX pump), and would

need to be manually cycled on and off to prevent overfilling the RPV. There is

no information that can be used to estimate the reliability of HPCI while it is

operating in a cyclical mode.

d. The CDETs probabilistically consider local manual operation of the RCIC

pump (termed “blackstart and blackrun”) if dc power fails.
2-12

e. No probabilistic consideration was taken for using the portable FLEX pump to

provide core cooling if the RCIC pump fails because there is not sufficient

time to get it aligned during the first phase.

f. There is no need to vent the containment during the first phase.

g. The plant operators will reduce RPV pressure using the safety relief valves

(SRVs) to a range of 200-400 psig in order to minimize SRV cycling and to

minimize heatup of the suppression pool.

h. The CDETs probabilistically consider local manual operation of SRVs if dc

power fails.

3. In the CDETs, the second phase of the SBO mitigation strategy was assumed to last

for 68 hours (i.e., the accident sequence analysis used a total mission time of 72

hours):

a. The 72-hour total mission time used in the accident sequence analysis

consists of the time needed for onsite diagnosis leading to a decision to

request assistance from the National SAFER Response Centers (NSRC)

(nominally 1 hour from the occurance of station blackout), the time needed to

load and transport offssite portable equipment from the NSRC to the site

(nominally 24 hours after the request is received), and the time needed to

align the offsite portable equipment once it arrives at the site. Each of these

times is highly uncertain; the 72-hour mission time is believed to be

conservative. It should be noted that the only basic event in the accident

sequence analysis whose probability depends on the assumed 72-hour

mission time is the event FLEXP-FTR, “FLEX pump fails to run.”

b. Core cooling is maintained by operation of the RCIC pump. Makeup to the

suppression pool is provided by the portable FLEX pump.

c. Except for regulatory analysis alternatives 4Biv and 4Civ, the operators will

initiate anticipatory containment venting in order to minimize heatup of the

suppression pool and prolong RCIC operation. (Regulatory analysis

alternatives 4Biv and 4Civ reflect a passive drywell-first venting strategy.)

d. The WW and DW vents are redundant, i.e., the DW vent can be used to

provide anticipatory venting if the WW vent fails closed, and vice versa.

e. The CDETs probabilistically consider local manual operation of the

containment vent valves if dc power fails.

f. There is no need to provide RCIC pump room ventilation.

g. A portable generator must be aligned to provide dc power within four hours.

Many plants have battery lifetimes longer than four hours, which may be

further extended by shedding non-essential dc loads. The scoping approach

used to conduct the human reliability analysis (HRA) does not depend on the

assumed battery lifetime.

h. If the RCIC pump fails, core cooling can be provided by aligning the portable

FLEX pump for RPV injection and depressurizing the RPV below the portable

FLEX pump’s shutoff head.

4. In the CDETs, it is assumed that the operators will attempt to reclose the

containment vent valves, in accordance with the BWR Owners’ Group Emergency

Procedure and Severe Accident Guidelines (EPG/SAGs), if they recognize that core

damage is occurring.
2-13

5. In the APETs, containment overpressurization failure is prevented by opening the

containment vent valves:

a. The WW and DW vents are redundant, i.e., the DW vent can be used to

provide anticipatory venting if the WW vent fails closed, and vice versa.

b. The CDETs probabilistically consider local manual operation of the

containment vent valves if dc power fails.

c. MELCOR calculations indicate that the containment must be vented prior to

vessel breach due to the buildup of non-condensable gases generated during

fuel-clad oxidation.

d. Successful post-core-damage containment venting is a controlled release of

radioactive materials to the environment, which is allowable under 10 CFR

Part 50, Appendix A, General Design Criterion 16 and consistent with the

Commission’s direction in the SRM to SECY-12-0157 to assume the

installation of severe accident capable hardened venting systems ordered

under Option 2.

6. In the APETs, post-core-damage water injection to the RPV can prevent vessel

breach if it is initiated prior to lower plenum dryout and the RPV is depressurized

below the shutoff head of the portable FLEX pump using the SRVs. In contrast,

regulatory analysis alternatives involving post-core-damage DW injection cannot

prevent vessel breach.

2.2.3.2 Modular Approach to Logic Model Development 

A modular approach was used to develop the CDETs and APETs in order to streamline the 

development of risk estimates. As shown in Figure 2-1, three CDETs and six APETS were 

developed and subsequently combined as appropriate to develop the set of RC frequencies

for each regulatory analysis sub-alternative. This modeling technique proved to be a

responsive and efficient approach during the CPRR accident sequence analysis since the 

set of regulatory analysis sub-alternatives slowly evolved as the analysis progressed.


)LJXUH 0RGXODUDSSURDFKWRHYHQWWUHHGHYHORSPHQW
2-15

2.2.3.3 Event Tree Development 

The CDETs and APETs do not utilize supporting fault tree logic (i.e., they are not linked event

tree/fault tree models that require the use of Boolean solution methods to generate minimal cut

sets for each accident sequence). The CDET diagrams are shown in Appendix A, and supporting

information is provided in Appendix B. The APET diagrams are shown in Appendix C, and 

supporting information is provided in Appendix D.

Each CDET has 280 core-damage sequences that were binned into 139 PDSs. Each APET

contains 72-84 release sequences (depending on the regulatory analysis alternative being

modeled) that were mapped into 24 release categories. Table 2-3 and Table 2-4 present the PDS

and RC naming schemes.

Table 2-3 Plant damage state naming scheme

A plant damage state name consists of five attributes, arranged as follows: 

T-PP-VV-DD-FF

PDS Attribute Possible Values 

Code Description Code Description 

T RCIC failure time E Early (0-4 hours) 

M Mid-term (4-16 hours) 

L Long-term ( at 16 hours) 

PP RPV pressure at 

time of core 

damage 

HP High pressure (SRV cycling) 

MP Medium pressure (200-400 psig) 

LP Low pressure (below portable FLEX pump shutoff head) 

VV Containment vent 

status 

DW Drywell vent is open at time of core damage 

IS Both vents are closed at the time of core damage 

WW Wetwell vent is open at the time of core damage 

DD dc power status LT dc power fails long-term (unrecovered battery depletion) 

OK dc power is available throughout the accident 

ST dc power fails short-term (before battery depletion) 

XX Indeterminate (dc power status not important to 

subsequent logic) 

FF FLEX pump status OK Portable FLEX pump is working 

F FLEX pump hardware has failed 

H Operator fails to align FLEX prior to core damage 

XX Indeterminate (FLEX status not asked in the coredamage sequence) 
2-16

Table 2-4 Release category naming scheme

2.2.4 Supporting Data Analysis 

The following sections describe how the frequency of ELAP events was estimated, identify the 

data source used to estimate hardware-related failure events, and discuss the reasons why a 

scoping human reliability analysis was used. 

2.2.4.1 ELAP Frequency 

The accident sequence analysis included ELAPs that are initiated by the four categories of lossof-offsite-power events (LOOPs) that are included in the staff’s Standardized Plant Analysis of

Risk (SPAR) internal event models (plant-centered, switchyard-centered, grid-related, and 

weather-related, as defined in NUREG/CR-6890 [11]) and by seismic events. It was recognized 

that CPRR strategies may also be beneficial during other types of accidents (e.g., short duration 

SBOs, LOOPs that do not degenerate into SBOs, accidents that do not involve LOOP such as

loss-of-coolant accidents, and accidents that are initiated by internal floods, internal fires, external

floods and other types of external events). A complete assessment of CPRR strategies that

includes these types of accidents would have required the development of plant-specific, internal

and external event Level 3 PRAs for each BWR Mark I plant, which exceeds the required level-ofdetail established in NUREG/BR-0058 and NUREG/BR-0184 for regulatory analyses.

The frequencies of ELAP events used in the accident sequence analysis are semi-plant-specific 

since they are based on the plant’s emergency power system (EPS) class (the amount of

redundancy provided by the onsite emergency ac power sources), the SBO coping time, and the 

site-specific seismic hazard.

The frequency of internal event ELAPs was estimated by probabilistic convolution of the 

associated LOOP frequency, the probability of emergency power system (EPS) failure, and the 

probability that offsite power was not recovered within the SBO coping duration. Specifically:

A release category name consists of five attributes, arranged as follows: 

RRRRR-VV-DDD

Attribute Possible Values 

Code Description Code Description 

RRRRR RPV

depressurization

SRV RPV is depressurized using the SRVs 

HP RPV is at high pressure (high-pressure melt 

scenario) 

MSLCR RPV depressurized due to main steamline creep 

rupture 

VV Containment vent 

status 

WW Wetwell vent is open 

DW Drywell vent is open 

OP Containment overpressurization failure 

DDD Core debris 

location 

IVR In-vessel retention 

EVR Ex-vessel retention 

LMT Liner melt-through 
2-17

dte tt ff SBO t

ELAP LOOP

λ λ

σ

μ −

∞

∫ 



 



 



 



 −+ = Φ−

0

)ln( 1

where:

fELAP = ELAP frequency

fLOOP = LOOP frequency

Φ = standard normal cumulative distribution function

tSBO = SBO coping time

μ,σ = log-normal parameters of the offsite power recovery curve

λ = EPS failure rate

LOOP frequencies and offsite power recovery curves (which are assumed to have a log-normal 

distribution) were obtained from NUREG/CR-6890. The probability of EPS failure was obtained 

from the 2011 update to NUREG/CR-5500, Vol. 5 [12], which is based on the staff’s internal event 

SPAR models, reflects the number of onsite emergency power sources (the EPS class), and 

includes contributions from random equipment failures, common-cause failures, test/maintenance 

unavailability, and pre-initiator human failure events. Specific parameter values and the results of 

the probabilistic convolution are provided in Appendix E.

The frequency of seismic ELAPs was estimated by probabilistic convolution of the site-specific 

seismic hazard curve for peak ground acceleration with a model (top logic) of the EPS that 

included non-seismic failure modes and seismic failures of the emergency diesel generator 

equipment including the engine, generator, controls, day tank and fuel oil storage tank), ac 

switchgear, and dc switchgear. The convolution was performed using ten seismic bins that span 

the range of 0.03g to 3g peak ground acceleration. Seismic failures were assumed to be 

completely correlated (i.e., the seismically induced failure of a specific component implies the 

failure of all other similar and redundant components). No probabilistic consideration was given for 

recovering offsite power after a seismic event.

Seismic hazard curves were obtained from licensee responses (Browns Ferry [13], Brunswick 

[14], Cooper [15], Duane Arnold [16], Fermi [17], FitzPatrick [18], Hatch [19], Hope Creek [20], 

Monticello [21], Peach Bottom [22], Pilgrim [23], and Quad Cities [24]) to the 10 CFR 50.54(f) 

information request associated with NTTF Recommendation 2.1 [25].

Industry-average seismic fragilities were developed from information contained in the Individual 

Plant Examinations of External Events and the Screening, Prioritization and Implementation 

Details (SPID) document developed by the Electric Power Research Institute (EPRI) in response 

to NTTF Recommendation 2.1 [25]. Seismic fragility is characterized by three parameters:

C50 = median seismic capacity

βR = logarithmic standard deviation due to randomness

βU = logarithmic standard deviation due to uncertainty

For each component, a linear opinion pool (discrete mixture distribution) was formed to develop 

an industry-average value of the median seismic capacity:
2-18

=∑

i

C ,50 avg iCw 1,50

where:

C50,avg = industry average median seismic capacity

wi = subjectively assigned weight of the i’th IPEEE fragility data source

C50,I = median seismic capacity of the i’th IPEEE fragility data source

IPEEE fragility data sources that did not contain complete information (C50 and βU) were weighted 

less than complete sources.

A similar linear opinion pool was developed to estimate percentiles of the seismic capacity 

distribution function:

∑ 













 − Φ=

i iU

i

i

CC

w

,

,50 lnln

β

α α

where Cα denotes the α’th percentile of seismic capacity distribution function, which found using 

numerical methods for α = 0.05 and α = 0.95. The industry-average logarithmic standard deviation 

for uncertainty was then determined from:

05.0

95.0

95.0

,

1

C

C

z

βU avg =

where z0.95 denotes the 95th percentile of the standard normal distribution (approximately 1.645).

Appendix F provides the seismic top logic, a summary of the seismic fragility parameter s,

worksheets showing the details of the linear opinion pools used to develop industry-average 

fragility parameters, and plant-specific results of the seismic convolutions. 

2.2.4.2 Hardware-Related Failure Data 

Reliability parameters (failure rates and failure-on-demand probabilities) were based on estimates

used in the staff’s Standardized Plant Analysis of Risk (SPAR) models [28]. Appendix B lists the

hardware-related failure event probabilities that are incorporated in the CDETs. Appendix D

provides the hardware-related failure event probabilities that are incorporated into the APETs. 

2.2.4.3 Human Reliability Analysis 

The FLEX mitigation strategies represented in the CDETs and the CPRR strategies represented 

in the APETs rely on operator actions for their implementation. Some of these operator actions

are to be performed in the main control room, and some are to be performed at various locations

through the plant. All operator actions may need to be performed following a seismic event that is

large enough to cause the occurrence of an ELAP event.

The assessment of the human error probabilities (HEPs) needed to quantify the CDETs and APETs

proved to be challenging. The CPRR strategies are conceptual designs; accordingly, they are not

incorporated into the EPG/SAGs or into licensee training programs. The staff gained some insight

into how the CPRR strategies would be implemented through interactions with external stakeholders
2-19

during the development of the event trees. However, these interactions did not provide an adequate 

technical basis for completing the steps of an HRA of the CPRR strategies. No HRA method is 

capable of providing detailed HEP estimates at the conceptual design stage. Moreover:

1. Current HRA methods are inadequate for post-core-damage analysis because they are

geared to supporting at-power, Level 1, internal events PRA, and therefore fail to

recognize and appropriately capture the increased complexity of post-core-damage

scenarios.

2. There is little actual experience with severe accidents to guide our understanding of

operator responses in post-core-damage conditions.

3. EPG/SAGs differ from Emergency Operating Procedures (EOPs) in a number of ways

including format, level of detail, prescriptiveness, and requirements for decision-making.

4. In general, there is less frequent training on EPG/SAGs as compared to the EOPs. In

addition, most training simulators are not equipped to model plant behavior after the

onset of core damage.

5. Cues to the operators may not be available or may be ambiguous. As a result, there is

less accurate information on plant conditions that are important inputs to decision

making.

6. The nature of the teamwork among the licensee staff responding to a severe accident is

different as compared to responding to an off-normal situation. Prior to core damage, a

small cohesive team in the main control room is tasked with responding. In contrast,

following core damage, a larger number of people are involved who are situated at

multiple distributed locations.

7. Following core damage, assessment responsibilities shift from the control room

operators to technical support center.

8. Staffing may be inadequate for responding to site-wide events that involve multiple

radiological sources.

9. Access to vital plant locations may be impaired due to the damage caused by seismic

events or radiological hazards created by core damage.

Therefore, the accident sequence analysis used a set of scoping HEP estimates, supplemented 

with various sensitivity analyses that are focused on understanding the importance of operator

actions to the CPRR strategies. Actions that take place in the main control room were assigned a 

scoping failure probability of 0.1. Actions that take place outside of the main control were assigned 

a scoping failure probability of 0.3. Appendix B provides the pre-core-damage human failure 

events (HFEs) that are incorporated in the CDETs, along with their scoping HEPs. Appendix D

provides the post-core-damage HFEs that are incorporated into the APETs, along with their

scoping HEPs. 

2.2.5 Risk Quantification 

The initial step in the risk quantification process was the determination of fleet-average PDS

frequencies for each CDET, which was quantified twice for each plant included in the scope of the 


DFFLGHQWVHTXHQFHDQDO\VLV$VVKRZQLQ)LJXUHWKHILUVWTXDQWLILFDWLRQZDVGRQHXVLQJWKH

VHPLSODQWVSHFLILFLQWHUQDOHYHQWV(/$3IUHTXHQFLHVDQGWKHVHFRQGTXDQWLILFDWLRQZDVPDGH

XVLQJDVHWRIIRXUVHLVPLF(/$3IUHTXHQFLHVWKDWDFFRXQWIRUWKHVXFFHVVRUIDLOXUHRIWKHGF

SRZHUV\VWHPDQGWKH5&,&SXPS7KLVDSSURDFKZDVFRQYHQLHQWWRXVHVLQFHWKHRQO\KHDGLQJV

LQWKH&'(7WKDWLQFOXGHVHLVPLFIDLOXUHVDUH(/$3'&IDLOXUHRIGFSRZHULQ3KDVHDQG

5&,&IDLOXUHRI5&,&LQ3KDVH

)LJXUH &RUHGDPDJHHYHQWWUHHTXDQWLILFDWLRQ

7KHVHFRQGVWHSLQWKHULVNTXDQWLILFDWLRQSURFHVVZDVWKHGHWHUPLQDWLRQRI5&IUHTXHQFLHVIRU

HDFK$3(77KH$3(7GRZQEUDQFKSUREDELOLWLHVGHSHQGRQWKHVSHFLILF3'6WKDWLVLQSXWWRWKH

$3(7DVUHVXOWHDFK$3(7ZDVTXDQWLILHGWLPHVWKHQXPEHURIXQLTXH3'6VIRUHDFK

&'(77KHWRWDOVHWRI5&IUHTXHQFLHVZDVGHWHUPLQHGE\VXPPLQJWKH5&IUHTXHQFLHV

JHQHUDWHGE\WKH$3(7TXDQWLILFDWLRQSURFHVV

7KHWKLUGVWHSLQWKHULVNTXDQWLILFDWLRQSURFHVVZDVWKHLGHQWLILFDWLRQRIVLJQLILFDQW5&V$

VLJQLILFDQW5&LVRQHRIWKHVHWRI5&VWKDWZKHQUDQNHGFRPSRVHRIWKH&')RUWKDW

LQGLYLGXDOO\FRQWULEXWHVPRUHWKDQWRWKH&')

7KHIRXUWKVWHSLVWKHULVNTXDQWLILFDWLRQSURFHVVLVWKHGHWHUPLQDWLRQRIPHDQULVN5DYJZKLFKLV

GHILQHGDV

 ¦

L

DYJ &I5 LL

ZKHUHILGHQRWHVWKHIUHTXHQF\RIWKHL¶WKVLJQLILFDQW5&DQG&LGHQRWHVWKHFRQGLWLRQDO

FRQVHTXHQFHDVVRFLDWHGZLWKWKHL¶WK5&DVGHWHUPLQHGE\WKHFRQVHTXHQFHDQDO\VLV&RQGLWLRQDO

FRQVHTXHQFHVZHUHDVVLJQHGE\PDSSLQJHDFKVLJQLILFDQW5&IRUDJLYHQUHJXODWRU\DQDO\VLVVXE

DOWHUQDWLYHWRDQDVVRFLDWHG0(/&25FDVH6SHFLILFDOO\WKHPDSSLQJZDVDFKLHYHGE\PDWFKLQJ

WKHSODQWVWDWXVVSHFLILHGE\WKH5&DWWULEXWHVLHWKHFRPELQDWLRQRIVXFFHVVHVDQGIDLOXUHVWKDW

OHGWRUHOHDVHLQFOXGLQJWKHLUWLPHRIRFFXUUHQFHWRWKHLQLWLDODQGERXQGDU\FRQGLWLRQVRIWKH

0(/&25FDVH,QWXUQHDFK0(/&25FDVHZDVPDSSHGWRDQDVVRFLDWHG0$&&6ELQEDVHGRQ
2-21

consideration of the source term, as further discussed in the consequence analysis chapter. Some 

RCs could not be directly mapped to one of the MELCOR cases; in this situation, an appropriate 

“composite” source term was developed by reviewing the set of MELCOR cases for similar plant 

conditions, including the timing and quantity of releases. Table 2-5 presents the mapping of 

significant RCs to MELCOR cases for each regulatory analysis sub-alternative. In this table, 

composite MELCOR cases are indicated by the letter “C,” followed by a unique three-digit 

identifier. 1.1.1.1.1 APPENDIX A describes the build-up of composite MELCOR cases and 

MACCS bins.

Table 2-5 Mapping release categories to MELCOR cases

APET APET-1 APET-2 

Regulatory 

Analysis SubAlternative 

Release 

Category 

1 – 1 

2 – 2A

3 - 3A

5 - 4Ai(1) 

7 - 4Aii(1) 

9 - 4Aiii(1) 

11 - 4Bi(1) 16 - 4Ci(1) 

HP-DW-EVR

HP-WW-EVR

SRV-DW-EVR 11 11DF10 11DF1000 

SRV-WW-EVR 11 11DF10 11DF1000 

SRV-DW-IVR C108 C108DF10 C108DF1000 

SRV-OP-IVR C108 C108 C108 

SRV-WW-IVR C108 C108DF10 C108DF1000 

HP-DW-LMT 2 2 2DF10 2DF1000 

HP-OP-LMT 2 2 2DF10 2DF1000 

HP-WW-LMT 2 2 2DF10 2DF1000 

SRV-DW-LMT C104 C104 C104DF10 C104DF1000 

SRV-OP-LMT C104 C104 C104DF10 C104DF1000 

SRV-WW-LMT C104 C104 C104DF10 C104DF1000 

APET APET-3 

Regulatory 

Analysis SubAlternative 

Release 

Category 

4 - 3B 6 - 4Ai(2) 8 - 4Aii(2) 10 - 4Aiii(2) 12 - 4Bi(2) 17 - 4Ci(2) 

HP-DW-EVR

HP-WW-EVR C101 30 C105 28 C106DF10 C106DF1000 

SRV-DW-EVR 26dw 31dw 25dw 29dw 26dwDF10 26dwDF1000 

SRV-WW-EVR 26 31 25 29 26DF10 26DF1000 

SRV-DW-IVR 

SRV-OP-IVR 

SRV-WW-IVR 

HP-DW-LMT 2 2 2 2 2DF10 2DF1000 

HP-OP-LMT 2 2 2 2 2DF10 2DF1000 

HP-WW-LMT 2 2 2 2 2DF10 2DF1000 

SRV-DW-LMT C104 C104 C104 C104 C104DF10 C104DF1000 

SRV-OP-LMT C104 C104 C104 C104 C104DF10 C104DF1000 

SRV-WW-LMT C104 C104 C104 C104 C104DF10 C104DF1000 
2-22

2.3 Results

In addition to the risk estimates developed to support the draft regulatory basis attached to SECY15-0085, a variety of intermediate results were developed to provide risk insights about the CPRR

strategies. These results are presented in the following sections.

2.3.1 Results from the CDET Quantification

Table 2-6 provides the point-estimate ELAP frequency, core-damage frequency (CDF), and 

conditional core-damage probability (CCDP) by site for the wetwell-first containment venting strategy.

Table 2-7 and Table 2-8 provide the same information for the drywell-first (passive actuation) and the 

drywell-first (manual actuation) containment venting strategies. For ELAPs initiated by internal events,

plant-to-plant variations of ELAP frequencies and core-damage frequencies (CDFs) are due to 

differences in the EPS class and SBO coping time. For ELAPs initiated by seismic events, plant-toplant variations are due to differences in the seismic hazard curves.

The CCDP values given in Tables 2-6, 2-7 and 2-8 provide insight into the efficacy of the FLEX 

mitigating strategies. Overall, the FLEX strategies reduce the probability of core damage due to 

ELAP events by more than 50%. For a given pre-core-damage venting strategy:

• Each plant has the same CCDP for ELAPs initiated by internal events, which is due to

the use of a generic BWR plant model in the accident sequence.

• The CCDP values exhibit minor variations for ELAPs initiated by seismic events, which

is due to differences in the seismic hazard curves.

The CDF and CCDP values are somewhat lower for passive venting than for manual venting 

since the passive rupture disk is more reliable than operator action. 

APET APET-4 APET-5 APET-6 

Regulatory 

Analysis SubAlternative 

Release 

Category 

15 - 4Biv 20 - 4Civ 14 - 4Biii 19 - 4Ciii 13 - 4Bii 18 - 4Cii 

HP-DW-EVR 53DF10 53DF1000 53DF10 53DF1000 53DF10 53DF1000 

HP-WW-EVR

SRV-DW-EVR 51DF10 51DF1000 51DF10 51DF1000 51DF10 51DF1000 

SRV-WW-EVR C107DF10 C107DF1000 C107DF10 C107DF1000 

SRV-DW-IVR 

SRV-OP-IVR 

SRV-WW-IVR 

HP-DW-LMT 2DF10 2DF1000 2DF10 2DF1000 2DF10 2DF1000 

HP-OP-LMT 2DF10 2DF1000 2DF10 2DF1000 2DF10 2DF1000 

HP-WW-LMT 2DF10 2DF1000 2DF10 2DF1000 2DF10 2DF1000 

SRV-DW-LMT C104DF10 C104DF1000 C104DF10 C104DF1000 C104DF10 C104DF1000 

SRV-OP-LMT C104DF10 C104DF1000 C104DF10 C104DF1000 C104DF10 C104DF1000 

SRV-WW-LMT C104DF10 C104DF1000 C104DF10 C104DF1000 C104DF10 C104DF1000 
2-23

Table 2-6 Core-damage frequencies by site for Wetwell-First strategy Site EPS Class SBO Coping Time (h)

Internal Events Seismic Total 

ELAP Frequency (/y) CDF (/y)CCDPELAP Frequency (/y) CDF (/y)CCDPELAP Frequency (/y) CDF (/y)CCDP

Browns Ferry 4 4 4.1E-07 1.5E-07 36% 2.2E-05 1.0E-05 46% 2.3E-05 1.0E-05 45%

Brunswick 2 4 6.9E-06 2.5E-06 36% 1.1E-05 4.7E-06 44% 1.8E-05 7.2E-06 41%

Cooper 2 4 6.9E-06 2.5E-06 36% 4.9E-06 2.2E-06 45% 1.2E-05 4.7E-06 40%

Duane Arnold 2 4 6.9E-06 2.5E-06 36% 2.4E-06 1.1E-06 47% 9.2E-06 3.6E-06 39%

Fermi 4 4 4.1E-07 1.5E-07 36% 8.8E-06 4.1E-06 47% 9.2E-06 4.3E-06 46%

FitzPatrick 4 4 4.1E-07 1.5E-07 36% 3.5E-06 1.6E-06 46% 3.9E-06 1.7E-06 45%

Hatch 3 4 3.4E-06 1.2E-06 36% 9.6E-06 4.4E-06 46% 1.3E-05 5.6E-06 43%

Hope Creek 3 4 3.4E-06 1.2E-06 36% 7.6E-06 3.4E-06 44% 1.1E-05 4.6E-06 42%

Monticello 2 4 6.9E-06 2.5E-06 36% 6.7E-06 3.2E-06 47% 1.4E-05 5.6E-06 42%

Peach Bottom 3 8 2.0E-06 7.4E-07 36% 3.8E-05 1.9E-05 50% 4.0E-05 2.0E-05 50%

Pilgrim 2 8 4.1E-06 1.5E-06 36% 7.0E-05 3.4E-05 49% 7.4E-05 3.6E-05 48%

Quad Cities 4 4 4.1E-07 1.5E-07 36% 7.0E-06 3.3E-06 47% 7.4E-06 3.4E-06 47%

Fleet Average 3.5E-06 1.3E-06 36% 1.6E-05 7.6E-06 48% 1.9E-05 8.9E-06 46% 

Table 2-7 Core-damage frequencies by site for Drywell-First (passive) strategy. Site EPS Class SBO Coping Time (h)

Internal Events Seismic Total 

ELAP Frequency (/y) CDF (/y)CCDPELAP Frequency (/y) CDF (/y)CCDPELAP Frequency (/y) CDF (/y)CCDP

Browns Ferry 4 4 4.1E-07 1.4E-07 34% 2.2E-05 9.6E-06 43% 2.3E-05 9.8E-06 43%

Brunswick 2 4 6.9E-06 2.3E-06 34% 1.1E-05 4.5E-06 41% 1.8E-05 6.8E-06 38%

Cooper 2 4 6.9E-06 2.3E-06 34% 4.9E-06 2.0E-06 42% 1.2E-05 4.4E-06 37%

Duane Arnold 2 4 6.9E-06 2.3E-06 34% 2.4E-06 1.0E-06 44% 9.2E-06 3.4E-06 36%

Fermi 4 4 4.1E-07 1.4E-07 34% 8.8E-06 3.9E-06 44% 9.2E-06 4.0E-06 44%

FitzPatrick 4 4 4.1E-07 1.4E-07 34% 3.5E-06 1.5E-06 43% 3.9E-06 1.6E-06 42%

Hatch 3 4 3.4E-06 1.2E-06 34% 9.6E-06 4.1E-06 43% 1.3E-05 5.3E-06 40%

Hope Creek 3 4 3.4E-06 1.2E-06 34% 7.6E-06 3.2E-06 42% 1.1E-05 4.3E-06 39%

Monticello 2 4 6.9E-06 2.3E-06 34% 6.7E-06 3.0E-06 44% 1.4E-05 5.3E-06 39%

Peach Bottom 3 8 2.0E-06 6.9E-07 34% 3.8E-05 1.8E-05 48% 4.0E-05 1.9E-05 47%

Pilgrim 2 8 4.1E-06 1.4E-06 34% 7.0E-05 3.2E-05 46% 7.4E-05 3.4E-05 46%

Quad Cities 4 4 4.1E-07 1.4E-07 34% 7.0E-06 3.1E-06 45% 7.4E-06 3.3E-06 44%

Fleet Average 3.5E-06 1.2E-06 34% 1.6E-05 7.2E-06 45% 1.9E-05 8.4E-06 43% 
2-24

Table 2-8 Core-damage frequencies by site for Drywell-First (manual) strategy

As shown in Table 2-6, Table 2-7 and Table 2-8, the contribution of ELAPs initiated by seismic 

events is generally greater than the contribution of ELAPs initiated by internal events. However, 

this observation is reversed for plants located in areas with low seismicity with only two onsite 

emergency power sources (EPS Class) and 4-hour SBO coping times. Figure 2-3 shows the 

contribution to fleet-average ELAP frequency and CDF by seismic bin. In general, the contribution 

from ELAPs initiated by seismic events is due to earthquakes whose ground motion exceeds the 

plant design basis (the safe shutdown earthquake). Site EPS Class SBO Coping Time (h)

Internal Events Seismic Total 

ELAP Frequency (/y) CDF (/y)CCDPELAP Frequency (/y) CDF (/y)CCDPELAP Frequency (/y) CDF (/y)CCDP

Browns Ferry 4 4 4.1E-07 1.5E-07 36% 2.2E-05 1.0E-05 46% 2.3E-05 1.0E-05 45%

Brunswick 2 4 6.9E-06 2.5E-06 36% 1.1E-05 4.7E-06 44% 1.8E-05 7.2E-06 41%

Cooper 2 4 6.9E-06 2.5E-06 36% 4.9E-06 2.2E-06 45% 1.2E-05 4.7E-06 40%

Duane Arnold 2 4 6.9E-06 2.5E-06 36% 2.4E-06 1.1E-06 47% 9.2E-06 3.6E-06 39%

Fermi 4 4 4.1E-07 1.5E-07 36% 8.8E-06 4.1E-06 47% 9.2E-06 4.3E-06 46%

FitzPatrick 4 4 4.1E-07 1.5E-07 36% 3.5E-06 1.6E-06 46% 3.9E-06 1.7E-06 45%

Hatch 3 4 3.4E-06 1.2E-06 36% 9.6E-06 4.4E-06 46% 1.3E-05 5.6E-06 43%

Hope Creek 3 4 3.4E-06 1.2E-06 36% 7.6E-06 3.4E-06 44% 1.1E-05 4.6E-06 42%

Monticello 2 4 6.9E-06 2.5E-06 36% 6.7E-06 3.2E-06 47% 1.4E-05 5.6E-06 42%

Peach Bottom 3 8 2.0E-06 7.4E-07 36% 3.8E-05 1.9E-05 50% 4.0E-05 2.0E-05 50%

Pilgrim 2 8 4.1E-06 1.5E-06 36% 7.0E-05 3.4E-05 49% 7.4E-05 3.6E-05 48%

Quad Cities 4 4 4.1E-07 1.5E-07 36% 7.0E-06 3.3E-06 47% 7.4E-06 3.4E-06 47%

Fleet Average 3.5E-06 1.3E-06 36% 1.6E-05 7.6E-06 48% 1.9E-05 8.9E-06 46% 


)LJXUH &RQWULEXWLRQVWR(/$3IUHTXHQF\DQGFRUHGDPDJHIUHTXHQF\

7DEOHLGHQWLILHVWKHVLJQLILFDQWFRQWULEXWRUVWR&')DVGHWHUPLQHGE\LPSRUWDQFHDQDO\VLV

&RQVLVWHQWZLWK5HJXODWRU\*XLGHDQHYHQWLVDVLJQLILFDQWFRQWULEXWRULILWKDVDULVN

DFKLHYHPHQWZRUWK5$:JUHDWHUWKHQRUD)XVVHOO9HVHO\)9LPSRUWDQFHPHDVXUHJUHDWHU

WKDQ)LJXUHSUHVHQWVWKHLQIRUPDWLRQLQ7DEOHLQJUDSKLFDOIRUP7KHPRVWLPSRUWDQW

FRQWULEXWRUVWR&')DUHVHLVPLFIDLOXUHVRIWKHVWDWLRQEDWWHULHVRUGFVZLWFKJHDUVHLVPLFIDLOXUHV

RIWKHHPHUJHQF\GLHVHOJHQHUDWRUV('*VDQGWKHLUVXSSRUWLQJV\VWHPVUDQGRPIDLOXUHVRIWKH

SRUWDEOH)/(;SXPSIDLOXUHWRVWDUWDQGIDLOXUHWRUXQDQGUDQGRPIDLOXUHVRIWKH5&,&SXPS

IDLOXUHWRVWDUW

)DLOXUHVRIGFHTXLSPHQWEDWWHULHVVZLWFKJHDUDUHLPSRUWDQWEHFDXVHLWLVXVHGWRDFWXDWHWKH

('*V5&,&WKH659VDQGFRQWDLQPHQWYHQWV7KH&'(76DVVXPHWKDWIDLOXUHVRI5&,&WKH

659VDQGWKHFRQWDLQPHQWYHQWVFDQEHUHFRYHUHGE\ORFDOPDQXDORSHUDWRUDFWLRQV7KHVH

UHFRYHU\DFWLRQVZKLFKQHHGWRRFFXULQDGGLWLRQWRDOLJQLQJWKH)/(;SXPSWDNHSODFHDWGLYHUVH

SODQWORFDWLRQV7KHDFFLGHQWVHTXHQFHDQDO\VLVDVVXPHGWKDWWKHSODQWLVDGHTXDWHO\VWDIIHGWR

DFKLHYHDOORIWKHVHRSHUDWRUVDQGGLGQRWH[SORUHUHVRXUFHFRQVWUDLQWV)DLOXUHRISODQWSHUVRQQHO

WRVXFFHVVIXOO\SHUIRUPWKHVHUHFRYHU\DFWLRQVZRXOGFRQWULEXWHWRZDUGDQLQFUHDVHRQFRUH

GDPDJHIUHTXHQF\

 
2-26

Table 2-9 Importance measures with respect to core-damage frequency 


)LJXUH ,GHQWLILFDWLRQRIVLJQLILFDQWHYHQWVZLWKUHVSHFWWRFRUHGDPDJHIUHTXHQF\
2-28

APPENDIX G: provides the complete list of PDS frequencies, in alphabetic order, that were input 

to the APETs. Table 2-10, Table 2-11 and Table 2-12 list the significant PDS for each CDET in 

descending order. A significant PDS is one of the set of PDSs that, when ranked, compose 95% 

of the CDF or that individually contributes more than 1% to the CDF. Review of these tables 

shows that failures that occur during FLEX Phase 1 (before the portable FLEX pump can be 

aligned) contribute about 30% of the total CDF. 

Table 2-10 Significant plant damage states – Wetwell First strategy

Rank 

Wetwell First Strategy 

Plant Damage State Frequency (/y) Contribution 

(%) Cumulative (%) 

1 E-LP-IS-ST-XX 1.01E-06 11.4 11.4

2 E-LP-IS-XX-XX 9.81E-07 11.0 22.4

3 L-MP-IS-OK-F 6.74E-07 7.6 30.0 

4 M-LP-IS-OK-H 5.70E-07 6.4 36.4 

5 L-MP-IS-OK-H 4.44E-07 5.0 41.3 

6 E-HP-IS-ST-XX 4.34E-07 4.9 46.2 

7 M-MP-IS-LT-XX 3.67E-07 4.1 50.3 

8 M-LP-IS-LT-H 3.29E-07 3.7 54.0 

9 M-LP-IS-ST-H 2.94E-07 3.3 57.3 

10 M-MP-IS-LT-H 2.34E-07 2.6 60.0 

11 M-LP-IS-OK-F 2.00E-07 2.2 62.2 

12 L-MP-IS-LT-F 1.67E-07 1.9 64.1 

13 M-MP-IS-ST-XX 1.51E-07 1.7 65.8 

14 M-MP-IS-LT-OK 1.39E-07 1.6 67.3 

15 M-LP-IS-LT-F 1.16E-07 1.3 68.6 

16 L-MP-IS-LT-H 1.10E-07 1.2 69.9 

17 E-HP-IS-XX-XX 1.09E-07 1.2 71.1 

18 M-LP-WW-LT-H 1.09E-07 1.2 72.3 

19 M-LP-IS-ST-F 1.03E-07 1.2 73.5 

20 M-LP-IS-ST-XX 9.97E-08 1.1 74.6 

21 M-LP-WW-ST-H 9.68E-08 1.1 75.7 

22 M-MP-IS-ST-H 9.61E-08 1.1 76.8 

23 L-MP-IS-OK-OK 8.79E-08 1.0 77.7 

24 M-MP-IS-LT-F 8.22E-08 0.9 78.7 

25 M-MP-WW-LT-H 7.72E-08 0.9 79.5 

26 M-LP-IS-LT-XX 6.86E-08 0.8 80.3 

27 L-MP-IS-ST-F 6.84E-08 0.8 81.1 

28 L-MP-WW-OK-F 6.81E-08 0.8 81.8 

29 M-HP-IS-ST-XX 6.77E-08 0.8 82.6 

30 L-MP-IS-LT-OK 6.52E-08 0.7 83.3 

31 L-HP-IS-OK-F 6.51E-08 0.7 84.1 

32 M-MP-IS-OK-XX 6.42E-08 0.7 84.8 

33 M-LP-WW-OK-H 5.75E-08 0.6 85.4 

34 M-MP-IS-ST-OK 5.71E-08 0.6 86.1 

35 L-MP-WW-LT-F 5.50E-08 0.6 86.7 

36 M-MP-WW-LT-OK 4.59E-08 0.5 87.2 

37 L-MP-IS-ST-H 4.50E-08 0.5 87.7 

38 L-MP-WW-OK-H 4.48E-08 0.5 88.2 

39 M-HP-IS-ST-H 4.32E-08 0.5 88.7 

40 L-HP-IS-OK-H 4.28E-08 0.5 89.2 

41 L-HP-IS-ST-F 4.21E-08 0.5 89.7 

42 M-LP-WW-LT-F 3.81E-08 0.4 90.1 

43 L-MP-WW-LT-H 3.62E-08 0.4 90.5 
2-29

Rank 

Wetwell First Strategy 

Plant Damage State Frequency (/y) Contribution 

(%) Cumulative (%) 

44 M-LP-WW-ST-F 3.40E-08 0.4 90.9 

45 M-MP-IS-ST-F 3.37E-08 0.4 91.2 

46 M-HP-IS-LT-XX 3.33E-08 0.4 91.6 

47 M-LP-DW-LT-H 3.26E-08 0.4 92.0 

48 M-MP-WW-ST-H 3.17E-08 0.4 92.3 

49 M-LP-DW-ST-H 2.90E-08 0.3 92.7 

50 L-HP-IS-ST-H 2.77E-08 0.3 93.0 

51 M-MP-WW-LT-F 2.71E-08 0.3 93.3 

52 L-MP-IS-ST-OK 2.68E-08 0.3 93.6 

53 M-HP-IS-ST-OK 2.57E-08 0.3 93.9 

54 M-MP-DW-LT-H 2.32E-08 0.3 94.1 

55 L-MP-WW-ST-F 2.26E-08 0.3 94.4 

56 L-MP-WW-LT-OK 2.15E-08 0.2 94.6 

57 M-HP-IS-LT-H 2.12E-08 0.2 94.9 

58 L-HP-IS-LT-F 2.07E-08 0.2 95.1 

Table 2-11 Significant plant damage states – Drywell-First (passive) strategy. 

Rank 

Wetwell First Strategy 

Plant Damage State Frequency (/y) Contribution 

(%) Cumulative (%) 

1 E-LP-IS-ST-XX 1.01E-06 12.1 12.1

2 E-LP-IS-XX-XX 9.81E-07 11.7 23.7

3 L-MP-DW-OK-F 7.49E-07 8.9 32.7 

4 M-LP-DW-OK-H 6.33E-07 7.5 40.2 

5 M-LP-DW-LT-H 5.12E-07 6.1 46.3 

6 L-MP-DW-OK-H 4.93E-07 5.9 52.2 

7 M-LP-DW-ST-H 4.56E-07 5.4 57.6 

8 E-HP-IS-ST-XX 4.34E-07 5.2 62.8 

9 M-MP-DW-LT-H 3.64E-07 4.3 67.1 

10 L-MP-DW-LT-F 2.59E-07 3.1 70.2 

11 M-LP-DW-OK-F 2.22E-07 2.6 72.9 

12 M-MP-DW-LT-OK 2.16E-07 2.6 75.4 

13 M-LP-DW-LT-F 1.80E-07 2.1 77.6 

14 L-MP-DW-LT-H 1.70E-07 2.0 79.6 

15 M-LP-DW-ST-F 1.60E-07 1.9 81.5 

16 M-MP-DW-ST-H 1.49E-07 1.8 83.3 

17 M-MP-DW-LT-F 1.28E-07 1.5 84.8 

18 E-HP-IS-XX-XX 1.09E-07 1.3 86.1 

19 L-MP-DW-ST-F 1.06E-07 1.3 87.4 

20 L-MP-DW-LT-OK 1.01E-07 1.2 88.6 

21 L-MP-DW-OK-OK 9.77E-08 1.2 89.8 

22 M-MP-DW-ST-OK 8.88E-08 1.1 90.8 

23 L-HP-DW-OK-F 7.24E-08 0.9 91.7 

24 L-MP-DW-ST-H 6.99E-08 0.8 92.5 

25 M-HP-DW-ST-H 6.72E-08 0.8 93.3 

26 L-HP-DW-ST-F 6.55E-08 0.8 94.1 

27 M-MP-DW-ST-F 5.24E-08 0.6 94.7 

28 L-HP-DW-OK-H 4.76E-08 0.6 95.3 
2-30

Table 2-12 Significant plant damage states – Drywell-First (manual) strategy. 

Rank 

Wetwell First Strategy 

Plant Damage State Frequency (/y) Contribution 

(%) Cumulative (%) 

1 E-LP-IS-ST-XX 1.01E-06 11.4 11.4 

2 E-LP-IS-XX-XX 9.81E-07 11.0 22.4 

3 L-MP-IS-OK-F 6.74E-07 7.6 30.0 

4 M-LP-IS-OK-H 5.70E-07 6.4 36.4 

5 L-MP-IS-OK-H 4.44E-07 5.0 41.3 

6 E-HP-IS-ST-XX 4.34E-07 4.9 46.2 

7 M-MP-IS-LT-XX 3.67E-07 4.1 50.3 

8 M-LP-IS-LT-H 3.29E-07 3.7 54.0 

9 M-LP-IS-ST-H 2.94E-07 3.3 57.3 

10 M-MP-IS-LT-H 2.34E-07 2.6 60.0 

11 M-LP-IS-OK-F 2.00E-07 2.2 62.2 

12 L-MP-IS-LT-F 1.67E-07 1.9 64.1 

13 M-MP-IS-ST-XX 1.51E-07 1.7 65.8 

14 M-MP-IS-LT-OK 1.39E-07 1.6 67.3 

15 M-LP-IS-LT-F 1.16E-07 1.3 68.6 

16 L-MP-IS-LT-H 1.10E-07 1.2 69.9 

17 E-HP-IS-XX-XX 1.09E-07 1.2 71.1 

18 M-LP-DW-LT-H 1.09E-07 1.2 72.3 

19 M-LP-IS-ST-F 1.03E-07 1.2 73.5 

20 M-LP-IS-ST-XX 9.97E-08 1.1 74.6 

21 M-LP-DW-ST-H 9.68E-08 1.1 75.7 

22 M-MP-IS-ST-H 9.61E-08 1.1 76.8 

23 L-MP-IS-OK-OK 8.79E-08 1.0 77.7 

24 M-MP-IS-LT-F 8.22E-08 0.9 78.7 

25 M-MP-DW-LT-H 7.72E-08 0.9 79.5 

26 M-LP-IS-LT-XX 6.86E-08 0.8 80.3 

27 L-MP-IS-ST-F 6.84E-08 0.8 81.1 

28 L-MP-DW-OK-F 6.81E-08 0.8 81.8 

29 M-HP-IS-ST-XX 6.77E-08 0.8 82.6 

30 L-MP-IS-LT-OK 6.52E-08 0.7 83.3 

31 L-HP-IS-OK-F 6.51E-08 0.7 84.1 

32 M-MP-IS-OK-XX 6.42E-08 0.7 84.8 

33 M-LP-DW-OK-H 5.75E-08 0.6 85.4 

34 M-MP-IS-ST-OK 5.71E-08 0.6 86.1 

35 L-MP-DW-LT-F 5.50E-08 0.6 86.7 

36 M-MP-DW-LT-OK 4.59E-08 0.5 87.2 

37 L-MP-IS-ST-H 4.50E-08 0.5 87.7 

38 L-MP-DW-OK-H 4.48E-08 0.5 88.2 

39 M-HP-IS-ST-H 4.32E-08 0.5 88.7 

40 L-HP-IS-OK-H 4.28E-08 0.5 89.2 

41 L-HP-IS-ST-F 4.21E-08 0.5 89.7 

42 M-LP-DW-LT-F 3.81E-08 0.4 90.1 

43 L-MP-DW-LT-H 3.62E-08 0.4 90.5 

44 M-LP-DW-ST-F 3.40E-08 0.4 90.9 

45 M-MP-IS-ST-F 3.37E-08 0.4 91.2 

46 M-HP-IS-LT-XX 3.33E-08 0.4 91.6 

47 M-LP-WW-LT-H 3.26E-08 0.4 92.0 

48 M-MP-DW-ST-H 3.17E-08 0.4 92.3 

49 M-LP-WW-ST-H 2.90E-08 0.3 92.7 

50 L-HP-IS-ST-H 2.77E-08 0.3 93.0 

51 M-MP-DW-LT-F 2.71E-08 0.3 93.3 

52 L-MP-IS-ST-OK 2.68E-08 0.3 93.6 

53 M-HP-IS-ST-OK 2.57E-08 0.3 93.9 
2-31

2.3.2 Results from the APET Quantification

Table 2-13 summarizes the results of the APET quantification in terms of containment failure 

mode probabilities, which provides insight into the efficacy of the CPRR strategies. The capability 

to vent the containment during a severe accident, as required by Order EA-13-109, limits the 

probability of containment structural failure due to overpressurization to a few percent. However, it 

is noted that operation of the containment venting system during a severe accident also results in 

a release (albeit, controlled) of fission products to the environment. 

Rank 

Wetwell First Strategy 

Plant Damage State Frequency (/y) Contribution 

(%) Cumulative (%) 

54 M-MP-WW-LT-H 2.32E-08 0.3 94.1 

55 L-MP-DW-ST-F 2.26E-08 0.3 94.4 

56 L-MP-DW-LT-OK 2.15E-08 0.2 94.6 

57 M-HP-IS-LT-H 2.12E-08 0.2 94.9 

58 L-HP-IS-LT-F 2.07E-08 0.2 95.1 

The provision of post-accident water injection capability noticeably reduces the probability of 

containment failure due to liner melt-through. The largest reductions are associated with APET-2, 

which applies to regulatory analysis sub-alternatives that utilize RPV injection (as opposed to 

direct DW injection). Sub-alternatives involving post-accident injection to the RPV provide the 

capability to arrest a severe accident before vessel breach occurs. 

Table 2-13 Containment failure mode probabilities

Description 

Accident Progression Event Trees and 

Associated Regulatory Analysis Sub-Alternatives 

APET-1 

1 - 1 

2 - 2A

APET-2 

3 - 3A

5 - 4Ai(1) 

7 - 4Aii(1) 

9 - 4Aiii(1) 

11 - 4Bi(1) 

16 - 4Ci(1) 

APET-3 

4 - 3B

6 - 4Ai(2) 

8 - 4Aii(2) 

10 - 4Aiii(2) 

12 - 4Bi(2) 

17 - 4Ci(2) 

APET-4 

15 - 4Biv 

20 - 4Civ 

APET-5 

14 - 4Biii 

19 - 4Ciii 

APET-6 

13 - 4Bii 

18 - 4Cii 

Containment Pressure Control 

Drywell Venting 15% 15% 21% 99% 96% 85%

Wetwell Venting 81% 81% 77% 1% 3% 13%

Overpressurization Failure 4% 4% 3% 0.05% 0.1% 3%

Core Debris Location 

In-Vessel Retention 0% 42% 0% 0% 0% 0%

Ex-Vessel Retention 0% 13% 42% 40% 42% 42%

Liner Melt-Through 100% 45% 58% 60% 58% 58%

Conditional Containment 

Failure Probability 100% 47% 58% 60% 58% 58%

Note: Containment overpressurization failure and liner melt-through, taken together, contain cross-product terms 

which are eliminated in arriving at the total conditional containment failure probabilitiy. 

Table 2-14 lists the significant RCs for each APET, which were input to the risk calculations made 

for each regulatory analysis sub-alternative.
2-32

Table 2-14 Frequencies of significant release categories

1. U.S. Nuclear Regulatory Commission, SECY-15-0085, “Evaluation of the Containment

Protection & Release Reduction for Mark I and Mark II Boiling Water Reactors

Rulemaking Activities (10 CFR Part 50) (RIN-3150-AJ26),” June 18, 2015, ADAMS

Accession No. ML15022A218.

2. U.S. Nuclear Regulatory Commission, SECY-12-0157, “Consideration of Additional

Requirements for Containment Venting Systems for Boiling Water Reactors with Mark I

and Mark II Containments (REDACTED VERSION),” November 26, 2012, ADAMS

Accession No. ML12345A030.

3. U.S. Nuclear Regulatory Commission, Order EA-12-050, “Issuance of Order to Modify

Licenses with Regard to Reliable Hardened Containment Vents,” March 12, 2012,

ADAMS Accession No. ML12054A694.

4. U.S. Nuclear Regulatory Commission, Order EA-12-049, “Issuance of Order to Modify

Licenses with Regard to Requirements for Mitigation Strategies for Beyond-DesignBasis External Events”, March 12, 2012, ADAMS Accession No. ML 12054A736.

Index 

Release 

Cateogry 

Accident Progression Event Trees and 

Associated Regulatory Analysis Sub-Alternatives 

APET-1 

1 - 1 

2 - 2A

APET-2 

3 - 3A

5 - 4Ai(1) 

7 - 4Aii(1) 

9 - 4Aiii(1) 

11 - 4Bi(1) 

16 - 4Ci(1) 

APET-3 

4 - 3B

6 - 4Ai(2) 

8 - 4Aii(2) 

10 - 

4Aiii(2) 

12 - 4Bi(2) 

17 - 4Ci(2) 

APET-4 

15 - 4Biv 

20 - 4Civ 

APET-5 

14 - 4Biii 

19 - 4Ciii 

APET-6 

13 - 4Bii 

18 - 4Cii 

1 HP-DW-EVR 1.7E-07 2.1E-07 2.0E-07

2 HP-DW-LMT 1.2E-07 4.0E-07 4.0E-07 3.3E-07

3 HP-WW-EVR 1.8E-07

4 HP-WW-LMT 2.8E-07 1.6E-07 2.9E-07

5 SRV-DW-EVR 1.7E-07 5.6E-07 3.1E-06 3.4E-06 3.1E-06

6 SRV-DW-IVR 5.7E-07

7 SRV-DW-LMT 1.3E-06 5.5E-07 1.1E-06 4.5E-06 4.5E-06 3.8E-06

8 SRV-OP-IVR 1.6E-07

9 SRV-OP-LMT 3.5E-07 1.2E-07 1.6E-07 1.6E-07

10 SRV-WW-EVR 9.1E-07 2.9E-06 9.9E-08 3.6E-07

11 SRV-WW-IVR 3.0E-06

12 SRV-WW-LMT 6.9E-06 3.1E-06 3.4E-06 1.7E-07 6.8E-07

Total of Significant RCs 8.8E-06 8.8E-06 8.7E-06 8.2E-06 8.7E-06 8.6E-06

Core-Damage 

Frequency 

8.9E-06 8.9E-06 8.9E-06 8.4E-06 8.9E-06 8.9E-06

Fraction of CDF 

Captured in the APET 99% 99% 98% 98% 98% 97% 

2.4 References for Chapter 2
2-33

5. U.S. Nuclear Regulatory Commission, Order EA-13-109, “Issuance of Order to Modify

Licenses with Regard to Reliable Hardened Containment Vents Capable of Operation

Under Severe Accident Conditions,” June 6, 2013, ADAMS Accession No.

ML13143A334.

6. U.S. Nuclear Regulatory Commission, NUREG/BR-0058, Rev. 4, “Regulatory Analysis

Guidelines of the U.S. Nuclear Regulatory Commission,” September 2004, ADAMS

Accession No. ML042820192.

7. U.S. Nuclear Regulatory Commission, NUREG/BR-0184, “Regulatory Analysis Technical

Evaluation Handbook,” October 2010, ADAMS Accession No. ML050190193.

8. U.S. Nuclear Regulatory Commission, “Recommendations for Enhancing Reactor Safety

in the 21st Century: The Near-Term Task Force Review of Insights from the Fukushima

Dai-ichi Accident,” July 11, 2011, ADAMS Accession No. ML111861807.

9. U.S. Nuclear Regulatory Commission, RG 1.200, Rev. 2, “An Approach for Determining

the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed

Activities,” March 2009, available through the NRC’s public Web site under “Regulatory

Guides” at http://www.nrc.gov/reading-rm/doc-collections/.

10. Nuclear Energy Institute, NEI 12-06, Rev. 2, “Diverse and Flexible Coping Strategies

(FLEX) Implementation Guide,” December 2015, ADAMS Accession No. ML16005A625.

11. U.S. Nuclear Regulatory Commission, NUREG/CR-6890, Vol. 1, “Reevaluation of

Station Blackout Risk at Nuclear Power Plants,” December 2005, ADAMS Accession No.

ML060200477.

12. Idaho National Laboratory, INL/EXT-15-34441, “System Study: Emergency Power

System 1998–2013,” January 2015, http://nrcoe.inel.gov/resultsdb/publicdocs/

SystemStudies/eps-SPAR-unreliability-external-2013.pdf

13. Tennessee Valley Authority, “Tennessee Valley Authority’s Seismic Hazard and

Screening Report (CEUS Sites), Response to NRC Request for Information Pursuant to

10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review

of Insights from the Fukushima Dai-ichi Accident,” March 31, 2014, ADAMS Accession

No. ML14098A478.

14. Duke Energy Progress, Inc., “Seismic Hazard and Screening Report (CEUS Sites),

Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding the

Seismic Aspects of Recommendation 2.1 of the Near-Term Task Force Review of

Insights from the Fukushima Dai-ichi Accident,” March 31, 2014, ADAMS Accession No.

ML14106A461.

15. Nebraska Public Power District, “Nebraska Public Power District's Seismic Hazard and

Screening Report (CEUS Sites) - Response to NRC Request for Information Pursuant to

10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review

of Insights from the Fukushima Dai-ichi Accident,” March 31, 2014, ADAMS Accession

Nos. ML14094A040 and ML14094A042.
2-34

16. NextEra Energy Duane Arnold, LLC, “NextEra Energy Duane Arnold, LLC Seismic

Hazard and Screening Report (CEUS Sites), Response to NRC Request for Information

Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task

Force Review of Insights from the Fukushima Dai-ichi Accident,” March 28, 2014,

ADAMS Accession No. ML14092A331.

17. DTE Electric Company, “DTE Electric Company's Seismic Hazard and Screening

Report, Response to NRC Request for Information Pursuant to 10 CFR 50.54(f)

Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from

the Fukushima Dai-ichi Accident,” March 31, 2014, ADAMS Accession No.

ML14090A326.

18. Entergy Nuclear Northeast, “Entergy Seismic Hazard and Screening Report (CEUS

Sites), Response NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding

Recommendation 2.1 of the Near-Term Task Force Review of Insights from the

Fukushima Dai-ichi Accident,” March 31, 2014, ADAMS Accession No. ML14090A243.

19. Southern Nuclear Operating Company, Inc., Edwin I. Hatch Nuclear Plant - Units 1 and

2, Seismic Hazard and Screening Report for CEUS Sites,” March 31, 2014, ADAMS

Accession No. ML14092A017.

20. PSEG Nuclear LLC, “PSEG Nuclear LLC's Seismic Hazard and Screening Report

(CEUS Sites) Response to NRC Request for Information Pursuant to 10 CFR 50.54(f)

Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from

the Fukushima Dai-ichi Accident - Hope Creek Generating Station,” March 28, 2014,

ADAMS Accession No. ML14087A436.

21. Exel Energy, “MNGP Seismic Hazard and Screening Report (CEUS Sites), Response to

NRC Request for Information Pursuant to Title 10 of the Code of Federal Regulations

50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of

Insights from the Fukushima Dai-ichi Accident,” May 14, 2014, ADAMS Accession No.

ML14136A288.

22. Exelon Generation Company LLC, “Seismic Hazard and Screening Report (Central and

Eastern United States (CEUS) Sites), Response to NRC Request for Information

Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task

Force Review of Insights from the Fukushima Dai-ichi Accident,” March 31, 2014,

ADAMS Accession No. ML14090A247.

23. Entergy Nuclear Operations, Inc., “Entergy's Seismic Hazard and Screening Report

(CEUS Sites), Response to NRC Request for Information Pursuant to 10 CFR 50.54(f)

Regarding the Seismic Aspects of Recommendation 2.1 of the Near-Term Task Force

Review of Insights from the Fukushima Dai-ichi Accident,” March 31, 2014, ADAMS

Accession No. ML14092A023.

24. Exelon Generation Company LLC, “Exelon Generation Company, LLC, Seismic Hazard

and Screening Report (Central and Eastern United States (CEUS) Sites), Response to

NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation

2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi

Accident,” March 31, 2014, ADAMS Accession No. ML14090A526.
2-35

25. U.S. Nuclear Regulatory Commission, “Request for Information Pursuant to Title 10 of

the Code of Federal Regulations 50.54(F) Regarding Recommendations 2.1, 2.3, and

9.3, of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi

Accident,” March 12, 2012, ADAMS Accession No. ML12056A046.

26. Electric Power Research Institute, EPRI Report 1025287, “Seismic Evaluation Guidance:

Screening, Prioritization and Implementation Details (SPID) for the Resolution of

Fukushima Near-Term Task Force Recommendation 2.1: Seismic,” November 2012,

ADAMS Accession No. ML12333A170.

27. Idaho National Laboratory, “Summary of SPAR Component Unreliability Data and

Results: 2010 Parameter Estimation Update,” http://nrcoe.inel.gov/resultsdb/publicdocs/

AvgPerf/ComponentUR2010.pdf

28. U.S. Nuclear Regulatory Commission, NUREG/CR-7155 (draft), “State-of-the-Art

Reactor Consequence Analyses Project, Uncertainty Analysis of the Unmitigated LongTerm Station Blackout of the Peach Bottom Atomic Power Station,” August 2013,

ADAMS Accession No. ML13189A145.

3-1

3 ACCIDENT PROGRESSION ANALYSIS

This report documents analysis of selected accident scenarios in representative boiling-water

reactor (BWR) plants with Mark I and Mark II containments. This supports the staff’s ongoing

effort to address the Near-Term Task Force (NTTF) recommendation related to the containment

venting issue [1]. Specifically, the work reported herein was performed in support of the ongoing 

Containment Protection and Release reduction (CPRR) rulemaking activities in response to the 

staff requirement memorandum for SECY-12-0157 [2], using the U.S. Nuclear Regulatory

Commission (NRC) severe accident analysis code MELCOR [3]. This former reference 

documents staff’s previous analytical work of a similar nature, which formed the technical basis of

the SECY.

The scope of MELCOR analysis falls broadly into two categories: (1) reactor systems and 

containment thermal-hydraulics under severe accident conditions; and (2) assessment of source

terms, i.e., timing and magnitude of fission products releases to the environment. The outcome of

MELCOR analysis in the first category includes, for example, containment temperature and 

pressure signatures and hydrogen distribution in the containment, reactor building, vent line, etc. – 

all indicative of the state of containment vulnerability under severe accident conditions. These 

quantities provide needed information to assess containment integrity and also provide technical

basis for developing staff guidance; for example, for the severe accident capable hardened vent

order EA-13-109 [4]. The outcome of MELCOR analysis in the second category is environmental

release estimates that are used to calculate health consequence and offsite property damage 

assessment using the MELCOR Accident Consequence Code System, MACCS [5], discussed in 

Chapter 4. The MELCOR/MACCS results, along with the results of the accident sequence 

analysis, are used to estimate the relative public health risk reduction associated with the various 

accident prevention and mitigation strategies. 

The selection of accident scenarios considered for MELCOR analyses is informed by the recent

state-of-the-art reactor consequence analysis or SOARCA [6], the Fukushima study [7], and also 

by the previous analytical work documented in SECY-12-0157. Specifically, the accident

scenarios selected for MELCOR/MACCS analyses relate to extended loss of alternating-current

power (ELAP). The ELAP results in a loss of offsite power (LOOP), failure of onsite power, and 

failure of the grid. All systems dependent on alternating current (AC) power are unavailable. The 

turbine-driven reactor core injection cooling (RCIC) system (equivalently, the isolation condenser

in some plants) is the only system assumed available. It is assumed that the high pressure 

coolant injection or high pressure core spray system is not available.

For various accident progression events (defined in in terms of accident progression event tree or

APET), MELCOR cases were run simulating different possible outcomes or plant damage states

(e.g., containment failure by overpressurization, drywell liner melt-through, main steam line 

rupture). Consideration was given to various preventative and mitigative measures and how these 

influence the failure modes.

The MELCOR calculations, described in considerable detail in the rest of this Chapter, are 

deterministic in nature. The calculations produce point estimates of the quantities of interest (e.g.,

radionuclide release fractions). There are phenomenological uncertainties in the code models,

and, as a result, the predicted point estimates also have some uncertainties. With regard to the 

containment venting issue, pertinent uncertainties are discussed in a latter section of this chapter. 

MELCOR BWR models are described in considerable detail in Section 3.1 . The representative 

Mark I plant model is based on the SOARCA study referenced above, and the insights from the 
3-2

accident progression results (e.g., mode of containment failure or venting) are used in the risk

assessment of other Mark I containments as well. The representative Mark II plant model is based 

on a LaSalle-type configuration. Scoping analysis was done to examine the response of the plant

for other Mark II configurations. Section 3.2 discusses the basis of the accident scenario 

development and formulation of the MELCOR run matrix. Results of MELCOR analysis are 

discussed in detail in Section 3.3 for both Mark I and Mark II representative plants. Section 3.4

provides a discussion on uncertainties and the implications on the results. A summary of

important results and the conclusions are documented in Section 3.5 . 

3.1 MELCOR BWR Models

The following sections detail the representative Mark I and Mark II models used in the present

analyses. The goal of this section is to provide sufficient information for readers to understand the

system and models. The representative nuclear power plants for the BWR Mark I and Mark II are 

the Peach Bottom Unit 2 and LaSalle Unit 2 Power Stations.

3.1.1 Description of Mark I MELCOR Model

The BWR/4 Mark I input model described here follows the “best practice” used in the SOARCA

study and reflects the current understanding in severe accident modeling. Modification of the 

model was necessary to perform the present analyses given the inclusion of the FLEX equipment

and sequence events. The model was updated to permit the use of the most recent MELCOR 2.1 

computer code. Details of the model are provided in SECY-12-0157. Some of the major model

characteristics are given below.

3.1.1.1 Reactor Pressure Vessel and Reactor Coolant System

Figure 3-1 shows a representation of the MELCOR control volumes and flow paths for the reactor

coolant system, and Figure 3-2 provides a detailed reactor vessel nodalization comparing

MELCOR modeling features to actual vessel design. 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.

The reactor pressure vessel is modeled with seven control volumes outside of the core region:

• Lower plenum (CV320)

• Downcomer (CV310)

• Shroud dome or upper plenum (CV345)

• Steam separators (CV350)

• Steam dryers (CV355)

• Steam dome (CV360)

• Jet pumps (CV300)

Reactor vessel upper internals are modeled in detail. The steam dryer region includes all volume 

inside of the dryer skirt and the dryers from the top of the steam separators to the top of the steam 

dryers. Water stripped from steam in the separators and dryers is returned to the downcomer 

volume. 
3-3

Figure 3-1 MELCOR reactor coolant system nodalization for Mark I analysis 

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
3-4

Figure 3-2 MELCOR reactor vessel nodalization for Mark I analysis

Flow paths are designed to represent all potential fluid pathways between the control volumes 

defined above. The 9 flow paths modeled connecting reactor pressure vessel control volumes 

include flow between:

• Jet pumps and lower plenum (FL312)

• Shroud dome/upper plenum and steam separator standpipes (FL345)

• Steam separators and steam dryers (FL355)

• Steam dryers and steam dome (FL356)

• Steam dome and downcomer (FL357)

• Loop A suction flow from the downcomer to the jet pumps (FL310)

• Loop B suction flow from the downcomer to the jet pumps (FL311)

• Steam separators and downcomer (FL351)

• Steam dryers and downcomer (FL352)

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
3-5

The heat capacity and radionuclide deposition surface of a number of structures associated with 

the reactor pressure vessel are modeled via heat structures including: 

• Cylindrical portion in the lower downcomer region

• Cylindrical portion in the upper downcomer region

• Cylindrical portion adjacent to the steam dryers

• Hemispherical upper head

• Shroud baffle

• Standpipes and steam separators

• Steam dryers

• Core shroud

Depressurization of the RCS is performed with 11 flow paths (FL359-FL369) that represent the 

S/RVs, and flow path FL358, which represents the two safety valves (SV). Each valve within a 

group is represented with a differentiating pressure of 1 psi between each SRV actuation set point 

starting from the prescribed group actuation set pressure. This small differential pressure is 

sufficient to impose nearly all automatic SRV operations to a single SRV as modeled by 

MELCOR. 

3.1.1.2 Reactor Core

In MELCOR, the region tracked directly by the COR package model includes a cylindrical space 

extending axially from the inner surface of the vessel bottom head to the core top guide and 

radially from the vessel centerline to the inside surface of the core shroud. The region tracked by 

the COR package also includes the region of the lower plenum outside of the core shroud and 

below the downcomer. The core and lower plenum regions are divided into concentric radial rings 

and axial levels. A particular radial ring and a particular axial level define a core cell (node) which 

has a cell number defined with a three digit integer IJJ, where the first digit represents the radial 

ring number and the last two digits represent the axial level number. For example, core cell 

number 314 specifies a cell located in radial ring three and axial level 14. The numbering of axial 

segments begins at the bottom of the vessel. 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 particulate debris.

The MELCOR core nodalization for the current containment filtered venting study is the same as 

SOARCA analysis as shown in Figure 3-3. The entire core and lower plenum regions are divided 

into six radial rings and 17 axial segments. Axial levels 1 through 6 represent the entire lower 

plenum and the unfueled region of the core immediately above the lower core plate. Initially this 

region has no fuel and no internal heat source. However, during the core degradation phase, the 

fuel, cladding and other core components may enter the lower plenum in the form of particulate or 

molten debris by relocation from the upper core nodes. Axial node 6 represents the steel associated 

with assembly lower tie plates, fuel nose pieces and the lower core plate and its associated 

supports. Particulate debris formed by fuel, canister, and control blade failures above the lower core 

plate will be supported at this level until the lower core plate yields. Axial segments 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. Radial ring 6 represents the region in the lower 

plenum outside of the core shroud inner radius and below the downcomer region.
3-6

Core cell geometry and masses for nonfuel-related core components (e.g., control rod guide 

tubes, lower core plate, and core top guide) are obtained from a variety of references. Axial level 1 

through 5 in rings 1 through 5 contains control rod stub tubes, control rod drives, and instrument 

guide tubes. Axial level 1 includes the region from the lower head to the top of the control rod stub 

tubes. Control rod stubs are modeled as tubes with a specified inner diameter and an outer 

diameter. Control rod drives are modeled as a solid shaft with a specified diameter representative 

of a BWR Mark I design. Fifty-five instrument tubes are modeled with each one including a guide 

tube with a specified inner diameter and an outer diameter, and a central shaft with a specified 

diameter. Control rod stub/drive and instrument tubes are distributed between the rings. The

combined mass of the control rod stub tubes, control rod drives, and instrument tubes within axial 

level 1 are modeled as a stainless steel supporting structure. The surface area for this component 

is modeled as the outer surface area of the control rod stub tubes. Axial level 1 in ring 6, which is 

outside of the lower head curvature, does not contain any core components and is simply 

deactivated within the model.

Axial level 6 in rings 1 through 5 includes the fuel support pieces, lower core plate, lower core 

plate support structures and fuel assembly lower tie plates. The total mass for the fuel support 

pieces and lower core plate is distributed between the core rings based on the fraction of the area 

inside of the core shroud represented by the ring. Assembly lower tie plate mass depends on the 

type of fuel assemblies modeled, and is distributed based on the number of assemblies per ring.

The combined mass of these structures is modeled as a steel support structure representing the 

lower core plate.

All control blades are assumed to be inserted in the core region at the time of the accident. Axial 

levels 7 through 16 in rings 1 through 5 contain the control blades distributed as described in axial 

level 1. Axial level 17 in rings 1 through 5 contains the core top guide and the fuel assembly upper 

tie plates. 

Core cells within the five concentric rings modeling the active fuel region and the core top guide 

from axial levels 7 through 17 are coupled with a total of 40 hydrodynamic control volumes. Within 

each radial ring, five axially-stacked control volumes represent coolant flow through the core 

channels and five parallel (axially-stacked) control volumes represent the neighboring bypass 

regions of the core. This reflects a coupling between core cells and hydrodynamic control volumes 

within the core region.

Four distinct groups of flow paths are modeled to represent all potential flow within the core 

region. Axial core flow within the fuel assemblies is modeled with the channel flow area for each 

ring excluding flow area internal to the water rods. Axial flow paths from the lower plenum into the 

fuel assembly channel include pressure losses associated with flow through the fuel support piece 

orifices and the lower tie plate. Axial flow paths between volumes within the core region include 

friction losses for flow through fuel rods over a volume-center to volume-center length and form 

losses based on grid spacers. Axial flow from the upper fuel region control volume and the upper 

plenum includes form losses for flow through the upper tie plate. The MELCOR axial flow 

blockage model is activated for each of these flow paths. Axial bypass core flow between 

canisters and through the peripheral bypass is modeled with the bypass flow area in the core 

region, including flow area internal to the water rods.
3-7

Figure 3-3 MELCOR core nodalization for Mark I analysis

At each axial level of the core, horizontal flow paths model the possibility of coolant cross-flow 

between channel and bypass areas. The open fraction for these flow paths is connected to control 

logic that monitors channel box integrity (i.e., the flow paths are closed when the channel box is 

intact and open if the channel box fails in a particular ring). In addition, coolant cross-flow between 

bypass regions is modeled by horizontal flow paths between each ring at each axial level.

The lower head is modeled as a hemisphere with an inner radius and thickness representative of 

a BWR Mark I plant. The lower head region extends to the downcomer baffle plate where it 

connects with the reactor pressure vessel. The hemispherical region of the lower head is 

represented by eight segments, and the cylindrical region of the lower head below the baffle plate

is represented by a single segment. A one-dimensional model of the stress and strain distribution 

in the lower head is applied. Heat transfer coefficients from particulate debris to the lower head 

and penetrations are modeled with a temperature-dependent control function which reflects 

conduction-based heat transfer through a frozen crust at temperatures of 2650K and below, a 

conduction enhanced heat transfer coefficient as the debris reaches the eutectic melting 

temperature of UO2 and ZrO2, and a convective heat transfer coefficient as the debris exceeds the 

eutectic melting temperature and forms a circulating molten pool. By default, the penetration 

failure model is deactivated and the lower head failures occur due to creep rupture (same 

assumption as in SOARCA).

CV514CV524

CV534CV544CV554

CV345

Shroud Dome

CV320

Lower 

Plenum

Control Volume Nodalization

CV111CV121

CV131

CV141CV151

CV511CV521

CV531

CV541CV551

CV512CV522

CV532

CV542CV552

CV112CV122

CV132

CV142CV152

CV513CV523

CV533

CV543CV553

CV113CV123

CV133

CV143CV153

CV114CV124

CV134

CV144CV154

CV515CV525

CV535CV545CV555

CV115CV125

CV135

CV145CV155

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

407

507

109

209

309

409509

110

210

310

410510

111

211

311

411

511

112

212

312

412512

113

213

313

413

513

114

214

314

414514

115

215

315

415

515

116

216

316

416516

Penetration - R1

Control Rod 

Stub Tubes

Penetration - R2Penetration - 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

Core Nodalization

FL110

FL510FL520FL530FL540FL550

FL120 FL130 FL140 FL150

FL111

FL511FL521FL531FL541FL551

FL121 FL131 FL141 FL151

CV111

CV511

CV121 CV131 CV141 CV151 CV521 CV531 CV541

CV551

FL160 FL165 FL170 FL175 FL180

FL116 FL117 FL118 FL119

FL112

FL512FL522FL532FL542FL552

FL122 FL132 FL142 FL152

CV112

CV512

CV122 CV132 CV142 CV152 CV522 CV532 CV542

CV552

FL161 FL166 FL171 FL176 FL181

FL126 FL127 FL128 FL129

FL114

FL514FL524FL534FL544FL554

FL124 FL134 FL144 FL154

CV114

CV514

CV124 CV134 CV144 CV154 CV524 CV534 CV544

CV554

FL163 FL168 FL173 FL178 FL183

FL146 FL147 FL148 FL149

FL113

FL513FL523FL533FL543FL553

FL123 FL133 FL143 FL153

CV113

CV513

CV123 CV133 CV143 CV153 CV523 CV533 CV543

CV553

FL162 FL167 FL172 FL177 FL182

FL136 FL137 FL138 FL139

FL115

FL515FL525FL535FL545FL555

FL125 FL135 FL145 FL155

CV115

CV515

CV125 CV135 CV145 CV155 CV525 CV535 CV545

CV555

FL164 FL169 FL174 FL179 FL184

FL156 FL157 FL158 FL159

Flow Path Nodalization

Channel Volumes

Bypass Volumes

Flow Path Between Channel Volumes

Flow Path Between Bypass Volumes

Legend

Flow Path Between Channel and 

Bypass Volumes (open when channel 

box fails)

CV151

CV541
3-8

3.1.1.3 Containment 

The primary containment is subdivided into seven distinct control volumes. The drywell is 

represented by the following four control volumes:

• Region internal to the reactor pedestal including the drywell sumps (CV205)

• Region external to the drywell pedestal from the floor to an elevation of 165’ (CV200)

• Region from 165 feet to the drywell head flange (CV201)

• Region above the drywell head flange (CV202)

The wetwell is represented by the following two control volumes:

• 1/16th segment of the wetwell (CV221)

• Remaining 15/16th of the wetwell (CV222)

One control volume represents the vent pipes and downcomers connecting the drywell to the 

wetwell (CV210). The MELCOR nodalization of the primary containment is shown in Figure 3-4. 

A total of 18 flow paths represent the containment flow pathways. Of these, two flow paths (FL200 

and FL202) connect the three drywell regions external to the reactor pedestal. Each of these flow 

paths is modeled with 50% of the interfacing flow area between the control volumes. This 

assumes a 50% obstruction by equipment and structures of the interface between the drywell 

regions. In addition, two flow paths (FL201 and FL203) allow for natural circulation inside the 

drywell.

Three flow paths (FL014, FL015, and FL016) connect the reactor pedestal to the lower drywell.

The open fraction of the personnel doorway is reduced based on the core debris elevation in the 

reactor pedestal after vessel failure (debris elevation determined from CAV package). Two 

additional flow paths (FL012 for flow from the drywell to the vent pipes and FL017 for nominal 

drywell leakage from the lower drywell to the reactor building) represent flow from the drywell. The 

nominal drywell leakage flow area, friction, and form losses are defined to match the nominal 

drywell leak rate. The elevation of nominal containment leakage through the drywell is modeled at 

the dominant location of drywell penetrations.

The downcomer vent exits to the two wetwell control volumes are represented by FL801 and 

FL802. Each flow path has the SPARC fission product pool scrubbing model activated within 

MELCOR for aerosols and vapors across all fission product classes.

Four flow paths (FL821, FL822, FL022, and FL023) model vacuum breakers intended to limit 

under-pressure failures of the drywell and wetwell. The wetwell-drywell vacuum breakers (FL821 

and FL822, representing connections with each wetwell control volume) open whenever the 

wetwell pressure exceeds the vent pipe pressure by 0.5 psid. The reactor building-wetwell 

vacuum breakers connect the wetwell airspace of the larger wetwell control volume with the 

northeast and southeast torus corner rooms, and open whenever the pressure in the wetwell 

drops 2 psi below the pressure in the reactor building.
3-9

Figure 3-4 MELCOR nodalization of the primary containment for Mark I analysis

Four flow paths (FL901, FL902, FL903, and FL904) represent the flow through various potential 

breach locations. FL901 represents the torus failure location, FL902 the drywell liner shear, FL903 

the head flange leakage, and FL904 the drywell liner melt-through.

The wetwell is modeled using two control volumes to capture the RCIC and single SRV exhaust 

and improve pressurization trends observed during the SBO accidents observed at Fukushima. A 

single well-mixed control volume under-predicts containment pressurization rates observed during 

an SBO when only the lowest set point SRV is operating. Enhanced evaporation of pool water 

and uncondensed vapor escaping the pool due to increased local wetwell temperatures cannot be 

captured with a single well-mixed control volume. A 1/16th segment of the wetwell torus (CV221) 

represents the near-field region of the lowest set point SRV t-quencher to promote higher local 

temperature predictions. The remaining wetwell torus control volume (CV222) interacts with the 

remaining SRVs, the RCIC system, vacuum breakers, wetwell vent lines, and torus failure. The 

two control volumes are connected through flow path FL851 to permit the transport of water and 

atmospheric material between the wetwell control volumes. The SRV exhaust while using a single 

SRV in the MELCOR model was set to the larger wetwell control volume (CV222) to enforce 

uniform heatup in agreement with operators alternating among SRVs during relief mode 

operations. 

CV200

CV222 FL012

FL802

CV221

FL822

FL903

(DW head

flange failure)

FL901

(Torus Failure)

CV210 CV210

CV205

FL014

Personnel

access

FL902

(DW liner shear)

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)

CV201

CV202

FL202

FL200

FL910

(Wetwell Hard-Pipe vent

to atmosphere)

FL801

FL821

FL920

(Drywell Hard-Pipe vent

to atmosphere)

FL201

FL203

FL023
3-10

Heat structures in the containment model are represented include the following:

• Drywell floor outside of the reactor pedestal

• Drywell floor inside of the reactor pedestal

• Drywell liner-air gap-concrete wall (between primary and secondary containment)

• Drywell liner representing the cylindrical and dome portions

• Biological shield wall in the lower drywell

• Biological shield wall in the mid-drywell

• Reactor pedestal

• Misc. drywell steel in the lower drywell

• Misc. drywell steel in the mid-drywell

• Misc. horizontal deposition surfaces in the lower drywell

• Wetwell torus liner

• Wetwell miscellaneous steel (equipment and structures)

The heat structure film-tracking model is activated to connect water film flows between the 

appropriate drywell liner heat structures.

3.1.1.4 Reactor Cavity

The drywell floor is subdivided into three regions for the purposes of modeling moltencore/concrete interactions. The first region (which receives core debris exiting the reactor vessel) 

corresponds to the reactor pedestal and sump floor areas (CAV 0). Debris that accumulates in the 

pedestal can flow out into the second region (through a doorway in the pedestal wall), 

corresponding to a 90 degrees sector of the annular portion of the drywell floor (CAV 1). If 

sufficient debris accumulates in this region, it can spread further into the third region, which 

represents the remaining portion of the drywell floor (CAV 2). This discrete representation of 

debris spreading is illustrated in Figure 3-5. 

Two features of debris relocation within the three cavities are modeled. The first models debris 

overflow from one cavity to another. The second manages debris spreading radius within the 

drywell floor region cavities (CAV 1 and 2). Control functions monitor 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 (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 (more 

debris head required to flow).

Debris entering CAV 1 and CAV 2 are 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. When the cavity debris temperature is at or above the liquidus, the shortest transit time (and 

therefore maximum transit velocity) of the debris front to the cavity wall is determined (10 minutes 

for CAV 1 as defined in MELCOR control function CF960, and 30 minutes for CAV 2 as defined in 

control function CF961). When the debris temperature is at or below the solidus, the debris front is 

assumed to be frozen. A linear interpolation is performed to determine the debris front velocity at 

temperatures between these two values. The CAVITY package model implemented enforces full 

mixing of all debris into a single mixed layer.
3-11

Figure 3-5 Discrete representation of debris spreading in the cavity

The solidus and liquidus temperatures in the parametric model that govern the rate of debris 

spreading on the drywell floor were modified in the present study. Original values of solidus and 

liquidus temperatures in the SOARCA model were 1,420K and 1,670K, respectively. These 

temperatures are representative of concrete solidus and liquidus. For containment venting 

calculations, the solidus and liquidus temperatures were changed to 1,700K and 2,800K, 

respectively which are more representative of the debris properties rather than concrete to better 

approximate the debris composition shortly after lower head failure which impacts debris 

spreading and the potential for liner melt-through. The revised liquidus temperature is 

representative of the liquidus temperature of a eutectic UO2/ZrO2 mixture. The revised solidus 

temperature was set at 1,700K to represent the lower bound of average melt temperature at 

vessel breach, and happens to coincide approximately with the melting point of steel. In the 

model, spreading is disallowed at debris temperatures less than the solidus temperature and 

occurs at a maximum rate (0.259 m/min) when debris temperature is above the liquidus 

temperature. Spreading rate varies linearly at temperatures between the solidus and liquidus 

temperatures.

3.1.1.5 Balance of Plant

A total of 41 control volumes, 71 flow paths, and 85 heat structures are modeled to represent all 

pertinent structures external to primary containment. These model elements represent the reactor 

building, turbine building, radwaste building, and the environment. Given its importance as a 

fission product release pathway if the containment fails, the reactor building is modeled in 

significant detail (30 control volumes and 80 heat structures). 

CAV 0

CAV 2

CAV 1

6.706 m 3.086 m

0.914 m
3-12

A sectional view of the reactor building is shown in Figure 3-6. It is modeled on a level-by-level 

basis, beginning in the basement (i.e., torus room) and sequentially rising up through the main 

floors to the refueling bay.

Figure 3-6 MELCOR nodalization of the reactor building for Mark I analysis

CV412

(Refueling Bay)

CV411

CV409

(195' SE

Quadrant)

CV407

(195' NE

Quadrant)

CV405

(165' South Half)

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

El. 91' 6"

El. 135'

El. 165'

El. 195'

El. 234'

CV901

(Environment)

FL424 (nom. leakage)

FL421 (roof failure)

FL903
3-13

The torus room level of the reactor building is represented by eight control volumes. These 

include volumes representing the main torus room, the northeast corner room, the stairwell in the 

northeast corner of the building, the southeast corner room, and the RHR A, B, C, and D heat 

exchanger and pump rooms. The next higher level of the reactor building is modeled by five 

control volumes. These include volumes representing the southern half of the building, the 

northern half of the building, the southwest stairwell enclosure, the northeast stairwell enclosure, 

and the steam tunnel. The next higher level of the reactor building is represented by five control 

volumes. The next higher level of the reactor building is represented by eight control volumes.

These volumes represent the northeast, southeast, northwest, and southwest quarters of the floor; 

the reactor building ventilation room; the drywell enclosure; the southwest stairwell enclosure; and 

the northeast stairwell enclosure. The refueling bay level (highest level) of the reactor building is 

represented by four control volumes. These volumes represent the open refueling bay (including 

the spent fuel pool but neglecting the separator/dryer storage pit), the southwest stairwell 

enclosure, the northeast corner room and the northeast stairwell enclosure. The flow paths 

modeled within the reactor building can be classified into the following categories: same level 

flows between distinct control volumes, open hatches, doors, blowout panels, flow pathways 

through walls, leakage pathways, stairwells and concrete hatches.

3.1.2 Description of Mark II MELCOR Model

To support the filtered/venting analyses, the original model from Reference [8] was converted to 

the input format standards of MELCOR 2.1. Unlike the extensive SOARCA input development 

performed for the Mark I analysis, the original BWR/5 Mark II input deck was modified primarily 

with the intention of performing the present analyses. 

Any significant variation between the input used here and that of the original model are presented 

below in addition to the general details necessary to illustrate the model. 

A detailed description of the secondary containment system is not provided. Given the 

containment failures assumed for the Mark II model, discussed in Section 3.1.3.4 , liner melt 

through is not considered. Furthermore, neither head flange leakage nor failure were observed 

during the accident analyses. These findings reduce the interactions between the primary and 

secondary containment systems to heat transfer through the boundary structures for the analyses 

provided. The reactor building is not a significant contribution to the findings of the analyses for 

Mark II.

3.1.2.1 Reactor Pressure Vessel and Reactor Coolant System

The RPV model is composed of 7 control volumes, 14 flow paths, and 18 heat structures. Figure 

3-7 depicts the facility nodalization and corresponding flow path connections of the RPV, with the

exception of the channel and bypass control volumes.

The RPV modeled includes the following control volumes:

• Lower plenum (CV100)

• Channel (CV111)

• Bypass (CV121)

• Upper plenum/separators (CV103)

• Dryers/steam dome (CV104)

• Downcomer (CV105)

• Jet pumps (CV300)
3-14

The downcomer control volume (CV105) represents the volume between the core barrel and the 

reactor vessel wall (excluding jet pump volume) from the baffle plate to the top of the steam 

separators. The downcomer control volume includes all volume external to the steam separators 

in the region above the core shroud dome. The lower plenum control volume (CV100) includes all 

reactor vessel volume below the top of the core support plate excluding the downcomer region 

and jet pumps. All volume internal to the jet pumps is represented by CV300. 

Figure 3-7 MELCOR reactor vessel nodalization for Mark II analysis 

The total jet pump volume was extracted from the original model’s lower plenum control volume, 

which included the jet pumps. The inlet and outlet height of the recirculation lines as well as the 

steady state flow control logic were adapted from the Mark I model. This practice was also 

performed to include the feedwater lines, feedwater pump control logic, and main steam lines. The

original volumes of the recirculation loops and jet pumps were preserved from the original model. 

A single lumped control volume represents the upper plenum mixing region in the shroud dome as 

well as the stand pipes and steam separators (CV103). In total, CV103 models the region 

CV104

CV103

CV102

CV105

FL312 FL312

Recirc-A 

FL382

Suction flow-A 

FL018

FL015 FL016

Main Steam A

FL373

FL017

Suction flow-B 

FL318

Main Steam

B, C & D

FL370

Feedwater A

FL376

Feedwater B

FL976

Recirc-B 

FL383

Drive flow - A 

FL386

Drive flow - B 

FL387
3-15

between the core top guides to the top of the steam separators. CV104, which is composed of the 

steam dryers and steam dome, receives steam from CV103. Water separated from the steam is 

returned to the downcomer control volume (CV105).

The flow paths present in the RPV are the following:

• Downcomer to jet pump A entrance (FL018)

• Downcomer to jet pump B entrance (FL318)

• Jet pump exit to lower plenum (FL312)

• Loop A recirculation water drawn from downcomer (FL382)

• Loop A recirculation water injected into jet pump loop (FL386)

• Loop B recirculation water drawn from downcomer (FL383)

• Loop B recirculation water injected into jet pump (FL387)

• Lower plenum to bypass (FL061)

• Lower plenum to channel (FL051)

• Bypass to upper plenum/steam separators (FL081)

• Channel to upper plenum/steam separators (FL071)

• Upper plenum/steam separators to downcomer (FL017)

• Upper plenum/steam separators to dryers/steam dome (FL015)

• Dryers/steam dome to downcomer (FL016)

Structures located within the RPV, not internally modeled as a core component, are represented 

as heat structures. The following heat structures are present in the RPV model:

External heat structures

• Cylindrical portion in the downcomer region (HS10501)

• Upper cylindrical portion of the vessel wall above the downcomer region (HS10403)

• Upper reactor pressure vessel hemispherical head (HS10402)

Internal heat structures

• Dryers (HS10401)

• Separators (HS10303)

• Core shroud in axial level JJ, (HS120JJ (HS12106 – HS12113))

• Lower plenum shroud (HS10004, HS10005, and HS10014)

• Upper plenum shroud (HS10301)

• Shroud dome (HS10302)

Given the single control volume representation of the drywell, the convective boundary for the 

outer surface of the heat structures representing the vessel cylinder and head (HS10402, 

HS10403, and HS10501) are bounded by the drywell volume (CV205). The cylindrical HS10403 

and HS10501 inner convective boundary is set to the downcomer (CV105) and dryers/steam 

dome (CV104), respectively. The hemispherical HS10402 convective boundary is assigned to the 

dryers/steam dome (CV104). The separators (HS10303) are bounded by the downcomer (CV105) 

and upper plenum/separators (CV103). However, the dryers are bound on both convective 

surfaces to the dryers/steam dome (CV104) region. HS11101, the support plate, has convective 

boundaries assigned to the channel (CV111) and the lower plenum (CV100). 
3-16

The modeling requirements for heat structures representing the shroud are divided between two 

regions, the shroud located above the core, which includes the shroud dome (HS10302) and 

upper plenum shroud (HS10301), and the shroud heat structures that are associated with core 

assigned control volumes. The representative heat structures modeling the shroud above the core 

are similarly modeled as the heat structures already discussed. The inner and outer convective 

boundaries of HS10301 and HS10302 are the upper plenum/separators and the downcomer 

(CV105). The shroud heat structures located in the core region (HS12106 – HS12113) each 

identify the bypass (CV121) as the inner convective boundary and the downcomer (CV105) as the 

outer convective boundary, the shroud located in the lower plenum (HS10004, HS10005, 

HS10014) identifies the lower plenum (CV100) and the downcomer (CV105) as the inner and 

outer convective boundaries, respectively.

The lowest set point SRV, represented by FL021, was separated from the lumped representation 

of all the SRVs modeled in the original input as FL022. Representing the lowest set point SRV as 

a single flow path allowed direct integration of the failure mode models from the Mark I model 

discussed in Section 3.1.3.3 . More significant pressure events may still actuate additional SRVs.

3.1.2.2 Reactor Core

The general discussion provided in Section 3.1.1.2 for the Mark I model is largely applicable to 

Mark II model; therefore, the provided discussion is limited to the model description and 

necessary modification. 

The core and lower plenum region, which represents the physical space where core degradation 

is evaluated, is divided into 5 radial rings and 13 axial segments. Axial levels 1 through 5 

represent the lower plenum and unfueled region immediately above the core plate. Axial level 5 

represents the steel associated with the assembly lower tie plate, fuel support pieces, and core 

plate. Axial segments 6 through 12 represent the fueled region and axial segment 13 represents 

the region above the active fuel, e.g., tie plate, cladding, and core top guide. Ring 5 is only present 

in the lower plenum and characterizes the region outside the inner radius of the core shroud 

located beneath the downcomer baffle plate; no intact structural material exists within this region. 

The core cell geometry and masses are retained from the original model. However, some 

modifications were necessary to accommodate the current lower head modeling approach in 

MELCOR. The current lower head model considers the curvature of the lower head unlike the flat 

cylindrical representation in the original model. The region impacted by the curvature of the lower 

head (axial levels 1 and 2) was therefore adjusted. These two axial levels were combined to form

a single axial level, axial level 1, and the total component masses were redistributed per core cell 

volume fraction. In addition, a new axial level was added, dividing the original axial level 3, to 

indicate the base of the downcomer baffle plate. The vessel inner radius was used as an 

approximation for the radius for the hemispherical lower head. The resulting modifications 

maintained the same total number of axial segments within the model. 

Coupling between the core cells and the hydrodynamic modeling performed in the control 

volumes is provided in Figure 3-8. Given the reduction in control volume fidelity, in comparison to 

the SOARCA model, the channel region for each core cell is associated with the control volume 

representing the channel (CV111) and bypass region of each core cell is associated with the 

control volume representing the bypass (CV121). Penetration failure modeling was deactivated in 

accordance with the Mark I modeling methodology discussed in Section 3.1.1.2 . 
3-17

Figure 3-8 MELCOR core nodalization for Mark II analysis 

3.1.2.3 Containment

The Mark II containment is represented by the following 5 control volumes:

• Wetwell (CV200)

• Downcomer vents (CV201)

• Lower reactor cavity (CV203)

• Upper reactor cavity (CV204)

• Drywell (CV205)

A total of 9 flow paths provide the interconnection among the control volumes representing the 

containment. The MELCOR nodalization of the primary containment is shown in Figure 3-9. 
3-18

Figure 3-9 MELCOR nodalization of the primary containment for Mark II analysis

Interconnectivity between the drywell and wetwell is modeled by 4 flow paths. The inlet and outlet 

of the downcomer vents (CV201) are represented by FL024 and FL025, respectively to CV205 

and CV200. The containment vacuum breakers are modeled by FL026 with a linear open fraction 

model with a differential pressure range of 0.0 and 3.0 psi. FL029 models an open 1⁄2-inch inerting 

system line connecting the drywell and wetwell.

Interconnectivity between the upper reactor cavity and drywell is modeled with 2 flow paths.

Various accesses and ports are lumped into FL027 between the two volumes. Two 8-inch drain 

lines from the drywell floor to the upper reactor cavity are modeled with FL028. The access ports 

connecting the lower reactor cavity and wetwell are lumped together and modeled as FL041 

(assumed open throughout the transient). FL040 represents two 4-inch sump drain lines through 

the upper reactor cavity floor to model prescribed failure between the upper and lower reactor 

cavity after lower head failure and debris relocation to the upper reactor cavity. Once the drain line 

fails, this flow path becomes fully open.
3-19

Containment heat structures present in the model are listed below.

• Wetwell wall (HS20001)

• Wetwell base slab (HS20002)

• Drywell floor support columns (HS20003)

• Wetwell pedestal section 2 (HS20004)

• Wetwell pedestal section 3 (HS20005)

• Wetwell pedestal section 4 (HS20006)

• Wetwell misc. steel (above water) (HS20007)

• Wetwell misc. steel (below water) (HS20008)

• Downcomers (HS20101)

• Drywell pedestal floor (HS20301)

• Drywell floor (HS20501)

• Reactor shield (HS20502)

• Drywell wall (HS20503)

• Drywell pedestal Section 1 (HS20504)

• Drywell head (HS20505)

• Drywell steel (vertical) (HS20506)

• Drywell steel (horizontal) (HS20507)

• Lower reactor cavity floor (HS10002) [added in the current model]

3.1.2.4 Reactor Cavity

Two cavities were modeled to represent the upper and lower reactor cavities in the representative 

Mark II containment. After lower head failure, debris relocates into the upper reactor cavity and 

initiates molten core concrete interaction. The ex-vessel debris is modeled as a single well-mixed

metal/oxide layer assumed instantly covering the cavity floor. 

Drain lines leading to the in-pedestal sump are a potential failure mode for the upper reactor 

cavity [9]. The challenge is that after core debris contacts the drainlines and fails them, it could 

provide a bypass from the drywell directly to the wetwell without going through the suppression 

pool. Within the model, failure of the upper reactor cavity may be identified through user input by 

selecting one of two methods. The first option initiates a timer to relocate the debris from the 

upper cavity to the lower cavity using a default delay of 20 minutes based on information provided 

in [9]. As a second option, the debris relocates to the lower cavity once the ablation depth 

exceeds the drywell pedestal floor thickness. However, this second option is used to investigate 

different cavity configuration among the fleet. The flow path representing the drain lines, FL040, 

opens upon upper reactor cavity failure, increasing the total bypass area available between the 

wetwell and drywell vapor space.

The addition of the debris from the upper reactor cavity instantly mixes with the existing steel 

mass in the lower cavity and core concrete interaction can resume once debris temperature 

exceed the ablation temperature.

3.1.3 Description of Mitigation Systems and Failure Modes

Performing the ELAP sequences with the Mark I and II models required modeling sequence 

events detailing operator actions as well as system availability and failure. The initial and 

boundary conditions for these systems are based on the interaction with the industry through 

public meetings (see Reference [10]). 
3-20

The descriptions provided in this section are applicable to both the Mark I and II models, unless 

stated otherwise below. The following sections discuss the reactor core isolation cooling system, 

FLEX injection, SRV operations and failure modes, containment failure modes, and vent 

operations.

3.1.3.1 Reactor Core Isolation Cooling

Reactor core isolation cooling (RCIC) is the only considered in-situ injection system for the 

sequences analyzed. Nodalization for the RCIC system includes:

• RCIC turbine (CV611)

• Flow from main steam line C to the RCIC turbine (FL611)

• Flow from the RCIC turbine to the suppression pool (FL613)

• Flow from the CST to feedwater piping including the RCIC pump (FL614)

• Flow from the suppression pool to feedwater piping including the RCIC pump (FL606)

The model includes a constant-flow pump, delivering 600 gpm via velocity-specified flow paths, 

with suction initially aligned to the condensate storage tank (CST). Switchover of pump suction to 

the suppression pool occurs upon receipt of a low CST water level signal. The model permits 

initialization with the CST failed at the time of scram if desired (in this case suction is aligned to 

the suppression pool). 

Steam flow through the RCIC turbine is modeled to account for the transfer of energy from the 

steam line to the suppression pool during RCIC operation. The flow of steam from main steam line 

to the RCIC turbine is modeled as a function of the pressure difference between the main steam 

line and the suppression pool. RCIC is modeled with automatic initiation and termination criterion.

RCIC is initiated on receipt of a reactor vessel low level-2 signal2

. RCIC is terminated on receipt of 

a reactor vessel high level-8 signal when DC power is available. For this analysis, manual pump 

operation is modeled where operators throttle the RCIC turbine/pump to maintain RPV water level 

after automatic initiation.

In general, RCIC failure is assumed if the temperature of the suction source for the RCIC pump 

exceeds 383 K (230 F) since a portion of water is used as coolant for the pump. This can happen 

only when the suction is aligned to the suppression pool. RCIC can also fail upon loss of dc power 

(in SOARCA, this led to overfilling of the steamline and flooding the turbine). In addition, 

consistent with assumption in SECY-12-0157, RCIC is assumed to fail at 16 hours if the other 

conditions are not met. These assumptions can lead to failure of RCIC anytime between zero to 

16 hours. 

3.1.3.2 FLEX Injection

FLEX injection provides a constant 500 gpm with a water temperature of 300K assuming 

availability of on-site water supplies. Injection can be specified either to the vessel or the drywell 

via the core sprays or drywell sprays, respectively. FLEX injection occurs at the time of lower 

head failure for the all the scenarios considered here if water addition is to be probabilistically 

considered in the accident sequence analysis. 

2 For the Mark II analysis, RCIC initiates at 31 seconds, which is prior to L2 water level signal. There is no 

significant impact on the overall response of the plant.
3-21

Injection can be specified as either continuous injection, injection termination upon reaching the 

top of wetwell level instrumentation range, or controlled injection prior to exceeding the wetwell 

level instrumentation range. Injection is full rated flow during for the continuous injection and 

injection with termination option. Controlled option injects at full flow until the wetwell level reaches 

close to the upper instrumentation range, and the flow is decreased to maintain water in the 

pedestal region of the drywell (but sufficiently low to inhibit drainage from the drywell through the 

downcomers into the wetwell). 

3.1.3.3 SRV Operations and Failure Modes

The SRV operations performed for the Mark I analyses include two pressure management stages 

(see Reference [10]). The first, occurring 10 minutes into the transient, models the operator 

performing managed SRV operations, operating available SRVs to manage wetwell heat-up, while 

maintaining RPV pressure between 800 and 1000 psig. After 1 hour, the second pressure 

management stage initiates as an operator depressurizes the reactor to 200 psig and maintains 

pressure between 200 psig and 400 psig3

. The 200 psig limit maintains sufficient turbine pressure 

to continue RCIC operation during the transient. For the Mark II analysis, the pressure band for 10 

minutes to 1 hour was set between 900 and 1000 psig. The SRV operation between 10 minutes 

and 1 hour does not significantly impact the Mark I and Mark II analyses. 

The SRV failure models are unchanged from the values provided in NUREG/CR-7110 [11]. The 

present model assumes either a stochastic failure or high temperature failure. Valve failure is 

modeled as fully open by default for either criterion. User options were included to permit further 

investigations into the SRV failure states by allowing the user to specify the open fraction of the 

failed SRV as well as disabling either the stochastic and/or high temperature failure criteria.

3.1.3.4 Containment Failure Modes

Two containment pressure boundary failure modes modeled in the Mark I containment venting 

analysis include the drywell liner failure resulting from debris interaction and drywell head flange 

leakage due to containment over pressure or drywell head flange failure due to over temperature. 

The Mark II model only incorporates the Mark I drywell head flange failure. Liner melt through is only 

modeled for the Mark I model since it was not identified as a failure mode in the Mark II design.

The Mark I debris spreading model, discussed in Section 3.1.1.4 , calculates the spreading of 

material from the pedestal region into the drywell as well as the time of debris-liner contact. Given 

limitations in the cavity model, temperature response of the liner due to heat transfer between the 

liner and debris is not explicitly modeled. Without direct simulation of liner failure, an 

approximation of 15 minutes was assumed for the duration of the liner remaining intact after 

debris contact (SOARCA assumption). Once failed, a flow path (FL904) is opened between the 

lower drywell control volume (CV200) and adjacent reactor building compartment (CV401), but 

debris is not permitted to relocate into the reactor building.

Drywell head flange overpressurization leakage and high temperature failure are modeled with

flow path (FL903). Overpressurization simulates head flange bolt strain in response to 

containment pressure. The head flange bolts are pretensioned, mating the flange surfaces while 

compressing a gasket seal. As pressure increases within containment, the pretensioning is 

overcome, the gasket seal decompresses over a distance of 0.03in, and continual strain opens 

3 The operators try to control the depressurization rate to 80oF/hr. No controlled depressurization is performed in 

the present analysis for simplicity and is not believed to significantly impact the results.
3-22

the drywell head. The high temperature failure mode corresponds to the temperature (644K) at 

which ductility of the gasket material is lost. When drywell atmosphere temperature exceeds 

644K, an assumed open area of 0.1 ft2

 is initiated [10].

3.1.3.5 Vent Operations

Hardened vent lines connecting the wetwell and drywell atmosphere spaces with the environment 

have been included in the models to investigate containment venting strategies. The wetwell vent 

line is modeled as a 16 in. line between the reactor building stack and wetwell using a discharge 

coefficient of 0.3 (see Reference [10]). The physical volume of the wetwell vent line is represented 

with CV223 and is connected to the wetwell volume with FL910 and to the environment with 

FL911. A simpler representation was applied for the drywell vent line, which is simply represented 

with a single flow path (FL920) using a 12 in. line (see Reference [10]). 

Anticipatory venting (pre-core damage) is performed to cool the suppression pool and avoid RCIC 

failure. This action is to be performed prior to RCIC failure, but an option to permit anticipatory 

venting after RCIC failure has occurred is available. However, if the RPV water level reaches the 

minimum steam cooling (2/3 core height), the vent is closed (see Reference [10]). 

Post core damage venting operation stage is performed to protect containment integrity and/or control 

releases. User options provide selection of either wetwell or drywell vent line operation when 

containment pressure exceeds primary containment pressure limit (prescribed as 60 psig) or at 

pressure suppression pressure (prescribed as 30 psig) as the actuation criterion (see Reference [10]).

Additionally, vent cycling can be imposed with a pressure band of 10 or 20 psi [10] below the actuation 

pressure, i.e., 50-60 psig or 40-60 psig. The cycling option is only exercised for the PCPL venting 

cases. The Mark I PCPL of 60 psig was retained for the Mark II analysis, but additional sensitivities 

were performed with a corrected value of 45 psig (see Table 3-3). 

If severe accident water addition (SAWA) is in effect, and wetwell venting is isolated upon high 

suppression water level (i.e., closing of the vent), then a switchover to drywell venting is required.

However, the switchover to the drywell vent is not immediate. The pressure is allowed to return to 

PCPL or PSP before drywell venting actuates. Cycling can be enforced following switchover but 

only for PCPL venting cases.

If severe accident water management (SAWM) is in effect, and wetwell venting is the preferred 

path, then no switchover to drywell venting is required.

3.2 Scenario Development and Run Matrix

In developing the MELCOR calculation matrix for containment filtered venting system analysis, a 

set of accident prevention and mitigation measures were considered, informed by the lessons 

learned from the Fukushima event, accident management alternatives contemplated by the 

industry, the current state of knowledge of severe accident progression in a BWR and mitigation 

alternatives, and by the experience gained from the previous effort (SECY-12-0157) to address 

the Near Term Task Force (NTTF) Recommendation 5.1. The accident scenarios considered are 

those associated with an extended loss of alternating power (ELAP) event related to BWRs with 

Mark I and Mark II containments caused by an external hazard (e.g., beyond design basis 

earthquake), thereby resulting in one of three possible outcomes: containment overpressure or 

over-temperature failure, liner melt-through failure, or maintaining the containment largely intact 

(i.e., without any significant loss of its radioactivity confinement function) as a result of venting or 

other mitigation measures. 
3-23

In an ELAP with the loss of all cooling function and absent any mitigation measures, the core is 

going to uncover leading to heatup, degradation, relocation of degraded core into the lower 

plenum, thermal loading of the reactor pressure vessel (RPV) lower head and consequent lower 

head failure, relocation of core debris into the reactor cavity, and ultimate containment failure by 

overpressure or other mechanisms. For this type of situation, the reactor core isolation cooling 

(RCIC) system is designed to provide core cooling, thus delaying core uncovery and subsequent 

accident progression until such time other DC-powered (battery or diesel generator) and portable 

mitigation systems become available. Hence, the operation of RCIC is considered as an important 

element in developing the MELCOR calculation matrix. 

Containment venting is considered as another strategy to prevent catastrophic containment failure 

(by overpressure or otherwise) and consequent large release of radioactivity. SECY-12-0157 

considered venting when the containment pressure exceeded the primary containment pressure 

limit (PCPL). Since then, the BWR industry introduced the concept of anticipatory early venting 

(pre-core damage venting at a pressure significantly below PCPL) in the process of updating its 

severe accident management guidelines (SAMGs). The idea behind the anticipatory early venting 

is to reduce the containment load pre-core damage, thereby affording more opportunity to put in 

place mitigation measures to address post-core damage containment performance. The early 

containment venting feature is included in the MELCOR matrix as are vent cycling and transition 

from wetwell venting to drywell venting. 

By far the most important mitigation strategy, considered in the development of MELCOR matrix, 

relates to water addition, both in the reactor vessel and in the containment. Supplementing precore damage water addition, this mitigation strategy calls for post-core damage water addition to 

both RPV and containment (using FLEX [10]). Moreover, this strategy calls for water management 

(i.e., controlled water addition to achieve a specified purpose or goal, for example, ensuring the 

wetwell is not flooded). 

In summary, the MELCOR run matrix presented in this section and the analysis presented in the 

next section are based on a number of assumptions which fall into two broad categories: (1) 

general assumptions and (2) specific assumptions. The general assumptions are as follows (see 

Reference [10]):

• All MELCOR transients start with an ELAP

• All transients are 72 hour in duration4

• Industry (BWROG) EPG/SAG Rev. 3 is in place

• FLEX is in place both pre- and post-core damage

• Possible end states of accident progression: liner melt-through (LMT), main steam line

creep rupture (MSLCR), drywell head flange leakage by overpressure and

overtemperature – all consequential to environmental release

• Recirculation pump leakage of 18 gpm per pump starts at the time of the initiating event

Specific MELCOR assumptions are focused on mitigation systems functions and include RCIC, 

RPV pressure control, containment venting, and water injection into either RPV or drywell. The 

assumptions are as follows:

4 The assumptions regarding offsite support in this study are similar to those used in the SOARCA study (NUREG1935) and the spent fuel pool study (NUREG-2161). The 72-hour time was chosen to perform a detailed 

assessment of the accident progression and to capture the release characteristics. In all cases where external 

water addition is successful, the radioactive releases have stabilized well before 72 hours. 
3-24

• RCIC suction is generally from SP (though option for suction initially from CST, then

transitioning to SP, is also considered as a sensitivity)

• Initial RCIC flow rate is 600 gpm and throttling of RCIC considered as an option for RPV

level control

• RPV pressure control in 800 – 1000 psig band at 10 minutes into ELAP followed by

controlled depressurization in one hour; subsequent pressure control in 200 – 400 psig

band for continued RCIC operation

• Anticipatory early containment venting prior to core damage (generally at 15 psig

containment pressure but a sensitivity case was run with 5 psig)

• Upon entry into SAG, vent closes; reopens at PCPL (60 psig nominally) with option to

reopen at PSP also considered

• Transition from wetwell to drywell venting is at high SP water level (nominally 21 feet

above the bottom of torus)

• The option of vent cycling is considered in (PCPL/PCPL-20) band and sensitivity cases

were run with (PCPL/PCPL-10) band

• Water injection into drywell at vessel breach from an external source under severe

accident conditions with 500 gpm flow rate; sensitivity of flow rate control (water

management) considered for drywell injection to prevent wetwell flooding to avoid the

need for switchover to drywell venting

• Initial buildup of water in the drywell from nominal leakage

The above assumptions were vetted with the industry stakeholders in several public meetings to 

assure that they are consistent with either the current or the planned SAMG practices. The 

industry has used largely the same or similar assumptions in their analysis using the MAAP code.

It should be noted, however, that the industry is in the midst of updating their SAMG. 

Selection of accident sequences covered by MELCOR calculation matrix was informed by staff’s 

accident sequence analysis. The accident sequence analysis, discussed elsewhere in detail, 

consisted of core damage event tree (CDET) and accident progression event tree (APET) 

development and binning of a rather large number of possible end states to a manageable fewer 

categories of similar outcome, and ranking these categories in descending order of frequencies. The 

MELCOR calculation matrix covered all event categories comprising 98% of possible end states. 

Another important consideration factored into the development of the MELCOR calculation matrix 

is tied to alternatives considered for regulatory analysis. Specifically, these alternatives fall into 

three categories discussed below:

Category 1—Overpressure Protection (Alternatives 1 and 2): This category considers the 

severe accident capable vent, as called for in Order EA-13-109, is in place to provide overpressure protection to the containment. Venting action, both pre- and post-core damage, is 

assumed, and available vent paths (wetwell and/or drywell) are considered. However, no further 

action (i.e., SAWA) is considered in this category. As such, liner melt-through in the Mark I 

containment and consequent uncontrolled release of radioactivity to the environment are not 

prevented. 

Category 2—Liner Melt-through Protection (Alternative 3): This category involves Category 1 

over-pressure protection in addition to water injection (i.e., SAWA/SAWM) to either the RPV or the 

containment (drywell) or both for Mark I BWRs. This action has a high likelihood of preventing 

liner melt-through and slowing down the containment basemat erosion. This results in a smaller 

environmental release due, in large part, to the removal of fission products by the wetwell. 
3-25

Category 3—Enhanced Measures to Reduce Radioactivity Release (Alternative 4): This 

category considers various enhanced measures to further reduce radioactivity release to the 

environment. This measure includes Category 2 water management (i.e., controlled water 

addition to ensure that the wetwell vent remains available (SAWM, vent cycling, etc.) and the use 

of engineered filters (small or large)).

Within Categories 2 and 3, several subcategories are considered delineating specific actions, 

individually and collectively. These actions include vent cycling, wetwell to drywell vent transition, 

etc. Accordingly, the alternatives (including sub-alternatives) considered in the construction of the 

MELCOR calculation matrix are listed in Table 3-1.

The MELCOR calculation matrix for a representative BWR with Mark I containment is shown in

Table 3-2. The calculation matrix for a representative BWR with Mark II containment is shown in

Table 3-3. The gray boxes in the tables signify the major deviations from the assumed initial and 

boundary conditions as part of sensitivity calculations. 

The Mark I analyses in Table 3-2 include several sensitivities to investigate uncertainties in 

equipment availability and operator actions. The outcome of these sensitivities as well as the

scenarios investigated were used to inform the analyses selected for the Mark II calculation matrix.

In addition, the water management option was removed from consideration for the Mark II 

calculation matrix. Given the increased suppression chamber volume, wetwell flooding up to the 

vent line was perceived as less significant for the Mark II analyses. These considerations resulted in 

a significant reduction to the total number of analyses performed for the Mark II calculation matrix.
3-26

Table 3-1 Listing of the alternative options and actions

Alternative 1 Order EA-13-109, EPG/SAG Rev 35, anticipatory venting, RPV pressure 

control.

Alternative 2A Make generically applicable Alternative 1 in rulemaking.

Alternative 3A Alternative 2 plus wetwell venting (no vent cycling) and severe accident 

water addition (SAWA) into RPV.

Alternative 3B Alternative 2 plus wetwell venting (no vent cycling) and SAWA into drywell.

Alternative 4Ai(1) Alternative 3A with vent cycling.

Alternative 4Ai(2) Alternative 3B with vent cycling.

Alternative 4Aii(1) Alternative 2 plus wetwell venting (no vent cycling) and severe accident 

water management (SAWM) into RPV.

Alternative 4Aii(2) Alternative 2 plus wetwell venting (no vent cycling) and SAWM into drywell.

Alternative 4Aiii(1) Alternative 4Aii(1) with vent cycling.

Alternative 4Aiii(2) Alternative 4Aii(2) with vent cycling.

Alternative 4Bi(1) Alternative 4Ai(1) with an external filter (Decontamination factor [DF]=10).

Alternative 4Bi(2) Alternative 4Ai(2) with an external filter (DF=10).

Alternative 4Bii Alternative 4Bi(2) with both pre- and post-core damage manual venting 

through drywell and an external filter (DF=10).

Alternative 4Biii Alternative 4Bi(2) with pre-core damage manual venting and post-core 

damage passive venting through drywell and an external filter (DF=10).

Alternative 4Biv Alternative 4Bi(2) with both pre- and post-core damage passive venting 

through drywell and an external filter (DF=10).

Alternative 4Ci(1) Alternative 4Ai(1) with an external filter (DF=1000).

Alternative 4Ci(2) Alternative 4Ai(2) with an external filter (DF=1000).

Alternative 4Cii Alternative 4Bi(2) with both pre- and post-core damage manual venting

through drywell and an external filter (DF=1000).

Alternative 4Ciii Alternative 4Bi(2) with pre-core damage manual venting and post-core 

damage passive venting through drywell and an external filter (DF=1000).

Alternative 4Civ Alternative 4Bi(2) with both pre- and post-core damage passive venting 

through drywell and an external filter (DF=1000).

The MELCOR results, presented in the next section, can be classified into two broad categories: 

(1) thermal-hydraulic output; and (2) source term output. The thermal-hydraulic output includes

the following:

• RPV pressure, temperature, and water level – determine the likelihood of main steam

line creep rupture failure

5 Elements of BWR Owners’ Group (BWROG) Emergency Procedure guidelines (EPG) and Severe Accident 

Guidelines (SAG), Revision 3, have been communicated to NRC by Nuclear Energy Institute (NEI) and BWROG.

Alternative 

Number Alternative Action 
3-27

• Containment (drywell) pressure and temperature – determine the likelihood of failure of

containment and various components (e.g., drywell head flange) by overpressure and/or

overtemperature

• Hydrogen and other non-condensable gas generation and migration – contribute to

containment overpressurization and hence, the timing of vent operation; also, determine

the potential for combustion in reactor building, vent line, etc.

• Wetwell (suppression pool) temperature and water level – determine the effectiveness of

pool scrubbing and the likelihood of wetwell vent becoming unavailable

The source term output provides input to consequence analysis using MACCS and includes, 

among others, the following quantities:

• Cesium release – an important contributor to latent cancer fatality risk and land

contamination

• Iodine release – an important contributor to early fatality risk
3-28

Table 3-2 Calculation matrix for a representative BWR with Mark I containment 
3-29

Table 3-3 Calculation matrix for a representative BWR with Mark II containment 

RPV Pressure 

control

Anticipatory 

Venting

SRV 

Operation

Availability (hr)

RCIC 

Availability 

(hr)

RCIC Suction Failure Temp (F) Open SRV after RCIC fails Setpoint (psig) Injection @ LH failure SAWA Injection rate (gpm) Allow SRV stuck open failure? Location Setpoint (psig)

Option Case

1 1 72 16 SP 230 N 15 - - Y WW 60

1p1 72 16 SP 230 N 15 - - Y WW 45

1 3 4 4 SP 230 N 15 - - N WW 60

1 5 72 16 CST 230 N 15 - - Y WW 60

1 6 72 16 SP 230 N 15 - - Y WW 30

2A 10 72 16 SP 230 N 15 RPV 500 Y WW/DW 60

10p1 72 16 SP 230 N 15 RPV 500 Y WW/DW 45

2A 11 72 16 SP 230 Y 15 RPV 500 Y WW/DW 60

11p1 72 16 SP 230 Y 15 RPV 500 Y WW/DW 45

3A 24 72 16 SP 230 N 15 DW 500 Y WW/DW 60

24p1 72 16 SP 230 N 15 DW 500 Y WW/DW 45

2A 42 4 4 SP 230 N 15 RPV 500 N WW/DW 60

3A 44 4 4 SP 230 N 15 DW 500 N WW/DW 60

3A 45 - 16 SP 230 - - DW 500 Y DW 60

3A 49 - 0 - - - - DW 500 Y WW/DW 60

3A 51 - 16 SP 230 - 15 DW 500 Y DW 15

3A 52 - 16 SP 230 - 15 DW 500 N DW 15

1 1a11 1b11 1b2

1 1b3

Pre Core Damage Post Core Damage

RCIC Operation Flex Operation Venting

 Same as case 1 but assuming suppression pool bypass is delayed until upper reactor cavity (URC) floor is completely ablated

 Same as case 1 but assuming lower reactor cavity is filled with water

 Same as case 1b1 but assuming immediate suppression pool bypass (0 min delay)

 Same as case 1b1 but assuming suppression pool bypass is delayed until URC floor is completely ablated

Sensitivity to DW-WW bypass and lower reactor cavity (LRC) pool
3-30

3.3 Accident Progression Results

This section provides a detailed analysis of the MELCOR calculations for the Mark I and Mark II 

representative models. The initial and boundary conditions for the analysis and the accident 

scenarios are provided in Sections 3.1.3 and 3.2 . 

3.3.1 BWR Mark I Results

It is important to understand the basic response of the plant. Three cases were selected to 

examine accident progression under severe accident conditions for cases with and without water 

addition or management. These cases are described in detail to represent the phenonmena 

pertinent to the run matrix. Cases 9 and 10 from the run matrix (see Table 3-2) are selected as 

representative runs for the water management and water addition (Section 3.3.1.1 ). Case 1 is 

without water addition and is discussed in Section 3.3.1.2 . A timing of key events is shown in 

Table 3-4. 

Table 3-4 Timing of key events for selected Mark I cases

Event Timing (hr) Case 1 (no 

water) 

Case 9 

(SAWM) 

Case 10 

(SAWA) 

Start of ELAP 0.0 0.0 0.0 

Operators first open SRV to control pressure 0.17 0.17 0.17 

Low-level 2 and RCIC actuation signal 0.18 0.18 0.18 

Operators open SRV to control pressure (200-400 psig) 1.0 1.0 1.0 

RCIC flow terminates 9.6 9.6 9.6 

SRV sticks open or operators open SRV after RCIC fails 16.0 9.6 16.0 

Downcomer water level reaches TAF 12.0 11.6 12.0 

First hydrogen production 13.7 13.2 13.7 

First fuel-cladding gap release 13.7 13.2 13.7 

Start of containment venting at 60 psig 14.9 14.4 16.3 

Relocation of core debris to lower plenum 15.6 15.5 15.5 

RPV lower head dries out 18.1 18.2 18.9 

RPV lower head fails grossly 23.0 23.4 23.1 

Drywell head flange leakage 27.1 - - 

Hydrogen burn in reactor building refueling bay 28.8 - - 

Drywell liner melt-through 31.4 - - 

Calculation terminated 72 72 72

Selected MELCOR Results Case 1 Case 9 Case 10

Debris mass ejected (1000 kg) 292 280 287 

In-vessel hydrogen generated (kg) 1195 1032 1232 

Iodine release fraction at 72 hr 2.28E-01 7.86E-02 8.10E-02

Cesium release fraction at 72 hr 1.94E-02 6.12E-03 7.26E-03
3-31

3.3.1.1 Accident Progression with Water Management/Addition

Figure 3-10 shows the RPV pressure for Case 9. During the first 10 minutes, there is no pressure 

control and a single SRV keeps the pressure at its lowest pressure setpoint. The hydraulic 

transient immediately following reactor scram and isolation results in a gradual decrease in water 

level because of coolant evaporation and discharge through the cycling SRV to the suppression 

pool as shown in Figure 3-11. At 10 minutes, the operators maintain the pressure between 800-

1000 psig by opening one or more SRVs. At about the same time, the RCIC starts automatically 

when the water level in the RPV reaches the low level L2. 

After RCIC initiation, the RPV water level is gradually restored by injecting water from the 

suppression pool at 600 gpm. At one hour, the operators begin controlled depressurization of the 

RPV and maintain the pressure between 200-400 psig (the lower pressure of 200 psig would 

allow enough steam for the continued operation of the RCIC turbine). The operators take manual 

control of the RCIC to maintain the level by throttling the water injection. The cycling of the RPV 

pressure after this time causes changes in the effective level of water because of variations in the 

average coolant density. 

Figure 3-10 Mark I RPV pressure for Case 9 (SAWM)

RCIC fails at 9.6 hours due to over-temperature in the suppression pool (230°F). For case 9, the 

operators depressurize the RPV by opening the SRV. Between the start of initial depressurization 

at 1 hour and RCIC failure, there are six openings of the SRVs. The operators cycle different 

SRVs during this time to provide a more uniform heatup of the suppression pool. Without water 

injection from RCIC, the RPV water level decreases gradually and reaches the top of active fuel 

by 11.6 hours. It takes more than 2 hours for the water to reach the bottom of the active fuel and 

an additional 2 hours for the core to relocate to the lower plenum.
3-32

Figure 3-11 Mark I RPV water level for Case 9 (SAWM)

The thermal response of fuel in the core is illustrated in Figure 3-12 and Figure 3-13, which show 

the calculated temperature of fuel cladding in the inner core ring and across the core mid-plane. 

Temperatures of fuel cladding at the top of the core begin to rise when the mixture level 

decreases below the top of active fuel. 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 3-14, 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 overheated fuel. This mechanically weak material fragments into particulate debris, which relocates 

toward the lower core plate as rubble. Particulate and molten debris continue to move downward 

in the core until 16 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 3-11. 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 3-14. 

The core debris cools as it enters the water-filled lower plenum. When residual water in the lower 

plenum is completely evaporated at 18 hours, debris temperatures begin to increase. Heat 

transfer from debris to the inner surface of the lower head causes the lower head temperature to 
3-33

increase as well. 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.

Figure 3-14 also illustrates changes in the configuration of core debris and lower plenum 

structures between the time of RPV dryout (18 hrs) and lower head failure (23.4 hours). At the 

time of lower head dryout, most of the core fuel assemblies (i.e., the central four 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 ring of the core. Debris in the lower plenum surrounds a 

forest of intact control rod guide tubes. As the temperature of lower plenum debris increases, it 

causes the structural failure of the control rod guide tubes, which collapse and are mixed into the 

growing debris bed. Immediately prior to lower head failure at 23.4 hours, the debris bed 

represents the mass of most of the core plus structural materials below the core. Failure of the 

lower head results in the rapid ejection of 280 metric tons of core debris onto the floor of the 

reactor pedestal in the drywell. By the time of the lower head failure, the outer ring is still intact 

and the injection of water into the RPV causes the temperature to decrease.

Figure 3-12 Mark I fuel cladding temperature in Ring 1 for Case 9 (SAWM)
3-34

Figure 3-13 Mark I fuel cladding temperatures at core mid-plane for Case 9 (SAWM)

Figure 3-14 Mark I core degradation and relocation for Case 9 (SAWM)
3-35

During the pre-core damage phase of the accident, the thermal response of the containment is 

governed by the operation of RCIC and SRVs. The RCIC injection of suppression pool water into 

the RPV and the removal of decay heat through the SRVs back to the suppression pool caused 

the gradual heat up of the water as shown in Figure 3-15. The step changes in the pool 

temperature coincide with the opening of the SRVs (see Figure 3-10). The pool temperature 

reaches 230°F at 9.6 hours causing the failure of RCIC due to high water temperature. The 

saturated pool continues to heat up for about an additional 5 hours as a result of opening of the 

SRVs until the water level in the RPV reaches the bottom of the active fuel. There is a gradual 

decrease in the suppression pool temperature following the opening of the wetwell vent at 14.3 

hours. The suppression pool remains saturated for the remainder of the accident, as the wetwell 

vent is kept open to stabilize the pressure in the containment.

The changes in the containment pressure and the total integral vent flow are shown in Figure 

3-16. By the time RCIC fails, the containment pressure has reached 12.5 psig, which is below the

threshold of 15 psig for anticipatory venting. The containment pressure starts to rapidly increase

due to release of hydrogen through the SRV to the torus (i.e., over 500 kg before venting). At 14.3

hours after the initiating event, the containment pressure increases above the venting pressure of

60 psig, and the wetwell vent is opened. The rapid rise in containment pressure post core damage

requires a timely response to vent the containment.

Figure 3-15 Mark I suppression pool temperature for Case 9 (SAWM)
3-36

Figure 3-16 Mark I containment pressure and vent flow for Case 9 (SAWM)

Figure 3-17 shows the containment water level. By the time of the lower head failure, there is an 

accumulation of 1.4 ft of water above the drywell floor as a result of leakage from the recirculation 

lines. After venting the containment, the water level in the suppression pool continues to decrease 

as coolant is discharged from the containment. Following lower head failure, water injection into 

the RPV begins as shown in Figure 3-18. The initial flow rate is 500 gpm, which continues for the 

next 15.6 hours when the water level in the torus approaches 21 ft. At this time, there is a gap of 

about 2 hours6 with no water injection as the operators try to throttle the water and maintain the 

level below 21 ft.

Containment atmosphere temperatures (see Figure 3-19) remain relatively low even after core 

damage because the steam/hydrogen mixture cools as it bubbles through the suppression pool. 

Immediately following vessel breach, containment atmosphere temperatures increase due to the 

accumulation of core debris on the reactor pedestal and drywell floors. However, water injection 

maintains the drywell temperature below 500 K in the long term. 

Soon after debris is released onto the reactor pedestal floor, it flows laterally out of the cavity

through the personnel access doorway and spreads out across the main drywell floor. Lateral 

movement and spreading of debris across the drywell floor shown in Figure 3-20 indicates that the

debris does not reach the steel shell at the outer perimeter of the drywell. A combination of water 

injection and containment venting prevents containment failure.

6 This is a modeling assumption to control the water flow rate.


)LJXUH 0DUN,FRQWDLQPHQWZDWHUOHYHOIRU&DVH6$:0

)LJXUH 0DUN,539ZDWHULQMHFWLRQUDWHIRU&DVH6$:0
3-38

Figure 3-19 Mark I containment atmosphere temperature for Case 9 (SAWM)

Figure 3-20 Mark I debris spreading in DW for Case 9 (SAWM)
3-39

Volatile fission products release from fuel begins at 13.2 hours, when a portion of the fuel gap 

inventory is released due to early fuel cladding failures. As fuel temperatures rise (see Figure 3-12, 

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 volatile species 

such as cesium and iodine from the fuel and the distribution in the RPV and the containment are

shown in Figure 3-21 and Figure 3-22. 

The spatial distribution of cesium does not significantly change with time in the same way the 

distribution changed for iodine. This is most evident by the behavior inside the RPV. Even though the 

fraction of the initial inventory released from fuel that is deposited initially (e.g., on the steam 

separators and dryers) is consistent, a larger proportion of deposited cesium is retained inside the 

vessel. These differences in iodine and cesium behavior can be attributed to differences in the 

physical properties of their dominant chemical forms. Iodine is transported as CsI. The cesium 

contribution to CsI represents only 6% of the total cesium inventory. The vast majority 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. However, iodine is preferentially 

transported to the torus, but cesium remains deposited on in-vessel structures. This is clear from 

Figure 3-22 that shows between the time of containment venting at 14.3 hours and vessel breach, 

most of the iodine inventory in the RPV is purged to the suppression pool through the open SRV (see 

Figure 3-10).

It is important to note that the release of radionuclides that follows containment venting at 14.3 hours is 

characterized by a single sudden release about 9 hours before vessel breach and start of water 

injection. The injection of water after lower head failure stabilizes the environmental release.

The particle size distribution of CsI released through the wetwell vent (see Figure 3-23) clearly shows 

that the majorty of aerosols are in the submicron size range (~99%) with almost 80% in the lower bin 

size of 0.15 micron. The insert in the figure is the mass averaged particle size. The transport and 

scrubbing of the fission product aerosols and vapors through the suppression pool in the short time 

between the time of core damage at 13.2 hours and containment venting at 14.4 hours (see Table 

3-4) results in small particle sizes that make it through the wetwell vent in the initial sudden release. 

Any scrubbing of aerosols in this size range is expected to be minimal.
3-40

Figure 3-21 Mark I cesium fission product distribution for Case 9 (SAWM)

Figure 3-22 Mark I iodine fission product distribution for Case 9 (SAWM)
3-41

Figure 3-23 Mark I CsI particle size distribution for Case 9 (SAWM)

Since the release to the environment is dominated by containment venting before vessel breach 

and vessel injection, the mode of water control (i.e., water addition vs. water management) is not 

expected to significantly change the source term. Case 10 is a representative run to see the 

effects of continuous water injection that requires a switchover to drywell venting. In this case, it is 

assumed that once the water level in the torus reaches 21 ft, the wetwell vent is isolated and a 

switchover to drywell venting is initiated when the containment pressure once again reaches 60 

psig. Two other changes in this scenario compared to Case 9 include the assumption that there is 

no depressurization of the RPV following failure of the RCIC and anticipatory venting is allowed 

even after RCIC failure.7

Figure 3-24 shows the RPV pressure for Case 10. As mentioned before, the RPV is not 

depressurized after RCIC failure but continues to cycle between 200-400 psig for an additional 6 

hours until the SRV is stuck in an open position. The higher frequency of SRV cycles is a direct 

result of core degradation during this time. Even though the SRV fails to open due to thermal 

failure, there is a period of vessel pressurization following core debris relocation to the lower head.

The containment pressure response is shown in Figure 3-25. There is only an hour difference 

between the failure of RCIC at 9.6 hours and the time that the pressure inside the containment 

reaches the threshold of 15 psig for early venting. Here it is assumed that anticipatory venting 

7 These assumptions were made to see the response of the RPV to elevated pressure that could lead to possible 

main steam line creep rupture following core degradation and the timing of the stuck open SRV due to thermal 

seizure.
3-42

occurs and the wetwell vent is opened. Once the water level inside the RPV reaches the minimum 

steam cooling at 12.8 hours, the wetwell vent is isolated. The containment starts to pressurize as 

a result of core degradation as shown before (see Figure 3-16 for Case 9). The early venting in 

this case results in a time delay for post core damage containment venting at 60 psig as 

compared to Case 9 (at 16.3 hours or about 2 hours later). With water injection at lower head 

failure (23.8 hours), it takes 18.4 hours to increase the torus water level to 21 ft. The containment 

is once again isolated and the pressure continues to increase due to decay heat addition and noncondensable gas generation as a result of molten core concrete interaction. The containment 

pressure reaches 60 psig at 54.3 hours or more than two days after the initiating event. 

The sudden containment depressurization from the drywell causes a back flow of water from 

downcomer vents onto the drywell floor as shown in Figure 3-26. The vacuum breaker model 

takes into account the hydrostatic head of water inside the vent lines as water injection leads to an 

increase in the downcomer level. This back flow of water from the suppression pool and water 

injection into the RPV leads to a buildup of water of more than 15 ft above the bottom of 

containment vessel by the end of the calculation at 72 hours.

The fractional distribution of cesium and iodine are shown in Figure 3-27 and Figure 3-28. The 

total release of cesium (7.3E-3 for Case 10 and 6.1E-3 for Case 9) and iodine (8.1E-2 for Case 10 

and 7.9E-2 for Case 9) are comparable between the water addition case (case 10) and water 

management case (case 9) even though the fractional distributions inside the RPV and 

containment are initially somewhat different due to differences in boundary conditions and core

degradation process.

Figure 3-24 Mark I RPV pressure for Case 10 (SAWA)


)LJXUH 0DUN,FRQWDLQPHQWSUHVVXUHDQGYHQWIORZIRU&DVH6$:$

)LJXUH 0DUN,FRQWDLQPHQWZDWHUOHYHOIRU&DVH6$:$
3-44

Figure 3-27 Mark I cesium fission product distribution for Case 10 (SAWA)

Figure 3-28 Mark I iodine fission product distribution for Case 10 (SAWA)
3-45

3.3.1.2 Accident Progression without Water Management/Addition

This section describes the thermal response of the plant without water injection after core 

damage. The representative scenario is Case 1 (see Table 3-2 for the list of boundary conditions).

The focus here is only on parameters of interest such as the water level and thermodynamic 

conditions inside the containment, the mode of containment failure, and the source term. The 

timing of key events is provided in Table 3-4 that shows comparable timings with Cases 9 and 10. 

As mentioned before, the leakage of water (total of 36 gpm) from the recirculation lines leads to a 

buildup of water inside the pedestal and the lower drywell as shown in Figure 3-29. By the time of 

lower head failure at 23 hours there is an accumulation of 1.6 ft of water on the drywell floor. Once 

the debris is ejected from the RPV (see Figure 3-30), it starts vaporizing the water and at the 

same time the heat transfer to water cools the debris. It takes 2.7 hours to completely deplete the 

existing water on the drywell floor. Without interaction with the water, the debris starts to heat up 

once again. 

Figure 3-29 Mark I containment water level for Case 1
3-46

Figure 3-30 Mark I ex-vessel debris temperature for Case 1

The movement of debris out of the pedestal and on to the drywell floor is shown in Figure 3-31. 

The debris flows laterally out of the cavity through the open personnel access doorway and 

spreads out across the main drywell floor. However, the cooling of debris temporarily stops the 

flow towards the drywell liner. After the debris heats up again, the debris reaches the steel shell at 

the outer perimeter of the drywell and the thermal attack of the debris against the steel shell 

results in shell penetration. Because of wetwell venting and containment depressurization earlier, 

the shell failure and opening of a release pathway for fission products into the torus room of the 

reactor building does not result in any significant release to the environment. 
3-47

Figure 3-31 Mark I debris spreading in DW for Case 1

The containment atmosphere temperatures remain relatively low before lower head failure as 

seen before (see Figure 3-33). The presence of the water in the drywell also limits the 

temperature rise until the dryout in the drywell at 25.7 hours. At this time, the atmosphere 

temperature in the pedestal increases to nearly 1900 K and continues to increase in the long term.

High gas temperature near the top the drywell leads to failure of the drywell head seals and the 

model assumes leakage through the head flange with a constant leak area of 0.1 ft2

. The leak rate 

is relatively small since the venting has already depressurized the containment.

Before drywell shell melt-through occurs, hydrogen leaks through the drywell head flange and 

accumulates in the reactor building refueling bay. A flammable mixture quickly develops in the 

refueling bay, causing a hydrogen combustion (see Figure 3-32). Small increases in internal 

pressure (0.25 psig) cause the blowout panels in the refueling bay to open; the roof also fails at an 

overpressure of 0.5 psig (see Reference [11] and Figure 3-6)8

. Therefore, the failure of the 

refueling bay offers another release pathway to the environment immediately after a hydrogen 

burn occurs within the building. After melt-through of the drywell liner, additional hydrogen is 

released from the drywell into the torus room and is transported upward through open floor 

gratings into the ground level of the reactor building. The pressure rise within the building causes 

several doorways to open, including the large equipment access doorway at grade level. Several 

8 In the MELCOR model, combustible gases can still accumulate in the control volume leading to the second 

hydrogen combustion even though the roof has failed. This does affect the overall results since the containment 

fails by liner melt-through shortly after. 
3-48

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 flow areas in the 

refueling bay at the top of the building, creates an efficient transport pathway to the environment 

for material released from containment. 

The fractional distribution of cesium and iodine are shown in Figure 3-34 and Figure 3-35. The 

environmental release is dominated by the initial sudden venting of containment followed by a 

gradual release after vessel breach due to revaporization of CsI as discussed before.

Figure 3-32 Mark I refueling bay gas concentration for Case 1
3-49

Figure 3-33 Mark I containment atmosphere temperature for Case 1

Figure 3-34 Mark I cesium fission product distribution for Case 1
3-50

Figure 3-35 Mark I iodine fission product distribution for Case 1 

3.3.1.3 Source Term and Containment Response

A summary of cesium release fractions for the different scenarios in the run matrix (see Table 3-2) 

is given in Figure 3-36. The summary of main parameters is provided in Table 3-5. In most cases 

with water injection, the cesium release fraction remains below 1% (or a DF of greater than 100) 

and the mode of containment venting and water injection do not greatly affect the cesium release.

Changes in the boundary conditions such as opening of the SRVs by the operators before core 

damage, the fractional open area of thermally seized SRV, early (pre-core damage) containment 

venting, and injection source of RCIC (SP vs. CST) can affect the release fractions. These 

boundary conditions affect the core degradation and the thermodynamic conditions inside the 

RPV and ultimately the distribution of the fission products in the RPV and containment. 

In general, the cases without water injection or when water injection stops at high water level (see, 

for example, case 7 in Table 3-2) show higher release fractions. However, the release fraction of 

cesium is not significantly affected by the containment failure (liner melt-through and DW head 

leakage) because water injection occurs at the time of vessel breach whereas controlled 

containment venting has occurred much sooner during core damage. Therefore, in all cases the 

early release is characterized by a sudden release at the time of venting.
3-51

Figure 3-36 Mark I cesium environmental release fraction

Figure 3-37 shows the effects of containment venting, water injection, and suppression pool 

scrubbing on fission product releases to the environment for several cases (with and without water 

injection at lower head failure). The injection of water after lower head failure stabilizes the 

environmental release, but clearly does not affect the magnitude of release at the time of venting.

The highest releases are associated with main steam line creep rupture cases since suppression 

pool can be bypassed. For these scenarios, the DW head leakage and hydrogen combustion in 

the refueling bay occurs almost immediately after the release into the drywell. This provides a

direct path for release to the environment.

Figure 3-38 shows that for all the cases with sustained water injection into the drywell or the RPV 

(that eventually accumulates on the drywell floor), the maximum structure temperatures at the 

drywell upper head or the drywell liner near the elevation of the drywell vent remains below 500o

F.

The cases without water injection in general experience the highest temperatures at the time of 

vessel breach, since the exposure of the debris to the drywell atmosphere as it exits the vessel 

and the circulation of hot gases inside the drywell can heat up both the atmosphere and the 

surrounding structures. Water injection and submergence of the debris result in direct heat 

transfer from the debris to the overlying water, and the drywell atmosphere directly transfers heat 

to the cooler pool surface. In addition, the water cools the concrete decomposition gases before 

they enter the drywell atmosphere.

The relation between the maximum gas and structure temperatures in the middle drywell is shown 

in Figure 3-39. As stated before, the cases without water injection or main steam line creep 
3-52

rupture result in higher temperatures in the drywell. For the sustained water injection cases, the 

structure temperature on the average remains about 100o

F cooler than the atmosphere.

Figure 3-37 Effect of water injection on selected cesium releases for selected cases

Figure 3-38 Mark I containment gas and structure temperatures
3-53

Figure 3-39 Mark I maximum containment gas and DW liner temperatures

The behavior of hydrogen in the containment is shown in Figure 3-40 and Figure 3-41. The total 

mass of hydrogen produced during the transient for Case 9 is about 2600 kg with about 1000 kg 

generated in-vessel. The blue line represents the total hydrogen generation, which should be 

almost identical with the amount remaining inside the containment and the amount that is vented 

(represented by the green line). The amount of hydrogen that remains inside the containment 

(both the drywell and the wetwell air space as shown by the red line) quickly decreases as a result 

of venting. With the wetwell vent open during the transient, the total amount of hydrogen is kept 

very low in the long term (below 30 kg). Therefore, containment venting is very efficient in purging 

the hydrogen from the containment. The presence of water seems to avoid containment failure 

and any uncontrolled release of hydrogen to the reactor building, which remains intact for the 

duration of the transient. 

The conditions inside the containment show that following vessel breach and water injection, 

evaporation of water leads to a high mole fraction of steam in excess of 80% (see Figure 3-41 and

Figure 3-42). The mole fraction of steam in the wetwell at the time of containment venting is also 

very high, which leads to steam inerting of the containment vent line in a very short time (see 

Figure 3-43 and Figure 3-44). These conditions are not conducive to an energetic hydrogen 

combustion. 


)LJXUH 0DUN,K\GURJHQJHQHUDWLRQDQGWUDQVSRUWIRU&DVH

)LJXUH 0DUN,GU\ZHOOJDVGLVWULEXWLRQIRU&DVH


)LJXUH 0DUN,ZHWZHOOJDV GLVWULEXWLRQIRU&DVH

)LJXUH 0DUN,ZHWZHOOYHQWOLQH JDVGLVWULEXWLRQIRU&DVH


)LJXUH 0DUN,ZHWZHOOYHQWOLQH JDVGLVWULEXWLRQDWYHQWLQJIRU&DVH

$FDOFXODWLRQZDVSHUIRUPHGWRLQYHVWLJDWHWKHLPSDFWRIZDWHUDGGLWLRQSULRUWRYHVVHOEUHDFKRQ

WKHILVVLRQSURGXFWUHOHDVHWRWKHDWPRVSKHUH7KLVLVDYDULDWLRQRI&DVHZLWKWKHDVVXPSWLRQ

WKDW539LQMHFWLRQEHJLQVDWKRXUVRUVKRUWO\DIWHUFRUHKHDWXSVHH7DEOH7KHZDWHU

OHYHOLVVKRZQLQ)LJXUH7KHZDWHUOHYHOLVQHDUWKHERWWRPRIDFWLYHIXHOE\WKHWLPHWKH

JSP)/(;LQMHFWLRQVWDUWV,WWDNHVDERXWKRXUVEHIRUHWKHOHYHOLVUHVWRUHGDERYHWKHWRSRI

DFWLYHIXHODQGVRPHILVVLRQSURGXFWUHOHDVHIURPWKHIXHOKDVDOUHDG\RFFXUUHG)LJXUH

LQGLFDWHVWKDWDERXWRIWKHFHVLXPLQYHQWRU\LQWKHFRUHKDVDOUHDG\EHHQUHOHDVHGIURPWKH

IXHO+RZHYHUWKHUHOHDVHWRWKHHQYLURQPHQWLVVLJQLILFDQWO\ORZHUDVVKRZQLQ)LJXUHDQGLW

RFFXUVDWDPXFKODWHUWLPH7KHFRROLQJRIWKHFRUHDIIHFWVWKHFRQWDLQPHQWSUHVVXUL]DWLRQDQG

GHOD\VWKHWLPLQJRIYHQWLQJ7KHUHLVRQO\DVPDOOWLPHZLQGRZEHIRUHWKHFHVLXPUHOHDVHIURPWKH

FRUHLVDOPRVWFRPSOHWHDERXWKRXUVDVVKRZQLQ)LJXUH,QDOOWKHEDVH0(/&25

FDOFXODWLRQVLWZDVDVVXPHGWKDWLQMHFWLRQEHJLQVDWWKHWLPHRIORZHUKHDGIDLOXUHDQGWKHLQLWLDO

YHQWLQJVXGGHQUHOHDVHZDVSUHGLFWHGWRRFFXUPXFKVRRQHU7KHUHIRUHWKHLQMHFWLRQWLPLQJLV

LPSRUWDQWLQGHWHUPLQLQJLIWKHUHLVDQ\UHGXFWLRQLQUHOHDVH
3-57

Figure 3-45 Mark I RPV water level for Case 9 (IVR)

Figure 3-46 Mark I cesium release fraction from fuel for Case 9 (IVR)
3-58

Figure 3-47 Mark I cesium release fraction to environment for Case 9 (IVR)
3-59

Table 3-5 Summary of main parameters for Mark I analysis

3.3.2 BWR Mark II Results

The following analyses are presented to facilitate a general understanding of the Mark II 

containment response. The representative cases discussed below document cases with and 

without SAWA. Case 11 has been selected as the representative water injection case and is 

discussed in Section 3.3.2.1 whereas Case 1 is discussed in Section 3.3.2.2 to document a case 

without water injection. Event outlines are provided in Table 3-6 for the selected representative 

cases. 
3-60

Table 3-6 Timing of key events for selected Mark II cases

*RCIC actuates erroneously prior to low level signal. Corrected runs give RCIC actuation at 0.07 hours. No significant deviation is 

anticipated due to this error.

3.3.2.1 Accident Progression with Water Addition

Case 11 is presented here as the representative water addition case. The options in Table 3-3

dictate that containment venting will commence using the wetwell vent line when containment 

pressure reaches PCPL and FLEX injection will begin at lower head failure. FLEX injection is 

modeled at a continuous volumetric rate of 500 gpm injection sourced to the RPV. As the wetwell 

fills due to continuous FLEX injection, switchover from the wetwell vent to the drywell vent will be 

available should this occur. 

The initiating event results in the loss of AC power, containment isolation, reactor scram, MSIV 

closure, feedwater coastdown, and recirculation pump trip. The loss of injection and closure of the 

MSIVs permits RPV pressure to increase due to continual heat generation. RPV pressure 

9 This is the time when the water level in the downcomer reaches the top of active fuel. The water level in the core 

is slightly different.

Event Timing (hr) Case 1 

(no water) 

Case 11 

(SAWA) 

Start of ELAP 0.0 0.0 

Operators first open SRV to control pressure 0.17 0.17 

RCIC actuation signal 0.01 0.01 

Operators open SRV to control pressure (200-400 psig) 1.0 1.0 

RCIC flow terminates 8.4 8.4 

SRV sticks open or operators open SRV after RCIC fails 16.8 8.4 

Water level reaches TAF9 10.7 10.3 

First hydrogen production 12.9 12.1 

First fuel-cladding gap release 13.0 12.2 

Start of containment venting at 60 psig 22.8 20.3 

Relocation of core debris to lower plenum 15.3 14.5 

RPV lower head dries out 20.3 18.5 

RPV lower head fails grossly 22.0 20.0 

Calculation terminated 72 72

Selected MELCOR Results Case 1 Case 11

Debris mass ejected (1000 kg) 248 220 

In-vessel hydrogen generated (kg) 1232 1307 

Iodine release fraction at 72 hr 1.98e-1 4.50e-3 

Cesium release fraction at 72 hr 2.46e-2 4.22e-4 
3-61

eventually increases until the lowest set point SRV actuates automatically. Figure 3-48 displays 

the RPV pressure for Case 11. 

RCIC injection initiates shortly after the reactor scram. RCIC pump suction is aligned to the 

suppression pool. RCIC injects approximately 600 gpm of suppression pool water into the vessel 

while steam drawn to operate the RCIC turbine is exhausted to the suppression pool. Operators 

assume control of the RCIC system by manually throttling RCIC injection to prevent a high reactor 

water level trip shutting down RCIC. Figure 3-49 illustrates the RPV water level for Case 11.

Operators initiate controlled depressurization of the RPV after 1 hour. Pressure is maintained 

above the operating pressure of the RCIC turbine as operators cycle available SRVs to maintain 

RPV pressure within the range of 200-400 psig. SRV operations and RCIC turbine exhaust results 

in an increase of the suppression pool temperature. At 8.4 hours, RCIC fails as a result of 

suppression pool water temperature exceeding 230o

F. The RPV is depressurized by operator 

opening an SRV following loss of RCIC for Case 11.

Without RCIC injection, boil-off of the RPV water level commences. The water level falls to the top 

of active fuel at 11.0 hours and below the bottom of active fuel at 13.3 hours. The thermal 

response of intact cladding in the active fuel region is illustrated in Figure 3-50 and Figure 3-51, 

which show the calculated temperature of the fuel cladding in the inner core ring and across the 

core mid-plane, respectively. As temperatures increase in the uncovered regions of the fuel, 

exothermic oxidation of the Zircaloy cladding initiates. The integral mass of hydrogen generated 

in-vessel is displayed in Figure 3-51. 

Figure 3-48 Mark II RPV pressure for Case 11

0

200

400

600

800

1000

1200

1400

0 6 12 18 24 30 36

RPV Pressure (psia)

Time (hr)

Start of Controlled

Depressurization

RCIC Fails

SRV Opens

Core Relocation

Lower Head Failure


)LJXUH 0DUN,,539ZDWHUOHYHOIRU&DVH

)LJXUH 0DUN,,IXHOFODGGLQJWHPSHUDWXUHLQ5LQJIRU&DVH

0

10

20

30

40

50

60

0 2 4 6 8 10 12 14 16 18 20

RPV Water Level (ft)

Time (hr)

Top of Active Fuel (TAF)

Bottom of Active Fuel (BAF)

Automatic RCIC 

Actuation

Start of Fuel Relocation

into Lower Head

Top of Active Fuel (TAF)

Bottom of Active Fuel (BAF)

Automatic RCIC 

Actuation

Start of Fuel Relocation

into Lower Head

0

500

1000

1500

2000

2500

3000

10 11 12 13 14 15 16 17 18 19 20 21 22

5LQJ&ODG7HPSHUDWXUH.

7LPHKU

Axial Level 7

Axial Level 8

Axial Level 9

Axial Level 10

Axial Level 11

Axial Level 12

Indicates intact fuel no

longer present at this location


)LJXUH 0DUN,,IXHOFODGGLQJWHPSHUDWXUHVDWFRUHPLGSODQHIRU&DVH

$VWKHFRUHGHJUDGHVGHEULVPRYHVGRZQZDUGWKURXJKDYDLODEOHIUHHYROXPH/DUJHUHORFDWLRQ

LQLWLDWHVLQWKHE\SDVVUHJLRQDVFRQWUROPDWHULDOIDLOVHDUO\RQ:KLOHLQWDFWIXHOLVPRGHOHGWRUHVWULFW

GRZQZDUGSURSDJDWLRQRIIXHOGHEULVGXHWRWKHOLPLWHGDYDLODEOHYROXPHFDQLVWHUZDOOIDLOXUHVSHUPLW

UDGLDOUHORFDWLRQRIIXHOGHEULVLQWRWKHE\SDVVUHJLRQ5DGLDOUHORFDWLRQRIIXHOGHEULVEDFNLQWRWKH

FKDQQHOUHJLRQLVSHUPLVVLEOHDVWKHFDQLVWHUZDOOIDLOVLQORZHUUHJLRQVRIWKHFRUH

$VWKHFRUHGHJUDGDWLRQFRQWLQXHVGHEULVEHJLQVWRDFFXPXODWHRQWRWKHFRUHSODWH:KLOHLQ

FRQWDFWFRQGXFWLRQEHWZHHQWKHGHEULVDQGWKHFRUHSODWHFRPPHQFHVDQGWKHFRUHSODWH

WHPSHUDWXUHEHJLQVWRLQFUHDVH7KHLQFUHDVHLQWHPSHUDWXUHRIWKHFRUHSODWHDORQJZLWKWKH

SK\VLFDOORDGLQJRIWKHGHEULVHYHQWXDOO\FDXVHWKHSODWHWR\LHOGORFDOO\DOORZLQJGHEULVWRUHORFDWH

LQWRWKHORZHUSOHQXP5DSLGTXHQFKLQJRIWKHGHEULVLVLPSRVHGDVGHEULVHQWHUVWKHSRRORI

ZDWHULQWKHORZHUKHDG,QWHUDFWLRQEHWZHHQWKHGHEULVFRRODQWDQGORZHUKHDGGHWHUPLQHVWKH

HYHQWXDOERLORIIRIWKHUHPDLQLQJFRRODQWDQGWKHUPDOUHVSRQVHRIWKHORZHUKHDG8OWLPDWHO\

IDLOXUHRIWKHORZHUKHDGRFFXUVDWKRXUVIURPVWUDLQ9DULRXVVWDJHVRIFRUHGHJUDGDWLRQDUH

SUHVHQWHGLQ)LJXUHWRLOOXVWUDWHWKHFRUHGHJUDGDWLRQSURJUHVVLRQ

0

250

500

750

1000

1250

1500

0

500

1000

1500

2000

2500

3000

10 14 18 22 26 30 34

0LGSODQH&ODGGLQJ7HPSHUDWXUH.

,Q9HVVHO+\GURJHQ*HQHUDWLRQNJ

7LPHKU

Ring 1

Ring 2

Ring 3

Ring 4

In-Ves. H2

Indicates intact fuel no

longer present at this location
3-64

Figure 3-52 Mark II core degradation and relocation for Case 11

Upon lower head failure, approximately 220 metric tons of corium are ejected to the upper reactor 

cavity within the reactor pedestal. Corium concrete interaction commences as the relocated debris 

within the upper reactor cavity begins to decompose the concrete. Sump drain lines penetrating 

the concrete structure between the upper and lower reactor cavity are assumed to fail 20 minutes 

after debris relocation to the upper reactor cavity. At this time, the sump drain lines are modeled in 

the analysis as fully open, thus allowing debris to relocate to the lower reactor cavity along with 

water. Concrete ablation continues in the lower reactor cavity throughout the remainder of the 

accident sequence, as shown in Figure 3-53. 

The containment pressure response throughout the transient can be seen in Figure 3-54. The 

operation of the RCIC system and actuation of the SRVs result in a slow pressurization of the 

containment prior to suppression pool saturation (see Figure 3-55). Once the pool becomes 

saturated, the containment pressure rapidly increases due to steam released from the vessel and 

the subsequent vaporization of wetwell pool water. Since the containment pressure does not 

reach 15 psig prior to RCIC failure, venting is not performed. However, the post core damage 

venting initiates at 20.3 hours as the containment pressure exceeds the prescribed PCPL of 60 

psig. The containment pressure excursion during core damage fails to achieve 60 psig.

Vessel leakage through the reactor circulation pumps and to a lesser extent steam condensed 

and drained from the drywell has over time increases the water level in the upper reactor cavity, 

as seen in Figure 3-56. At the time of lower head failure, debris comes into contact with the upper 

reactor cavity water and sufficient containment pressure results in actuation of the wetwell vent.

The integral vent flow is provided in Figure 3-54. 

13.3 hr 14.6 hr 17.4 hr 20.0 hr

Water level below BAF Core relocation to lower head Lower plenum dryout Lower head failure

Intact fuel (UO2) and cladding Steam and/or hydrogen (no structure)

Particulate debris Steel support structure

Molten pool Control material

Water
3-65

Figure 3-53 Mark II concrete ablation depth due to MCCI for Case 11

Figure 3-54 Mark II containment pressure and vent flow for Case 11

-20

-17.5

-15

-12.5

-10

-7.5

-5

0 12 24 36 48 60 72

Lowest Cavity Elevation (m)

Time (hr)

Upper Reactor

Cavity Floor

Original

Lower Reactor

Cavity Floor

0

50000

100000

150000

200000

250000

300000

350000

400000

450000

0

10

20

30

40

50

60

70

80

0 12 24 36 48 60 72

Integral Vent Mass Flow (kg)

DW Pressure (psia)

Time (hr)

WW Vent Opens at PCPL

DW Vent Opens at PCPL

WW Vent Closes

at High WW Level


)LJXUH 0DUN,,VXSSUHVVLRQSRROWHPSHUDWXUHIRU&DVH

)LJXUH 0DUN,,FRQWDLQPHQWZDWHUOHYHOVIRU&DVH

300

320

340

360

380

400

420

440

0 12 24 36 48 60 72

Temperature (K)

Time (hr)

Wetwell

Wetwell Pool Saturation

230F

-30

-25

-20

-15

-10

-5

0

0 12 24 36 48 60 72

Containment Water Elevations [m]

Time (hr)

Drywell Upper

Reactor

Cavity

Lower

Reactor

Cavity

Wetwell

(Suppression

Pool)

D

C

V

s


539)/(;LQMHFWLRQLQLWLDWHVDWWKHWLPHRIORZHUKHDGIDLOXUHDQGZDWHULVSHUPLWWHGWRGUDLQIURP

WKHYHVVHOWKURXJKWKHORZHUKHDGIDLOXUHLQWRWKHXSSHUUHDFWRUFDYLW\2QFHWKHVXPSGUDLQOLQHV

IDLOZDWHULVDEOHWRIUHHO\IORZLQWRWKHORZHUUHDFWRUFDYLW\DQGHQWHUWKHZHWZHOO:LWKFRQWLQXRXV

)/(;LQMHFWLRQRIJSPWKHVXSSUHVVLRQSRROOHYHOHYHQWXDOO\H[FHHGVWKHDVVXPHGKLJK

ZDWHUOHYHODQGWKHZHWZHOOYHQWLVLVRODWHG&RQWDLQPHQWSUHVVXUHLVSHUPLWWHGWRSUHVVXUL]HXQWLO

SVLJLVDFKLHYHGDWKRXUVDWZKLFKWLPHWKHGU\ZHOOYHQWOLQHLVRSHQHG

&RQWDLQPHQWDWPRVSKHUHWHPSHUDWXUHDQGSUHVVXUHUHPDLQVXIILFLHQWO\ORZWRSUHYHQWGU\ZHOO

KHDGIDLOXUHVHH)LJXUH([YHVVHOJDVJHQHUDWLRQDQGKHDWWUDQVIHUIURPWKHGHEULVFDXVH

WHPSHUDWXUHVLQWKHFRQWDLQPHQWWRULVHZKLOH)/(;LQMHFWLRQSUHYHQWVVLJQLILFDQWWHPSHUDWXUH

LQFUHDVH

)LJXUH 0DUN,,FRQWDLQPHQWDWPRVSKHUHWHPSHUDWXUHVIRU&DVH

'XHWRWKHWHPSHUDWXUHWUDQVLHQWDQGIXHOIDLOXUHILVVLRQSURGXFWVDUHHYHQWXDOO\UHOHDVHGIURPWKH

FRUHDQGWUDQVSRUWHGWKURXJKRXWWKHSULPDU\FRQWDLQPHQWXQWLOYHQWRSHUDWLRQVSHUPLWUHOHDVHWR

WKHHQYLURQPHQW7KHRQVHWRIILVVLRQSURGXFWUHOHDVHRFFXUVRQFHIXHOFODGJDSIDLOVDW

KRXUV7KLVWLPLQJFRLQFLGHVZHOOZLWKWKHRQVHWRIYLJRURXVR[LGDWLRQRIWKHFODGGLQJ$VIXHO

WHPSHUDWXUHVUDSLGO\LQFUHDVHILVVLRQSURGXFWGLIIXVLRQHQKDQFHVDQGYRODWLOHILVVLRQSURGXFWVDUH

UHOHDVHGIURPWKHIXHO&HVLXPDQGLRGLQHZKLFKDUHVLJQLILFDQWFRQWULEXWRUVWRSXEOLFULVNDUH

GLVWULEXWHGDVVHHQLQ)LJXUHDQG)LJXUH

300

400

500

600

700

800

900

0 12 24 36 48 60 72

Temperature (K)

Time (hr)

Lower Reactor Cavity

Upper Reactor Cavity

Drywell
3-68

Figure 3-58 Mark II cesium fission product distribution for Case 11

Figure 3-59 Mark II iodine fission product distribution for Case 11

Oxidation results in a rapid temperature increase such that the release of cesium and iodine from 

the fuel is nearly complete within 2.5 hours of the initial gap release; more than 95% of the original 

inventory is predicted to be released from the fuel at this time. Released fission products that are 

not deposited within the RCS enter containment primarily through the open SRV. Fission product 

1.0E-05

1.0E-04

1.0E-03

1.0E-02

1.0E-01

1.0E+00

0 12 24 36 48 60 72

Fraction of Initial Cesium Core Inventory

Time (hr)

Released 

from fuel

Deposited in

Containment

Airborne in

Containment

Released to

Environment

Deposited/Airborne

in RPV

1.0E-05

1.0E-04

1.0E-03

1.0E-02

1.0E-01

1.0E+00

0 12 24 36 48 60 72

Fraction of Initial Iodine Core Inventory

Time (hr)

Released 

from fuel

Deposited in

Containment

Airborne in

Containment

Released to

Environment

Deposited/Airborne

in RPV
3-69

scrubbing occurs as steam is passed through the suppression pool. Uncaptured fission products 

are transported through the wetwell/drywell vacuum breakers to the drywell vapor space, where 

suspended aerosols continue to deposit over time. Airborne fission products at the time of vent 

operation are available for environment release.

Once the lower head has failed, fission products are transported from the reactor vessel to 

containment through the lower head failure. Pool scrubbing is performed, albeit less efficiently, 

through the downcomer vents while differential pressures between the drywell and wetwell are 

sufficient to clear the downcomer vents. The differential pressure is a product of the MCCI gas 

generation and the open wetwell vent line. However, with the assumed failure of the in-pedestal 

drain lines after 20 minutes, suppression pool bypass occurs and pool scrubbing through the 

downcomer vents terminates now that flow through the upper/lower reactor cavity is possible. 

After suppression pool bypass has occurred, the late in-vessel release, characterized by the 

revaporization of RCS deposits, is the dominant contributor to the environmental release. For 

Case 11, the operator action to open an SRV at the time of RCIC failure provided a continuous 

release path from the RCS to the containment throughout core degradation, permitting relatively 

large amounts of fission products to be released to the containmaint. By enhancing fission product 

release to the containment, the fission product inventory deposited within the RCS is reduced.

This limits the available fission product release from the deposited fission product in the RCS by 

suppressing the total deposited mass available for release, and similarly, the available decay heat 

to promote revaporization. Therefore, for Case 11, suppression pool bypass is not observed to 

produce a significant impact to the environmental releases. The initial vent operation and resulting 

sudden release are followed by revaporization of deposited cesium-iodide from the RCS where 

vapor pressures are sufficient. Iodine is subjected to greater revaporization from the RCS than is 

observed for cesium, as is discussed in Section 3.3.1.1 . A slow liberation of total iodine, and to a 

lesser extent cesium, is observed in RPV inventory seen in Figure 3-59 and corresponding 

environmental release.

Suppression pool water, due to the continual RPV FLEX injection of 500 gpm, eventually exceeds 

the assumed upper instrumentation level range and the wetwell vent line is closed at 49.9 hours, 

and the containment pressure is permitted to repressurize to 60 psig. At 68.8 hours environmental 

releases continue for the remainder of the analysis with the actuation of the drywell vent line.

3.3.2.2 Accident Progression without Water Addition

In contrast to the Mark II analysis with water addition, provided in Section 3.3.2.1 , Case 1 is 

presented to demonstrate system response in the absence of FLEX injection. Additionally, RPV 

depressurization after RCIC failure is not performed. Table 3-3 lists the specifics of the sequence 

employed for Case 1.

RPV pressure and water level responses are presented for Case 1 in Figure 3-60 and Figure 

3-61, respectively. As anticipated, the pressure response remains identical to that observed in

Case 11 until 8.4 hours when RCIC fails. In Case 1, pressure control, initiated at 1 hour, continues

and the RPV pressure is maintained between 200-400 psig. Without RCIC maintaining reactor

water level, oxidation of zirconium increases fuel and vapor temperatures. Transitioning from

intact components to particulate debris, fuel and in-core structures relocate downward, settling

upon the core plate. As heat is conducted from the debris, the core plate temperature rises and

the plate fails at 15 hours. Debris enters the pool in the lower head and the resulting steam

generation and pressure event are clearly seen in Figure 3-60 even though an SRV is open. As

the core degradation continues and vapor temperatures further increase, the internal components
3-70

of the operating SRV achieve temperatures in excess of 900K. Loss of mechanical strength due 

to excessive temperatures is assumed to fail the operating SRV in the open position, which 

permits the RPV to depressurize at 16.8 hours. 

Core degradation, unlike that observed in Case 11, continues after lower head failure as degraded 

fuel in core ring two becomes particulate debris and relocates to upper reactor cavity. This 

process increases the observed cumulative mass ejected from the vessel from the 220 metric 

tons observed in Case 11 to 248 metric tons observed in Case 1. Figure 3-62 outlines various 

core degradation states throughout the transient. 

Initial gap release and hydrogen generation occur at approximately 13 hours. SRV actuations 

allow airborne fission products to be released to the suppression pool until the lower head fails, 

whereby fission products are transported directly from the vessel to the containment. Downcomer 

vent scrubbing halts once the upper reactor cavity sump drain lines fail. Fission product 

distributions for cesium and iodine are provided in Figure 3-63 and Figure 3-64, respectively. Late 

in-vessel release significantly contributes to the environment releases as a larger fraction of 

fission products were observed to deposit in the RCS in comparison to those in Case 11. Unlike 

Case 11, suppression pool bypass timing impacts a greater fraction of the total fission product 

inventory released to the environment. Release fractions of total cesium and iodine inventories to 

the environment are calculated as 2.46E-2 and 1.98E-1, respectively. 

During the transient, the containment pressure is observed to slowly rise while the suppression 

pool has yet to saturate. SRV operations and hydrogen generation produce rapid pressurization of 

the containment system once saturation of the suppression pool occurs. As seen in Figure 3-65, 

the drywell pressure reaches 60 psig at 22.8 hours and the wetwell vent is actuated. Following 

lower head failure and the expulsion of debris, boiloff of the pool water accumulated in the upper 

reactor cavity as well as the progression of MCCI increases local temperatures within the upper 

reactor cavity, and eventually the lower reactor cavity following sump drain line failure. Pool water 

and debris are then transferred to the lower reactor cavity; where without the addition of FLEX 

injection, the containment temperatures observed in Figure 3-66 steadily increase. The debris bed 

becomes uncovered at 26.0 hrs, as shown in Figure 3-67. The wetwell vent line remains open for 

the remainder of the analysis.
3-71

Figure 3-60 Mark II RPV pressure for Case 1

Figure 3-61 Mark II RPV water level for Case 1

0

10

20

30

40

50

60

0 2 4 6 8 10 12 14 16 18 20

RPV Water Level (ft)

Time (hr)

Top of Active Fuel (TAF)

Bottom of Active Fuel (BAF)

Automatic RCIC 

Actuation

Start of Fuel Relocation

into Lower Head

RCIC Failure
3-72

Figure 3-62 Mark II core degradation and relocation for Case 1 

Figure 3-63 Mark II cesium fission product distribution for Case 1

14.0 hr 15.4 hr 19.8 hr 22.2 hr 72.0 hr

Water level below BAF Lower plenum dryout Final state

Core relocation to lower head Lower head failure

Intact fuel (UO2) and cladding Steam and/or hydrogen (no structure)

Particulate debris Steel support structure

Molten pool Control material

Water

1.0E-05

1.0E-04

1.0E-03

1.0E-02

1.0E-01

1.0E+00

0 12 24 36 48 60 72

Fraction of Initial Cesium Core Inventory

Time (hr)

Released 

from fuel Deposited in

Containment

Airborne in

Containment

Released to

Environment

Deposited/Airborne

in RPV
3-73

Figure 3-64 Mark II iodine fission product distribution for Case 1

Figure 3-65 Mark II containment pressure and vent flow for Case 1

1.0E-05

1.0E-04

1.0E-03

1.0E-02

1.0E-01

1.0E+00

0 12 24 36 48 60 72

Fraction of Initial Iodine Core Inventory

Time (hr)

Released 

from fuel

Deposited in

Containment

Airborne in

Containment

Released to

Environment

Deposited/Airborne

in RPV

0

100000

200000

300000

400000

500000

600000

700000

0

10

20

30

40

50

60

70

80

0 12 24 36 48 60 72

WW Vent Integral Mass Flow (kg)

DW Pressure (psia)

Time (hr)

WW Vent Opens at PCPL


)LJXUH 0DUN,,FRQWDLQPHQWDWPRVSKHUHWHPSHUDWXUHVIRU&DVH

)LJXUH 0DUN,,FRQWDLQPHQWZDWHUOHYHOVIRU&DVH

300

500

700

900

1100

1300

1500

1700

1900

0 12 24 36 48 60 72

Temperature (K)

Time (hr)

Lower Reactor Cavity

Upper Reactor Cavity

Drywell

-30

-25

-20

-15

-10

-5

0

0 12 24 36 48 60 72

Containmnet Water Elevations (m)

Time (hr)

Drywell Upper

Reactor

Cavity

Lower

Reactor

Cavity

Wetwell

(Suppression

Pool)

D

C

V

s
3-75

3.3.2.3 Source Term and Containment Response

The U.S. Mark II fleet employs several cavity designs. Should core degradation result in lower 

head failure, containment response and ultimately fission product release could vary among these 

designs. Regardless of containment design, a reasonable assumption is fission product transport 

prior to lower head failure would undergo similar decontamination within containment; therefore, 

differences are anticipated to occur after lower head failure among the various containment 

configurations, see Figure 3-68. 

Figure 3-68 Mark II containment configurations

The upper reactor cavity floor, drywell floor, downcomer vents (see for example, Figure 3-67), and 

drain lines (not shown) are the components which comprise the drywell pressure boundary 

separating the atmospheres of the drywell and wetwell. Debris exiting the vessel accumulates 

within the upper reactor cavity and interacts with the upper reactor cavity floor and penetrations 
3-76

within, which compromises the integrity of the pressure boundary. Should the pressure boundary 

between the drywell and wetwell fail, atmospheric material can be freely exchanged between the 

drywell and wetwell atmospheres. The loss of the pressure boundary prevents the downcomer 

vents from passing drywell atmosphere through the suppression pool; therefore, vapor 

condensation from drywell atmosphere bubbled through the pool and suppression pool scrubbing 

are lost. This event is commonly referred to as suppression pool bypass. 

Three distinct fission product transport phases are characterized in Figure 3-69 to emphasize the 

importance of suppression pool bypass. The first phase occurs when the RCS pressure boundary 

is maintained and the predominant fission product releases from the vessel occur through open 

SRVs. The second phase initiates once lower head failure has occurred and debris is transported 

to the upper reactor cavity. Fission products released from RCS enter the drywell atmosphere 

predominantly through the lower head failure. In additional, MCCI generated aerosols and gases 

enter the drywell atmosphere as well. The pressure differential between the drywell and wetwell 

results in the downcomer vents clearing and atmospheric material is transmitted to the 

suppression pool. The third phase follows suppression pool bypass, where flow between the 

atmospheres of the wetwell and drywell becomes prevalent. Aerosol and vapors released from 

submerged debris do undergo scrubbing; however, volatile fission products, such as cesium and 

iodine, have predominantly been released from the fuel prior to lower head failure.

Figure 3-69 Mark II fission product transport paths during accident phases

The analyses presented assume the timing of suppression pool bypass is the most significant 

contribution to containment decontamination. Rather than attempt to address each containment 

configuration, suppression pool bypass timing was modified as a surrogate for direct simulation of 

each containment configuration. This was performed under the limitation of available Mark II 

models for severe accident analysis. Furthermore, the in-pedestal region beneath the upper 

reactor cavity is flooded in all other configurations. To address this difference, analyses with and 

without lower reactor cavity flooding are presented in the sensitivities discussed below. The 

sensitivities presented use Case 1 due to the large fission product masses deposited in the RCS 
3-77

as compared to Case 11. Suppression pool bypass and cavity configurations will influence the 

environmental release during the late in-vessel release phase of the accident.

The first sensitivity presented increases the time to suppression pool bypass for the default Mark 

II model. In the original model, suppression pool bypass is assumed to occur 20 minutes after 

debris enters the upper reactor cavity as a result of in-pedestal sump drains failing due to debris 

contact. This failure criterion was suspended and debris was permitted to remain in the upper 

reactor cavity until the ablation depth due to MCCI exceeded the total thickness of the cavity floor.

This case is presented as Case 1 with URC MINALT (upper reactor cavity ablation depth reaching 

minimum altitude of the concrete).

A set of sensitivities is presented that employ a modified Mark II model to investigate a flooded 

lower reactor cavity. The lower reactor cavity of the representative Mark II model is a partially 

filled, dry volume. This cavity volume was increased by extending the volume downward until the 

base altitude of the lower reactor cavity agreed with the bottom of the wetwell volume. The lower 

reactor cavity volume was combined with wetwell control volume to create a contiguous 

representation of the suppression pool, a single, well-mixed pool region defining the in-pedestal 

and suppression pool water. The original suppression pool water level in the Mark II 

representative model was maintained. This configuration is presented in Figure 3-69 (see Figure 

3-9 for comparison with the original model). The three sensitivities performed with the modified

containment model include the following:

• Case 1 LRC MINALT – Case 1 was performed with the minimum ablation rule enabled.

• Case 1 LRC 20 min delay – Case 1 was performed with the default 20-minute delay after

debris enters the upper reactor cavity before suppression pool bypass.

• Case 1 LRC 0 min delay – Case 1 was performed; however, the debris relocating from

the reactor was passed directly to the suppression pool/lower reactor cavity, producing

an instant suppression pool bypass at the time of lower head failure.

The environmental releases of cesium and iodine are presented in Figure 3-70 and Figure 3-71

for the default Mark II model and the modified Mark II model with the flooded lower reactor cavity, 

respectively. Environment release initiates at the time of wetwell venting in the analyses 

performed. In the case of the base analysis (Case 1) and the Case 1 sensitivity (URC MINALT), 

significant reductions in the release of cesium and iodine result. The increased duration of 

downcomer vent scrubbing through prolonging the bypass of the suppression pool, a significant 

reduction in the released cesium and iodine are realized.

The sensitivities performed with the Mark II model with the flooded lower in-pedestal reactor cavity 

modification produce similar reductions in the environment release of cesium and iodine. These 

analyses deviate from the base case as the vent line operation, initiating at PCPL, was reached 

prior to lower head failure10. Regardless of chronology of vent line operation and lower head 

failure, a significant reduction in the overall release of cesium and iodine is released by extending 

the suppression pool bypass timing.

The timing of suppression pool bypass is considered uncertain and assumed to be the largest 

significant variation among the different containment configurations. While no qualitative 

discussion is presented regarding the variation of suppression pool bypass timing for each design, 

10

 The modified Mark II model increases the total suppression pool inventory, which prolongs RCIC operations. These 

differences ultimately permit the first significant pressurization event to exceed the PCPL. The Mark I and the

modified Mark II model pressure responses are similar (see the Figure 3-16 for the Mark I containment pressure 

response).


HDUO\VXSSUHVVLRQSRROE\SDVVDQGDSURWUDFWHGVXSSUHVVLRQSRROE\SDVVZDVSHUIRUPHGIRUWKH

0DUN,,DQGDPRGLILHG0DUN,,PRGHO5HGXFWLRQVLQWRWDOPDWHULDOUHOHDVHGZHUHZLWKLQDQRUGHURI

PDJQLWXGHIRUWKHVHQVLWLYLWLHVSHUIRUPHG,QFRPSDULVRQWKHUDQJHRIUHVXOWVREVHUYHGIRUWKH

&DVHVHQVLWLYLWLHVLVSUHVHQWHGDORQJZLWKWKHUHPDLQLQJ0DUN,,DQDO\VHVLQ)LJXUH7KH

UHOHDVHVDUHFRPSDUDEOHWRWKRVHREVHUYHGDPRQJWKHVFHQDULRVLQYHVWLJDWHGDQGDUHWKHUHIRUH

FRQVLGHUHGQRPRUHVLJQLILFDQWWKDQWKHYDULDWLRQVSHUIRUPHG,IWKHVHVHQVLWLYLWLHVVKRZHGPRUH

VLJQLILFDQFHDGGLWLRQDOLQYHVWLJDWLRQVZRXOGEHMXVWLILHG

)LJXUH 0DUN,,HQYLURQPHQWUHOHDVHVHQVLWLYLW\WRVXSSUHVVLRQSRROE\SDVVWLPLQJ

1.00E-04

1.00E-03

1.00E-02

1.00E-01

1.00E+00

0 12 24 36 48 60 72

Environment Release Fraction(-)

Time (hr)

Cesium - Case 1

Cesium - Case 1 URC MINALT

Iodine - Case 1

Iodine - Case 1 URC MINALT

Lower Head Failure

Suppression Pool Bypass

(Case 1 URC MINALT)

Suppression Pool Bypass

(Case 1)


)LJXUH 0DUN,,ZLWKIORRGHGORZHUUHDFWRUFDYLW\VHQVLWLYLW\WRVXSSUHVVLRQSRRO

E\SDVVWLPLQJ

7KHK\GURJHQGLVFXVVLRQSUHVHQWHGEHORZLQFRUSRUDWHV&DVHSIURP7DEOHWKH3&3/ZDV

FRUUHFWHGIURPSVLJWRSVLJ7KHZHWZHOOYHQWOLQHLVLQLWLDOL]HGZLWKQRQFRQGHQVLEOHPROH

IUDFWLRQVUHSUHVHQWDWLYHRIVWDQGDUGDLU1DQG2)RUWKH&DVHSWUDQVLHQW

DSSUR[LPDWHO\NJRIK\GURJHQLVSURGXFHGLQYHVVHOZKLOHWKHWRWDOPDVVRIK\GURJHQ

JHQHUDWLRQLVFDOFXODWHGDVDSSUR[LPDWHO\NJ)LJXUHVKRZVWKHLQWHJUDOK\GURJHQ

JHQHUDWHGERWKLQYHVVHODQGH[YHVVHOFRPSDUHGWRWKHK\GURJHQPDVVUHPDLQLQJLQ

FRQWDLQPHQW$WPRVSKHULFPROHIUDFWLRQVDUHSUHVHQWHGLQWKURXJK)LJXUHGHSLFWLQJWKH

GLVWULEXWLRQRIUHOHYDQWJDVHV&DUERQPRQR[LGHJHQHUDWHGGXULQJ0&&,LVDQDGGLWLRQDO

FRPEXVWLRQFRQFHUQDQGLVDGGHGWR)LJXUHWKURXJK)LJXUH&RPEXVWLEOHFRQFHQWUDWLRQV

DUHSUHFOXGHGZLWKLQFRQWDLQPHQWSULRUWRYHQWOLQHRSHUDWLRQVGXHWRKLJKFRQFHQWUDWLRQVRI

QLWURJHQDQGVWHDPHYHQWKRXJKVLJQLILFDQWDPRXQWVRIK\GURJHQDUHSUHVHQW2QFHWKHZHWZHOO

DQGHYHQWXDOGU\ZHOOYHQWOLQHDFWXDWLRQVDUHSHUIRUPHGK\GURJHQDQGFDUERQPRQR[LGHDUH

UDSLGO\GLVSHUVHGIURPFRQWDLQPHQWDWDQGKUV:LWKRXWFRQWDLQPHQWIDLOXUHK\GURJHQLVQRW

GLVWULEXWHGWRWKHUHDFWRUEXLOGLQJ

0.0001

0.001

0.01

0.1

1

0 12 24 36 48 60 72

Environment Release Fraction (-)

Time (hr)

Cesium - 0 min delay

Cesium - 20 min delay

Cesium - LRC MINALT

Iodine - 0 min delay

Iodine - 20 min delay

Iodine - LRC MINALT

Lower Head Failure

Suppression Pool Bypass

(LRC MINALT)

Suppression Pool Bypass

(20 min delay)


)LJXUH 0DUN,,K\GURJHQJHQHUDWLRQDQGWUDQVSRUWIRU&DVHS

)LJXUH 0DUN,,GU\ZHOOJDVGLVWULEXWLRQIRU&DVHS

0

500

1000

1500

2000

2500

3000

0 12 24 36 48 60 72

+0DVVNJ

7LPHKU

Generation

Containment

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

1

0 12 24 36 48 60 72

'U\ZHOO0ROH)UDFWLRQ

7LPHKU

H2O

O2

N2

H2

CO

CO2


)LJXUH 0DUN,,ZHWZHOOJDVGLVWULEXWLRQIRU&DVHS

)LJXUH 0DUN,,ZHWZHOOYHQWJDVGLVWULEXWLRQIRU&DVHS

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

1

0 12 24 36 48 60 72

:HWZHOO0ROH)UDFWLRQ

7LPHKU

H2O

O2

N2

H2

CO

CO2

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

1

0 12 24 36 48 60 72

9HQW/LQH&90ROH)UDFWLRQ

7LPHKU

H2O

O2

N2

H2

CO

CO2
3-82

Releases to the environment observed from the accidents analyzed (see Table 3-3) are presented 

in Figure 3-76. The summary of the main parameters for the Mark II analyses are provided in 

Table 3-7 (see Table 3-3 for details of the scenarios). The observations made in Section 3.3.1.3 

for the Mark I accident analyses are largely applicable to the Mark II accident analyses. Similarly, 

FLEX injection appears to reduce overall releases of cesium whereas drywell venting produces 

larger environmental releases, as would be anticipated.

Noticeably, cases representing MSLCR for the Mark II model are not necessarily dominant 

release cases as was observed in Mark I results. Unlike the Mark I sequence of events, MSLCR 

does not produce head flange failure in the Mark II model; therefore the environment release path 

remains the designated vent line in the analyses performed with the Mark II model11. 

Figure 3-76 Mark II cesium environmental release fraction

Figure 3-77 presents the cesium release observed for the cavity sensitivities performed as well as 

the calculation matrix for the Mark II analyses (M_I refers to Mark I and M_II refers to Mark II). The 

designation of PCPL as 60 psig was changed to 45 psig for several of the calculations provided.

While the final magnitude of the release remains reasonably comparable, it should be reflected 

upon that the event timings, in particular the vent actuation prior to lower head failure, does 

become prominent. Cesium releases remain below 4% of total inventory for the analyses 

performed.

11 The two models do differ in regard to drywell nodalization, where the Mark I model segregates the drywell

atmosphere space into three separate volumes, the drywell in the Mark II model is a single well mixed receptor. 

Note in addition to the nodalization variations, the vapor space volume in Mark II is larger than found at Mark I.
3-83

Figure 3-77 Cesium release fractions for Mark II cases and sensitivities as well as 

corresponding Mark I cases

Table 3-7 Summary of main parameters for Mark II analysis
3-84

3.4 Consideration of Uncertainty

MELCOR is considered a state-of-the-art system-level integral code for severe accident modeling 

and analysis, and it has reached a reasonably high level of maturity over the years as evidenced 

from its wide acceptability and its broad range of applications. MELCOR embodies the current 

state of knowledge of severe accident phenomena. However, as for any system-level codes with 

similar capabilities, it is important to recognize that there are phenomenological uncertainties in 

severe accident progression that have a direct bearing on modeling uncertainties. Moreover, it is 

important to understand the compounding effect of various uncertainties on the output parameters 

of interest (e.g., hydrogen generation, release of fission products to the environment, etc.). 

Given the state of severe accident modeling and the residual uncertainties therein, an “adequate 

for purpose” approach is to do bounding analysis reflecting best estimate outcome supplemented 

by some measures of uncertainties. The bounding values of the output parameters of interest are 

then compared to those considered acceptable from a safety margin standpoint to determine if 

further reduction of residual uncertainties is warranted. In the remainder of this section, a brief 

discussion is provided regarding some of the more important modeling uncertainties in MELCOR 

and potential implications of such uncertainties. Also, a brief discussion of uncertainties in

reference to mitigation systems modeling in MELCOR is provided.

The in-vessel melt progression modeling in MELCOR starting with the loss of intact core geometry 

to clad oxidation, in-vessel hydrogen generation, molten core relocation to lower plenum, and 

subsequent lower head failure are based on small scale experiments which were conducted with 

the primary objective of gaining an understanding of these phenomena in relation to the 

observation and experience from plant accidents such as Three Mile Island. 

MELCOR, for example, has a parametric model for evaluating fuel mechanical response whereby 

a temperature-based criterion is used to define the threshold beyond which normal (“intact”) fuel 

rod geometry can no longer be maintained, and the core materials at a particular location collapse 

into particulate debris. The relocation of molten and particulate debris to the lower plenum is 

controlled by the relocation time constant parameter in MELCOR. This parameter is used as a 

surrogate for the broad uncertainty in the debris relocation rate into water in the lower head. The 

choice of relocation time constant affects the potential for debris coolability in the lower head 

(faster relocation rates decrease coolability; slower rates improve coolability). These and other 

related in-vessel melt progression modeling attributes in MELCOR affect the timing of lower head 

failure as well as the characteristics of melt (temperature, mass, and composition) exiting the 

vessel which provide the initial and boundary conditions for ex-vessel melt progression. These 

attributes also affect the amount of hydrogen generation – a parameter of interest from the 

containment integrity standpoint. The current state of BWR modeling in MELCOR does not 

consider the effects of structures (such as control rod drive mechanisms) beneath the lower head. 

Such structures can provide an energy sink to materials that relocate to the lower head, thus 

potentially delaying the lower head failure. Also, the core materials can transfer heat and 

potentially freeze onto these structures as they are ejected from the vessel into the cavity.

As in the case of in-vessel melt progression, the ex-vessel phenomenological modeling is based 

on experiments which were conducted to gain an understanding of melt spreading on the drywell 

floor, debris quenching in the presence of water, and molten core-concrete interaction, among 

others. After the core debris is released from the reactor vessel lower head, it flows out of the 

reactor pedestal onto the main drywell floor. The precise conditions under which core debris 

would flow out of the pedestal and across the drywell floor are uncertain. These uncertainties are 

captured in MELCOR in a parametric manner. Phenomenological models are being continuously 
3-85

updated in MELCOR based on data generated through international research programs. The 

improved models are expected to reduce the uncertainties to some extent.

Partitioning the initial core inventory of radionuclides (cesium and iodine in particular) among 

certain allowable chemical forms (for release and transport) is performed within MELCOR input 

files that define the initial spatial mass distribution of each chemical species and its associated 

decay heat. Changes to the mass fractions assumed for a particular chemical group directly affect 

the mass fractions of other chemical groups. Due to the complexity of this modeling approach, five 

alternative sets of MELCOR input files are used to bound uncertainties by spanning the range of 

plausible combinations of chemical forms of key radionuclide groups.

Gaseous iodine remains another source term issue with uncertainties, especially with respect to 

long-term radioactive release mitigation issues after the comparatively much larger airborne 

aerosol radioactivity has settled from the atmosphere. Mechanistic modeling of gaseous iodine 

behavior is a technology still under development with important international research programs 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.

Several other sources of uncertainties, not specifically discussed here, can have an impact on 

MELCOR results. Moreover, there are uncertainties in modeling various mitigation features (e.g., 

reactor core injection cooling or RCIC performance, drywell water addition mode and 

effectiveness, suppression pool decontamination factor, and in-containment radionuclide retention 

factor). Given these various sources of uncertainties, it is not uncommon to find an order of 

magnitude or more variation in the MELCOR prediction of the source term. 

An uncertainty analysis for a long term station blackout was recently performed for Peach Bottom 

following the completion of SOARCA (a Surry uncertainty analysis is currently in progress). The 

key MELCOR model parameters for Peach Bottom included the following.

• Zircaloy melt breakout temperature

• Molten clad drainage rate

• Fuel failure criterion (transformation of intact fuel to particulate debris)

• Radial debris relocation time constants

• Debris Pool Interface Heat Transfer

• Debris lateral relocation—cavity spillover criteria and spreading rate

• Chemical forms of iodine and cesium (I2, CH3I, CsI, CsOH, and Cs2MoO4)

The results of the SOARCA uncertainty analysis confirmed that prediction of the source term can 

have an order of magnitude or more variation. 

The analyses presented in this report did not consider a detailed uncertainty analysis. Limited 

sensitivity analysis was carried out to assess the range of MELCOR results and to further confirm 

the bounding range of source terms. The run matrix given in Table 3-2 contains mainly variations 

in the accident boundary conditions such as operation of RCIC, suction source of RCIC injection, 

etc. In addition, the run matrix also includes sensitivity to the fractional open area of a thermally 

seized SRV, which was considered an important parameter in the SORACA uncertainty analysis.

This parameter was varied to observe its effect on the possibility of main steam line creep rupture

(MLSCR). In all the MELCOR calculations, MSLCR only occurred by disabling the SRV failure or 

intentional depressurization. MSLCR scenarios represent the highest environmental releases due 

to early bypass of the suppression pool. The results of the calculations showed that there are 
3-86

variations in the source term by as much as an order of magnitude, especially when the releases 

are low. 

3.5 Accident Progression Summary and Conclusions

The MELCOR analysis investigated detailed accident progression, source terms, and the 

containment response following an ELAP subject to appropriate initial and boundary conditions for 

representative Mark I and Mark II containment designs. The run matrix for the Mark I analysis 

included sensitivities to the following main parameters:

• Mode of venting (e.g., WW or DW first, vent cycling)

• Status of RPV depressurization (e.g., SRV open pre-core damage, SRV stuck open)

• Mode of FLEX injection (either to the RPV or the DW at lower head failure)

• Water injection control (e.g., water management by throttling flow at high level in

wetwell)

The run matrix for the Mark II analysis included a subset of the Mark I runs based on the insights 

from the Mark I MELCOR calculations. The base case Mark II MELCOR model had a dry lower 

cavity without in-pedestal downcomers that was chosen mainly because of the availability of the 

MELCOR model. However, sensitivities were performed to examine the impact of the pedestal 

and lower cavity designs among the fleet by modifying the base model (e.g., replace the concrete 

lower cavity with water). 

The following are the main observations from the calculations:

• A combination of venting and water injection is required to prevent containment failure

and is a beneficial strategy for mitigating radiological releases.

• In all MELCOR calculations for the Mark I analysis, containment venting after core

damage occurs well before the lower head failure and injection of water. For this reason,

the sudden release at the time of venting is only sensitive to core degradation and

fission product transport and deposition rather than the late water injection at the time of

lower head failure.

• Pre-core damage anticipatory venting reduces the containment base pressure at the

time of core damage and results in a delay when post core damage venting is required.

This time delay affects fission product behavior inside the RPV and containment and

impacts the sudden release at the time of venting.

• Creep rupture of the main steam line seems unlikely if the reactor pressure is maintained

low. For the cases that the creep ruptured was forced by disabling the stuck open SRV

model (either through high temperature or excessive number of cycles), the results show

the highest releases to the environment. The failure of the main steam line results in

failure of the containment (opening of the upper drywell head) and bypass of the

suppression pool. The failure also results in migration of hydrogen to the refueling bay of

the Mark I containment, and consequent hydrogen combustion and release of fission

products directly to the environment.

• Addition of water either into the RPV or the drywell has the benefit of cooling the core

debris and containment atmosphere and can prevent the over-temperature failure of the

upper drywell head. For the Mark I analysis with water addition, the maximum structure
3-87

temperatures at the drywell upper head or the drywell liner near the elevation of the 

drywell vent remains below 500o

F. 

• For the Mark I analysis, the presence of water in the pedestal and lower drywell can cool

the debris and delay liner melt through. However, without water injection at the time of

lower head failure, the debris eventually heats up and contacts the drywell liner, leading

to its failure. With water addition, liner melt through is averted.

• Containment venting is efficient in purging the hydrogen and non-condensibles from the

containment. Water injection is also helpful in maintaining a steam inerted atmosphere,

which can preclude an energetic hydrogen combustion.

• The calculations show that the environmental releases from the Mark II containment are

in general comparable to or lower than those from the Mark I containment.

• For the Mark II analysis, additional analysis was performed to investigate different lower

cavity configurations by modifying the base model. The environmental releases are

within the range of source terms predicted based on the variations in the scenario

boundary conditions.

• In the present analysis, hydrogen migration outside the containment through nominal

leakage pathways does not lead to accumulation of combustible mixtures in the reactor

building. Therefore, a hydrogen burn is not predicted in the reactor building as long as

the containment does not fail due to drywell head failure or liner melt-through.

3.6 References for Chapter 3

1. U.S. Nuclear Regulatory Commission, “Recommendations for Enhancing Reactor Safety

in the 21st Century, The Near-Term Task Force Review of Insights from the Fukushima

Dai-ichi Accident,” July 2011.

2. U.S. Nuclear Regulatory Commission, “Consideration Of Additional Requirements For

Containment Venting Systems For Boiling Water Reactors With Mark I and Mark II

Containments,” SECY-12-0157, November 2012.

3. “MELCOR Computer Code Manuals, Vol. 1: Primer and Users’ Guide, Version 2.1.6840,”

SAND 2015-6691 R, Sandia National Laboratories, August 2015 (ADAMS Accession No.

ML15300A479).

4. U.S. Nuclear Regulatory Commission, “Order to Modify Licenses with Regard to Reliable

Hardened Containment Vents Capable of Operation Under Sever Accident Conditions,”

Order EA-13-109, June 2013.

5. U.S. Nuclear Regulatory Commission, D. Chanin, M. L. Young, J. Randall, “Code Manual

for MACCS2: User’s Guide,” NUREG/CR-6613, Vol. I (1998).

6. U.S. Nuclear Regulatory Commission, “State-of-the-Art Reactor Consequence Analyses

(SOARCA) report”, NUREG-1935, January 2012.

7. Sandia National Laboratories, “Fukushima Daiichi Accident Study (Status as of April

2012),” SAND2012-6173, July 2012.
3-88

8. Shaffer, C.J., et al., “Integrated Risk Assessment for the LaSalle Unit 2 Nuclear Power

Plant,” NUREG/CR-5305 Volume 3, Nuclear Regulatory Commission: Washington DC,

October 1992.

9. Kelly, D.L., et al., “An Assessment of BWR Mark-II Containment Challenges, Failure

Modes, and Potential Improvements in Performance,” NUREG/CR-5528, Nuclear

Regulatory Commission, Washington DC, July 1990.

10. U.S. Nuclear Regulatory Commission, Filtering Strategies Rulemaking Public Meeting,

December 12, 2013 (ADAMS Accession No. ML13357A794).

11. Bixler N., et al. “State-of-the-Art Reactor Consequence Analyses Project Volume 1: Peach

Bottom Integrated Analysis,” NUREG/CR-7110 Vol. 1, Rev. 1, Nuclear Regulatory

Commission, Washington DC, May 2013
4-1

4 OFFSITE CONSEQUENCE ANALYSIS

This section documents the offsite consequence analyses of the accident progression cases 

discussed in the MELCOR accident analysis section. The MELCOR Accident Consequence Code 

System (MACCS) was used to calculate offsite doses and land contamination, and their effect on 

members of the public with respect to individual early and latent cancer fatality risk, land 

contamination areas, population dose, and economic costs. MACCS was selected as the 

consequence analysis tool for this project because it is NRC’s code for performing offsite 

consequence analyses for severe accident risk assessments.

This section begins with a general description of MACCS and is then followed by a discussion of 

the modeling approach for the Mark I and Mark II source terms. The discussion of the modeling 

approach begins with a description of the radionuclide release modeling including the source term 

binning strategy. The discussion continues with a description of the calculational grid; site data; 

meteorological data; atmospheric transport; early, intermediate, and long-term phase exposure 

pathways; protective actions and costs; dosimetry; and health effects. The results are then 

presented and explained. These results are used to estimate the relative public health risk 

reduction associated with the various CPRR alternatives. This section then continues with a 

description of the sensitivity analyses and finally concludes with a discussion of the major insights 

gained from this effort.

4.1 MACCS Conceptual Models

The MACCS code was developed for NRC to evaluate offsite consequences from a hypothetical 

release of radioactive materials into the atmosphere. The code is used as a tool to assess the risk 

and consequences associated with accidental releases of radioactive material into the 

atmosphere in probabilistic risk assessment studies. The code models atmospheric transport and 

dispersion, emergency response and long-term phase protective actions, exposure pathways, 

health effects, and economic costs. While MACCS models consequences of airborne releases 

depositing onto water bodies, MACCS does not model transport and dispersion of aqueous 

source terms consistent with MELCOR, which does not estimate aqueous releases. The 

Fukushima accidents demonstrated that large volumes of contaminated water can be generated 

which can disperse through surface water, sediments, soils, and groundwater; however, this is a 

gap in existing PRA modeling technology. Past assessments have shown that aqueous releases 

pose less overall health and environmental risk than airborne releases [1].

MACCS estimates consequences in four steps: 

1. atmospheric transport and deposition of radioactive materials onto land and water bodies,

2. the estimated exposures and health effects for up to seven days following the beginning of

release (early phase),

3. the estimated exposures and health effects during an intermediate time period of up to one

year (intermediate phase), and

4. the estimated long-term (e.g., 50 years) exposures and health effects (late-phase model).
4-2

The assessment of offsite property damage in terms of contaminated land and economic costs 

uses all four parts of the modeling. An overview of the code is provided below. 

MACCS12 version 3.7.5 was used for the consequence analyses [2] [3]. The WinMACCS 

graphical user interface (version 3.7.5) was used to input data into MACCS [4]. Site file data 

including population, economic values, and land use data was prepared using the SecPop 

preprocessor code version 4.3 [5]. MELCOR source terms were converted to MACCS input 

format using the MelMACCS version 1.7.3 code [6]. These codes have been developed by the 

NRC and Sandia National Laboratories (Sandia) over multiple decades and have rigorous quality 

control and quality assurance processes in place. Code capabilities described in this section are 

specific to the exact version used in the analysis. 

MACCS is used by U.S. nuclear power plant license renewal applicants to support the plant 

specific evaluation of severe accident mitigation alternatives (SAMAs) that may be required as 

part of the applicant’s environmental report for license renewal. MACCS is also routinely used in 

severe accident mitigation design alternative (SAMDA) or severe accident consequence analyses 

for environmental impact statements (EISs) supporting design certification, early site permit, and 

combined construction and operating license reviews for new reactors. The NRC’s regulatory 

analysis guidelines in NUREG/BR-0058, “Regulatory Analysis Guidelines of the U.S. Nuclear 

Regulatory Commission,” [7] and NUREG/BR-0184, “Regulatory Analysis Technical Evaluation 

Handbook,” [8] recommend the use of MACCS to estimate the averted “offsite property damage” 

cost (benefit) and the averted offsite dose cost elements. The information from MACCS code runs 

supports a cost-benefit assessment for various potential plant improvements as part of SAMAs or 

SAMDAs. 

MACCS has also been used in a variety of NRC research studies. MACCS was used in the Stateof-the-Art Reactor Consequence Analyses (SOARCA) project, which aimed to calculate the 

accident progression and consequences in a very detailed manner for the most important severe 

accident scenarios at the Peach Bottom Atomic Power Station (Peach Bottom) and the Surry 

Power Station (Surry). These analyses were documented in NUREG-1935 [9], NUREG/CR-7110 

[10] [11], and NUREG/BR-0359 [12]. The MACCS best practices as applied in the SOARCA

project were documented in NUREG/CR-7009 [13]. Following the SOARCA project was an

uncertainty analysis of one of the SOARCA scenarios, the Peach Bottom unmitigated long-term

station blackout (LTSBO), documented in NUREG/CR-7155 [14]. This study propagated

uncertainty for a variety of key uncertain MELCOR and MACCS parameters to develop insights

into the overall sensitivity of SOARCA results and conclusions and to identify the most influential

input parameters for consequences. The results of the Peach Bottom unmitigated LTSBO

uncertainty analysis corroborated the conclusions of the SOARCA project.

MACCS was also used in a consequence study of a beyond design basis earthquake affecting the 

spent fuel pool for a U.S. Mark I BWR and this is documented in NUREG-2161 [15]. MACCS was 

used in the SECY-12-0157 technical basis related to containment venting systems for BWRs with 

Mark I and Mark II containments [16]. In addition, MACCS is currently being used for the offsite 

consequence analyses supporting the NRC’s Full-Scope Level 3 Probabilistic Risk Assessment 

(PRA) for the Vogtle Electric Generating Plant [17].

12 Recent versions of MACCS have been known as “MACCS2” however NRC has decided to remove the “2” so that 

the code has just one version number (version 3.7.5 in this case). MACCS version 3.7.5 was developed in 2013 

and was the most current version available at the time.
4-3

4.1.1 Atmospheric Transport and Dispersion (ATD)

MACCS models dispersion of radioactive materials released into the atmosphere using the 

straight-line Gaussian plume segment model with provisions for meander and surface roughness 

effects. The ATD model treats the following: plume rise resulting from the sensible heat content 

(i.e., buoyancy), initial plume size caused by building wake effects, release of up to 200 plume 

segments, dispersion under statistically representative meteorological conditions, deposition 

under dry and wet (precipitation) conditions, and decay and ingrowths of up to 150 radionuclides 

and a maximum of six generations. The model does not treat in detail irregular terrain, spatial 

variations in the wind field, and temporal variations in wind direction.

The user has the option to use a single weather sequence or multiple weather sequences.

Sampling among multiple weather sequences is used in PRA studies to evaluate the effect of 

weather conditions at the time of the hypothetical accident.

The results generated by the ATD model include contaminant concentrations in air, on land, and 

as a function of time and distance from the release source; these results are subsequently used in 

early, intermediate, and long-term phase exposure modeling. 

4.1.2 Phases of the MACCS Conceptual Model

The U.S. Environmental Protection Agency (EPA), in its “Manual of Protective Action Guides and 

Planning Guidance for Radiological Incidents,” has characterized the response to a nuclear 

accident in three distinct phases of activity. MACCS has the capability to characterize the different 

exposure pathways, protective actions, and costs associated with the accident for each of the 

three phases. These are summarized in Table 4-1. First, the early (emergency) phase is used for 

the period of up to 7 days following the start of the initiating event that causes the accident. The 

intermediate phase starts at the end of the early phase and in MACCS can last up to one year.

The long-term phase starts at the conclusion of the intermediate phase and can last 50+ years.
4-4

Table 4-1 Phases of the MACCS conceptual model

Early Phase

(Emergency) Intermediate Phase Long-Term Phase

Primary Offsite

Accident 

Response 

Objective(s)

• Protect public

from plume

exposures.

• Protect public from

exposures to deposited

materials.

• Plan for long-term

cleanup and recovery

activities.

• Protect public from

exposures to deposited

materials.

• Conduct long-term

cleanup and recovery

activities.

Typical 

Duration and 

Time Frame 

Days to weeks, starting 

at the time of the 

accident’s initiating 

event. 7 days was used 

in all CPRR 

calculations.

Weeks to years, starting at 

the end of the early phase. 3 

months was used in CPRR 

base calculations.

Months to decades, starting at 

the end of the intermediate 

phase. 50 years was used in all 

CPRR calculations.

Exposure 

Pathways

• Cloudshine

• Groundshine

• Inhalation

• Skin deposition

• Groundshine

• Inhalation of

resuspended materials.

• Groundshine

• Inhalation of resuspended

materials.

• Food and water ingestion.

Protective 

Actions

• Sheltering

• Evacuation

• Relocation

• KI ingestion

• Relocation

• Interdiction

• Decontamination

• Condemnation

4.1.2.1 Early (Emergency) Phase Protective Actions and Exposure Pathways

The early phase model in MACCS assesses the time period immediately following a radioactive 

release. This period is commonly referred to as the emergency phase and it can extend up to 

seven days after the arrival of the first plume at any downwind spatial interval. Early exposures in 

this phase account for emergency planning (i.e., sheltering, evacuation, and relocation of the 

population). MACCS models sheltering and evacuation actions for user-specified population 

cohorts. Different shielding factors for the different exposure pathways (cloudshine, groundshine, 

inhalation, and deposition on the skin) are associated with three types of activities: normal activity, 

sheltering, and evacuation.

For population cohorts that are not explicitly modeled to evacuate in MACCS, dose-dependent 

relocation actions may take place during the emergency phase. If individuals at any location are 

projected to exceed either of two user-specified dose thresholds (a larger, “hotspot” threshold, and 

a smaller, “normal threshold”) over the duration of the emergency phase, they are relocated at a 

user-specified time after plume arrival and are modeled to receive no further early phase 

exposures.

MACCS also models the beneficial effect of populations consuming potassium iodide (KI) to 

reduce radioiodine inhalation doses to the thyroid. KI can saturate the thyroid with stable iodine 

and thereby reduce the amount of radioiodine that can be absorbed. KI is distributed near some 

nuclear power plants. MACCS allows the user to specify which population cohorts would take KI, 

the expected fraction of the population within each cohort that would take it, and the efficacy of the 

KI in reducing thyroid doses.
4-5

4.1.2.2 Intermediate Phase Protective Actions and Exposure Pathways

MACCS can model an intermediate phase with a duration of up to one year following the end of 

the early phase. The only protective action modeled in this phase is relocation. If the projected 

dose to a population exceeds a user-specified threshold over a user-specified time duration, the 

population is assumed to be relocated to an uncontaminated area for the entire duration of this 

phase. A corresponding per-capita per diem economic cost is defined by the user. If the projected 

dose does not reach the user specified threshold, exposure pathways for groundshine and 

inhalation of resuspended material are modeled. The food and water ingestion pathway is not 

modeled in the intermediate phase because of the assumption that uncontaminated food and 

water would be brought in from outside the affected region during this interim period.

4.1.2.3 Long-Term Phase Protective Actions and Exposure Pathways

In the long-term phase (typically 30-50 years following the end of the intermediate phase), 

protective actions are defined to minimize the dose to an individual by external (e.g., groundshine) 

and internal (e.g., food consumption and resuspension inhalation) pathways. Decisions on 

protective actions are based on two sets of independent actions — i.e., decisions relating to 

whether land, at a specific location and time, is suitable for human habitation (habitability) or 

agriculture production (farmability). Habitability and farmability are defined by a set of userspecified maximum doses and a user-specified exposure periods to receive those doses.

Habitability and farmability decision-making can result in four possible outcomes: 

1. land is immediately habitable or farmable,

2. land is habitable or farmable after decontamination,

3. land is habitable or farmable after decontamination and interdiction13, or

4. land is deemed not habitable after decontamination plus 30 years of interdiction or land

is deemed not farmable after decontamination plus 8 years of interdiction (i.e., it is

condemned).

Land is also condemned if the cost of decontamination exceeds the value of the land. The dose 

criterion for the MACCS modeling of individuals returning back to the affected (i.e., contaminated) 

area is a user input and is typically based on the U.S. Environmental Protection Agency (EPA) 

Protective Action Guides (PAGs) [18] or state-specific guidelines. The decision on whether land is 

suitable for farming is first based on prior evaluation of its suitability for human habitation. 

Decisions on decontamination are made using a decision tree. The first decision is whether land is 

habitable. If it is, then no further actions are needed. The population returns to their homes and 

receive a small dose from any deposited radionuclides for the entire long-term phase. If land is not 

habitable, the first option considered is to decontaminate at the lowest level of dose reduction, 

which is also the cheapest to implement. If this level is sufficient to restore the land to habitability, 

then it is performed. Following the decontamination, the population return to their homes and 

receive a small dose based on the residual contamination for the duration of the long-term phase.

If the first level of decontamination is insufficient to restore habitability, then successively higher 

levels are considered. MACCS considers up to three decontamination levels. If the highest level of 

13 In this context, interdiction generally refers to the period of time in which residents are not permitted to return to 

live on their property because the radiation doses they would receive (from external sources and inhalation) 

exceed the habitability criterion. Interdiction allows for radioactive decay, decontamination, and weathering to 

potentially bring these doses to a point where they would no longer exceed the habitability criterion.
4-6

decontamination is insufficient, then interdiction for up to 30 years is considered following the 

highest level of decontamination. During the interdiction period, radioactive decay and weathering 

work to reduce the dose rates that would be received by the returning population. If the highest 

level of decontamination followed by interdiction is sufficient to restore habitability, then it is 

employed and the population is allowed to return. Doses are accrued for the duration of the longterm phase. If habitability cannot be restored by any of these actions, then the land is condemned.

The land is also condemned if the cost of the required action to restore habitability is greater than 

the value of property.

The decision tree for farmability is first based on prior evaluation of its suitability for human 

habitation—land cannot be used for agriculture unless it is habitable. Furthermore, farmland must 

be able to grow crops or produce dairy products that meet the user-specified farmability criterion, 

which is an ingestion dose equivalent threshold, and in this analysis is set to equal the habitability 

criterion. If farmland is habitable and farmable, a food chain model is used to determine doses 

that would result from consuming the food grown or produced on this land. The COMIDA2 food 

chain model is the latest model developed for use in MACCS. This model contains data on 

expected radionuclide uptake in nine foodstuff types for different seasons of the year and for 

different contamination levels and food category consumption rates for an average adult.

MACCS values of total long-term population dose and health effects account for exposures 

received by workers performing decontamination. While engaged in cleanup efforts, workers are 

assumed to wear respiratory protection devices; therefore, they only accumulate doses from 

groundshine.

4.1.3 Offsite Consequence Measures

The results of the consequence analyses are presented in terms of health risks to the public, 

population dose, land contamination, population subject to long-term protective actions, and 

economic costs. All consequence results are presented in this section as conditional 

consequences (i.e., assuming that the accident occurs), and show the risks to individuals as a 

result of the accident (i.e., LCF risk per event or early fatality risk per event). Therefore in this 

section, there is no consideration of the different probabilities/frequencies of the different accident 

progression scenarios. The risks, population dose, and economic costs are mean values (i.e., 

expectation values) over sampled weather conditions representing a year of meteorological data 

and over the entire residential population within a circular or annular region. 

4.1.3.1 Health Effects

Populations located on the MACCS computational grid receive doses from the passing plume 

(cloudshine), by exposure to materials deposited on the ground (groundshine), by inhalation of 

airborne radioactive materials (from the plume or from mechanical or wind-driven resuspension of 

materials deposited on the ground), and by ingestion of contaminated food and water.

MACCS uses a dose conversion factor file based on EPA’s Federal Guidance Report No. 13, 

“Cancer Risk Coefficients for Environmental Exposure to Radionuclides,” [19] which converts the 

integrated air concentration and ground deposition of 825 radionuclides to a whole body effective 

dose and individual organ doses for 26 tissues and organs and for four exposure pathways. The 

whole body effective dose and individual organ doses are then used to calculate health effects. In 

general, the radiological dose to a receptor in a given spatial element is the product of the 

radionuclide concentration or quantity, the exposure duration, the shielding factor, the dose 

conversion factor, and the usage factor (e.g., breathing rate). The total dose to an organ or the 
4-7

whole body used for modeling of health effects or protective action decisionmaking is then 

summed across the relevant exposure pathways and radionuclides.

MACCS considers deterministic and stochastic health effects and estimates the likelihood that an 

exposed individual may experience a specific health effect (e.g., lung impairment, breast cancer).

Deterministic health effects (early injuries and early fatalities) are calculated using nonlinear dose 

response models. These models consider the dose equivalent delivered to the target organ, a 

dose equivalent threshold below which the effect is not expected to occur and thus the risk is 

zero, a dose equivalent that would induce the effect in half the exposed population, and a shape 

factor that affects the range over which the likelihood of the health effect goes from zero to one for 

the overall population. For stochastic health effects (latent cancer fatalities), the NRC uses the 

linear no-threshold (LNT) model for analyses serving regulatory purposes, and this model is 

adopted in MACCS here. The LNT dose-response relationship suggests that any increase in 

dose, no matter how small, results in an incremental risk. The parameters supporting these 

models are discussed in Section 4.2.10. 

4.1.3.2 Land Contamination and Populations Subject to Long-Term Protective Actions

Land contamination can be computed in two different ways in MACCS. The first is to compute the 

area of land exceeding a user-specified areal concentration of a user-specified radionuclide. For 

example, the area of land exceeding a cesium-137 ground concentration of 15 μCi/m2 (15 Ci/km2

) 

is a useful metric for quantifying the long-term land contamination caused by the accident 

because cesium-137 is the most important long-term contaminant that limits habitability. This 

metric was used to quantify the land contamination caused by the Chernobyl accident. The 

second method is to compute the area of land for which human habitation and farm production 

are temporarily restricted or permanently restricted (condemned). This metric can be reported 

over any user-specified circular area or ring. This metric is more useful for characterizing the 

extent of land over which long-term phase protective actions are necessary.

An additional consequence metric is the population subject to long-term protective actions, 

including land and property interdiction and condemnation. In addition to the area of land that 

must be restricted from human habitation and farm production, this metric is useful for assessing 

the societal consequence of an accident.

4.1.3.3 Economic Consequences

Every protective action modeled in MACCS to reduce radiation exposures to the public (except for 

sheltering and potassium iodide (KI) ingestion14 during the early phase) has an associated cost.

The offsite economic consequence model in MACCS sums the costs for protective actions over 

the region of interest and includes the six categories as follows. (Costs for medical care, lifeshortening, and litigation are not calculated by MACCS.)

1. Evacuation and relocation costs. These are per-diem costs associated with the population

that is temporarily relocated. This includes the population that is explicitly modeled to

evacuate during the early phase, the population that would relocate during the early phase

due to the hotspot or normal relocation criteria, and the population that would need to be

relocated for the entire duration of the intermediate phase. These costs are calculated by

14 There is a cost to maintain a KI acquisition, distribution, and awareness program; however, this cost is incurred 

regardless of the occurrence of the accident and is therefore not included in the offsite cost calculation. 
4-8

adding up the number of displaced people times the number of days they are displaced 

from their homes. 

2. Moving expenses for people displaced. This is a one-time moving expense for the

population displaced from their homes because of long-term phase relocation for

decontamination, interdiction, or condemnation. The modeling can include loss of wages.

3. Decontamination costs. These are the costs associated with decontaminating farm and

nonfarm property and include labor, materials, and equipment for performing the

decontamination as well as the cost to dispose of the contaminants. They depend on the

population and size of the area that needs to be decontaminated as well as the level of

decontamination that needs to be performed. The model estimates the costs only if

decontamination is cost effective.

4. Costs due to loss of use of property. These costs are associated with the lost return on

investment and on depreciation caused by lack of routine maintenance for farm and

nonfarm properties during the period of interdiction, the time when the property cannot be

used.

5. Disposal of contaminated food grown or produced locally. For farmland that exceeds the

farmability criterion and is modeled to be temporarily unable to produce crops, meat, and

dairy products that are suitable for human consumption, this food must be disposed for

the current growing season. This cost sums the expected food sales per area times the

affected area. The site data file includes a parameter to estimate the fraction of annual

farm sales for dairy products and thus the disposal cost for dairy products is reported

separately from the disposal cost for all other agricultural products.

6. Cost of condemned lands. For farmland and nonfarmland that cannot be restored to

usefulness or is not cost-effective to do so, the land is condemned. These are costs of

condemning property, i.e., the value of the property condemned.

All of the costs for the six cost categories are summed over the entire region of interest affected 

by the atmospheric release to get the total offsite economic costs. Many of the values affecting the 

economic cost model are user inputs and thus can account for a variety of costs and can be 

adjusted for inflation, new technology, or changes in policy. Other data for the cost model come 

from the site file, which is discussed in more detail in section 3.3. The site file uses external data 

from the U.S. Census, the U.S. Bureau of Economic Analysis, and the U.S. Agricultural Census. 

These values are scaled to the year of interest.

4.2 Modeling Approach

The CPRR technical evaluation is designed to be applicable to all Mark I and II BWR containment 

sites in the United States. There is considerable variation in many different characteristics at the 

15 Mark I and 5 Mark II sites. Sites differ with respect to population (number and distribution); 

economic values; land use (land vs. water, farmland vs. developed land, etc.); weather (wind, 

precipitation, etc.); emergency response characteristics (time to evacuate 10-mile emergency 

planning zone (EPZ), use of potassium iodide (KI), etc.); long-term protective actions (habitability 

criterion); and many others. To capture the variation among these characteristics in the most 

resource-efficient manner, one site-specific Mark I reference MACCS model was developed and 

one site-specific Mark II reference model was developed. These reference models were then 

adapted in a series of over 100 sensitivity calculations to assess the potential impact of sitespecific parameters on offsite consequence results.
4-9

The Peach Bottom Atomic Power Station and the Limerick Generating Station were selected as the 

site-specific reference models to enable greater modeling fidelity for the high population sites (Peach 

Bottom has the second highest population within a 50 mile radius among the 15 Mark I sites and 

Limerick has the highest population within a 50 mile radius among the five Mark II sites). The modeling 

approach used the most current sources of information and consequence modeling best practices to 

make the Peach Bottom and Limerick MACCS models as realistic as practically possible.

This section discusses the modeling approach for each of the components of the analysis.

Greater detail is provided for those modeling parameter selections that differ from those used in 

SOARCA, SECY-12-0157, and other recent analyses. 

4.2.1 Radionuclide Release

The preceding MELCOR accident progression modeling section describes the assumptions, 

calculations, and results of the different accident progression scenarios. For the reference BWR 

Mark I site MELCOR model (Peach Bottom), 41 unique MELCOR simulations were run. For the 

reference BWR Mark II site MELCOR model (LaSalle), 12 unique MELCOR simulations were run.

MELCOR provides the following data for each source term:

• Time-dependent release fraction of 9 chemical groups for each MELCOR release

pathway: Noble Gases (Xe), Alkali Metals (Cs), Alkali Earths (Ba), Halogens (I),

Chalcogens (Te), Platinoids (Ru), Early Transition Elements (Mo), Tetravalents (Ce),

and Trivalents (La)

• Time-independent distribution by particle size diameter for 10 aerosol size bins

characterized by geometric mean diameters from 0.15 μm to 41.2 μm for each chemical

group

• Height of each MELCOR release pathway

• Time-dependent plume rise data including rate of release of sensible heat (W), mass

flow (kg/s), and gas density (kg/m3

)

In addition, the following data is needed to characterize each source term. For each item, the 

modeling approach is described. 

• Radionuclides to include in consequence analysis and their assignment to chemical

groups

o Both the BWR Mark I reference site MACCS model and the BWR Mark II reference

site MACCS model use the identical modeling approach as the SOARCA project

for both the Peach Bottom and Surry analyses. The full details are available in

Table A-1 of NUREG/CR-7110, Vol. 1, Rev. 1 [10].

• Pseudostable radionuclides to include in consequence analysis

o Both the BWR Mark I reference site MACCS model and the BWR Mark II reference

site MACCS model use the identical modeling approach as the SOARCA project

for both the Peach Bottom and Surry analyses. The full details are available in

Table B-1 of NUREG/CR-7110, Vol. 1, Rev. 1 [10].

• Radionuclide inventory (activity) at the time of reactor shutdown
4-10

o The BWR Mark I reference site MACCS model uses the identical modeling

approach as the SOARCA project for the Peach Bottom analyses. The inventory

was computed using the SCALE code to model the specific fuel management

strategy used at Peach Bottom. 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 was

calculated by integrating the isotopic inventory over the whole core.

o The BWR Mark II reference site MACCS model uses the same radionuclide

inventory as the Mark I site reference MACCS model because the reference plants

are very similar and have essentially identical thermal power output; Peach Bottom

is licensed at 3,514 MWt and Limerick is licensed at 3,515 MWt [20]. In addition, the

same mid-cycle accident timing assumption is made for the Mark II reference site.

The full details are available in Table A-2 through Table A-10 of NUREG/CR-7110,

Vol. 1, Rev. 1 [10].

• Plume segmentation characteristics including duration and thresholds for inclusion

o Consistent with the SOARCA project, the radionuclide release is divided into hourly

plume segments to be consistent with the resolution of the accompanying

meteorological data. Also consistent with the SOARCA project, 0.001 is used as

the threshold for inclusion of each MELCOR release pathway and each plume

segment. Thus, a MELCOR release pathway is used only if more than 0.001

(0.1%) of the total release of any chemical group occurs through it. A plume

segment is evaluated only if any of the chemical groups in that segment contribute

0.001 (0.1%) to the total release of that chemical group.

• Identification of the most risk-significant plume segment

o For each source term, the most risk-significant plume segment needs to be

identified to align the release with the weather data for each weather bin. The risksignificant plume segment is considered to be the one that causes the highest risk

for early fatalities. The modeling approach was to select the plume segment

among the first few that has the highest iodine chemical group release fraction. All

representative cases for the Mark I and Mark II source term bins used either the

first or second plume segment as most risk-significant.

• Building height and initial vertical and horizontal plume size to characterize the initial

dispersion of the plume and address building wake effects

o Both the BWR Mark I and II reference plants are approximately 50 m in height and

width. Consistent with the SOARCA project, the initial horizontal dispersion uses

the equation σy0 = 0.23 * Wb and the initial vertical dispersion uses the equation σz0

= 0.47 * Hb [4]. Therefore both BWR Mark I and II MACCS models use an initial

horizontal dispersion of 11.6 m and vertical dispersion of 23.3 m.
4-11

• Ground height in the MELCOR reference frame used to adjust MELCOR release heights

to release heights relative to grade

o The BWR Mark I analyses use the Peach Bottom MELCOR model, which has

grade level at an elevation of -4.04 m. The BWR Mark II analyses use the LaSalle

MELCOR model, which has grade level at an elevation of -14.5 m. These values

were obtained by reviewing plant schematic diagrams and MELCOR model

nodalization diagrams.

4.2.1.1 Modeling the Decontamination Factor Provided by an External Filter

The CPRR rulemaking technical analysis for BWR Mark I and Mark II plants includes options for 

small and large external engineered filters attached to wetwell and drywell vents. MELCOR 

reference plant models allow for very detailed modeling of accident progression and they model 

filtration by the wetwell/suppression pool; however, they do not currently provide mechanistic 

modeling of the presence of external filters. Therefore the filtration provided by an external filter is 

modeled by applying a decontamination factor (DF) to the MELCOR source term. In many 

accident progression cases, a direct pathway to the environment other than a wetwell or drywell 

vent path is created. For these cases, the DF is applied only to the wetwell and/or drywell vent 

release paths. The presence of a release pathway other than the wetwell or drywell vent and the 

fraction that goes through this pathway, are therefore major drivers of the potential effectiveness 

of an external filter. This topic is discussed further in Section 3.1.3.

In the absence of a research program demonstrating the effectiveness of different filter designs for 

removing aerosols across the particle size spectrum, the team decided to consider a range of 

decontamination factors to apply to all source terms. The range includes a DF of 10, 100, and 

1000 (which is consistent with the range reported in the OECD/NEA/CSNI Status Report on 

Filtered Containment Venting [21]) and each is applied uniformly among all chemical groups

despite their different particle size distributions. An attempt to identify chemical group-specific DFs 

based on their particle size distribution was considered for this analysis but is being deferred to 

future analyses based on schedule and resource constraints. 

4.2.1.2 Source Term Binning Process

The CPRR rulemaking accident progression analysis produced 41 Mark I source terms and 12 Mark 

II source terms. Considering the range of external filter DFs for each (unfiltered, DF = 10, DF = 100, 

and DF = 1000), this yields 164 Mark I source terms and 48 Mark II source terms. Many of the 

source terms have very similar release fraction and release timing characteristics, so rather than 

running all 212 source terms in MACCS, a binning strategy was developed for each containment 

type. The binning process was based on cumulative cesium and iodine release fractions, because 

the cesium group is most important for long-term offsite consequences and the iodine group is most 

important for early offsite consequences. The time of release of the risk-significant plume segment 

was also considered. However, in all cases it was after the time at which the 10-mile emergency 

planning zone (EPZ) is expected to be evacuated, and therefore was not considered as important.

For example, for Peach Bottom, the EPZ evacuation is expected to be completed about four hours 

after the time of notification [22]. (This topic is discussed in more detail in Section 3.6.2). Assuming 

public notification via general emergency (GE) siren at 1.5 hours after the accident, the EPZ is 

cleared in 5.5 hours. In comparison, the earliest risk-significant plume segment starts at 7.3 hours 

from the start of the accident and is more generally in the range of 10 to 24 hours from the start of 

the accident.
4-12

The source term bins were developed based on logarithmic spacing with finer resolution at higher 

magnitudes. The bin definitions for the Mark I source terms are provided in Table 4-2 and they span 

a very large (> 4 orders of magnitude) range of cesium release from 0.0006% to about 16%. The 

representative source term from each bin was selected by choosing the source term that was most 

similar to the average cesium and iodine release fractions for all the source terms in that bin. Table 

4-2 shows the summary source term information for each of the 18 Mark I source term bins.Table

4-3 identifies which source term bin corresponds to each of the 164 unique Mark I source terms.

The binning strategy for the Mark II source terms was very similar to that used for the Mark I source 

terms but since there were far fewer source term cases, the bin spacing for iodine release fraction 

was chosen to be somewhat discontinuous for the highest consequence bins. The cesium and 

iodine release fractions were used to group the source terms into bins. Consistent with the Mark I 

source terms, the start of release to the environment was sufficiently later than the expected EPZ 

evacuation completion time so the Mark II source terms were not binned considering release start 

timing. Table 4-4 shows the summary source term information for each of the 9 Mark II source term 

bins. Table 4-5 identifies which source term bin corresponds to each of the 48 unique Mark II source 

terms. Additional details for each individual source term case are provided in Appendix A. 

Table 4-2 Binning strategy for Mark I source terms

Bin Bin Cs Range 

(%)

Bin I Range 

(%)

Representative

Case

Rep Case

Cs (%)

Rep Case

I (%)

Start of Release 

to Environment 

(hours)

1 0.0002 - 0.001 0.001 - 0.01 28DF1000 0.0006% 0.006% 15.9

2 0.001 - 0.003 0.01 - 0.03 48DF100 0.002% 0.02% 11.4

3 0.003 - 0.01 0.03 - 0.1 10DF100 0.01% 0.08% 16.3

4 0.01 - 0.03 0.1 - 0.3 7DF1000 0.02% 0.26% 14.9

5 0.03 - 0.1 0.3 - 1.0 11DF10 0.06% 0.78% 14.4

6 0.1 - 0.3 1.0 - 3.0 48 0.23% 1.69% 11.4

7 0.3 - 1.0 3.0 - 10.0 15 0.60% 5.85% 15.9

8 0.3 - 1.0 10.0 - 20.0 46 0.98% 11.01% 14.8

9 1.0 - 2.0 2.0 - 4.0 5DF10 1.05% 2.89% 24.2

10 1.0 - 2.0 4.0 - 10.0 5 1.39% 6.46% 24.2

11 1.0 - 2.0 10.0 - 20.0 8 1.49% 19.25% 14.9

12 1.0 - 2.0 20.0 - 40.0 1 1.93% 22.68% 14.9

13 2.0 - 4.0 3.0 - 10.0 41DF1000 3.40% 7.65% 9.8

14 2.0 - 4.0 10.0 - 20.0 22dw 2.82% 18.64% 15.9

15 2.0 - 4.0 20.0 - 40.0 53 2.79% 29.05% 18.4

16 4.0 - 10.0 10.0 - 20.0 41 4.54% 14.10% 9.8

17 4.0 - 10.0 20.0 - 40.0 3DF10 8.85% 24.65% 9.8

18 10.0 - 20.0 20.0 - 40.0 52 15.90% 34.32% 18.4
4-13

Table 4-3 Identification of source term bin for each Mark I source term case

Case Bin Case Bin Case Bin Case Bin Case Bin Case 

1 12 7 11 13 7 22dw 14 41 16 48 6 

1DF10 7 7DF10 6 13DF10 5 22dwDF10 6 41DF10 13 48DF10 4 

1DF100 7 7DF100 5 13DF100 3 22dwDF100 4 41DF100 13 48DF100 2 

1DF1000 7 7DF1000 4 13DF1000 1 22dwDF1000 2 41DF1000 13 48DF1000 1 

1S1 7 7dw 11 14 7 24 7 42 16 49 7 

1S1DF10 7 7dwDF10 6 14DF10 5 24DF10 5 42DF10 13 49DF10 5 

1S1DF100 6 7dwDF100 4 14DF100 4 24DF100 4 42DF100 13 49DF100 3 

1S1DF1000 6 7dwDF1000 2 14DF1000 4 24DF1000 2 42DF1000 13 49DF1000 2 

2 15 8 11 15 7 24dw 13 43 16 50 6 

2DF10 10 8DF10 6 15DF10 5 24dwDF10 6 43DF10 13 50DF10 4 

2DF100 10 8DF100 4 15DF100 3 24dwDF100 4 43DF100 13 50DF100 2 

2DF1000 10 8DF1000 2 15DF1000 1 24dwDF1000 2 43DF1000 13 50DF1000 1 

3 17 9 7 16 7 27 7 44 16 51 8 

3DF10 17 9DF10 5 16DF10 5 27DF10 5 44DF10 13 51DF10 5 

3DF100 17 9DF100 3 16DF100 3 27DF100 4 44DF100 13 51DF100 3 

3DF1000 17 9DF1000 1 16DF1000 2 27DF1000 4 44DF1000 13 51DF1000 1 

4 10 10 7 18 7 28 7 45 7 52 18 

4DF10 10 10DF10 5 18DF10 5 28DF10 5 45DF10 5 52DF10 13 

4DF100 10 10DF100 3 18DF100 3 28DF100 3 45DF100 3 52DF100 13 

4DF1000 10 10DF1000 1 18DF1000 3 28DF1000 1 45DF1000 1 52DF1000 13 

5 10 11 7 21 11 30 7 46 8 53 15 

5DF10 9 11DF10 5 21DF10 6 30DF10 5 46DF10 5 53DF10 6 

5DF100 9 11DF100 3 21DF100 5 30DF100 3 46DF100 3 53DF100 4 

5DF1000 9 11DF1000 2 21DF1000 5 30DF1000 2 46DF1000 2 53DF1000 2 

6 12 12 11 22 12 32 7 47 6 

6DF10 7 12DF10 6 22DF10 6 32DF10 5 47DF10 4 

6DF100 7 12DF100 4 22DF100 4 32DF100 3 47DF100 2 

6DF1000 7 12DF1000 2 22DF1000 2 32DF1000 1 47DF1000 1 

Bin
4-14

Table 4-4 Binning strategy for Mark II source terms

Table 4-5 Identification of source term bin for each Mark II source term case

Case Bin Case Bin Case Bin Case Bin Case Bin Case Bin

1 8 5 6 11 4 42 6 45 8 51 8 

1DF10 5 5DF10 4 11DF10 3 42DF10 4 45DF10 5 51DF10 5 

1DF100 4 5DF100 3 11DF100 2 42DF100 3 45DF100 4 51DF100 4 

1DF1000 3 5DF1000 2 11DF1000 1 42DF1000 2 45DF1000 3 51DF1000 3 

3 7 10 5 24 6 44 6 49 6 52 9 

3DF10 5 10DF10 4 24DF10 4 44DF10 4 49DF10 4 52DF10 5 

3DF100 4 10DF100 3 24DF100 3 44DF100 3 49DF100 3 52DF100 4 

3DF1000 3 10DF1000 2 24DF1000 2 44DF1000 2 49DF1000 2 52DF1000 3 

4.2.1.3 External Filter Effectiveness

For accident progression cases in which all releases to the environment are through a vent path, 

an external filter can reduce the source term and offsite consequences. However, for accident 

progression cases that lead to containment failure, for example, via drywell liner 

melt-through (DW LMT) or main steam line creep rupture (MSLCR), an external filter is less 

effective. Table 4-6 shows some of the variation in external filter effectiveness through examples 

of three MELCOR Mark I cases. 

MELCOR case 1 is shown as an example in which there is no post-core-damage external water 

addition (either because the plant lacks the capability or because external water additional is 

unsuccessful) and the accident results in an uncontrolled release via DW LMT. In MELCOR case 

1, much of the release (78.2% of the cesium) is through a vent pathway15 so the external filter can 

15 Chapter 3 provides a discussion of the reasons for why the filterable release percent varies among the different 

MELCOR cases resulting in DW LMT.

Bin Bin Cs Range (%) Bin I Range (%) Representative

Case

Rep Case

Cs (%)

Rep Case

I (%)

Start of Release 

to Environment 

(hours)

1 0.00001 - 0.0001 0.0001 - 0.001 11DF1000 0.00004% 0.0005% 20.3

2 0.0001 - 0.001 0.001 - 0.01 5DF1000 0.0006% 0.005% 32.2

3 0.001 - 0.01 0.01 - 0.1 42DF100 0.0043% 0.037% 14.3

4 0.01 - 0.1 0.1 - 1.0 11 0.042% 0.45% 20.3

5 0.1 - 0.4 1.0 - 3.0 51DF10 0.23% 2.01% 16.6

6 0.4 - 1.0 3.0 - 10.0 5 0.55% 4.94% 32.2

7 1.0 - 2.0 ~ 10.0 3 1.09% 10.26% 14.3

8 2.0 - 3.0 ~ 20.0 1 2.46% 19.81% 22.8

9 3.0 - 4.0 ~ 30.0 52 3.57% 28.67% 16.6
4-15

substantially reduce the environmental release. However, the incremental benefit of increasing the 

external filter DF becomes very small.

For a similar but less likely case in which the containment fails instead via MSLCR, most of the 

release is uncontrolled and goes through an unvented pathway so the external filter has a very 

small effect on the total source term released to the environment. Even though the release to the 

environment continues to decrease as the external filter DF increases, the source term remains 

within the existing source term bin (Mark I bin 17) so the offsite consequences remain essentially 

unchanged.

For MELCOR Mark I Case 10 in which external water addition is successful, all of the released 

cesium flows through a vented pathway and therefore the external filter can potentially reduce the 

environmental release. Note that the MACCS source term bin number decreases with each 

incremental DF applied.

Table 4-6 External filter effectiveness for three example BWR Mark I cases

4.2.2 Calculational Grid

MACCS uses a polar grid to model radionuclide releases to the environment, exposures to 

people, land contamination, and protective actions of people and land. MACCS allows the user to 

divide the grid into 16, 32, 48, or 64 angular sectors. Consistent with modeling best practices and 

recent consequence analyses studies, 64 angular sectors were used in this project because this 

provides the greatest resolution.

MACCS allows the user to divide the polar grid into a maximum of 35 radial rings. For this project, 

26 radial sectors were selected. The boundaries of each sector are described in Table 4-7. The 

CPRR 

Alternatives

MELCOR 

Mark I 

Case and 

External 

Filter DF

Percent of 

Source Term 

Released 

Through Vented 

Pathway

Total Source Term 

Released to 

Environment

MACCS 

Source 

Term 

Bin

Description of 

External Filter 

Effectiveness 

Cesium Iodine Cesium Iodine

No External 

Water 

Addition 

Resulting in 

DW LMT 

1 

78.2% 85.5%

1.93% 22.70% 12 External filter has 

a notable effect on 

reducing 

environmental 

release for DF=10 

but smaller 

incremental 

benefit for higher 

DF

1DF10 0.57% 5.24% 7 

1DF100 0.44% 3.49% 7 

1DF1000 0.42% 3.32% 7 

No External 

Water 

Addition 

resulting in 

MSLCR 

3 

11.5% 21.6%

9.88% 30.20% 17 External filter has 

an insignificant 

effect on reducing 

environmental 

release 

3DF10 8.85% 24.32% 17

3DF100 8.75% 23.74% 17

3DF1000 8.74% 23.68% 17

External Water 

Addition 

Successful 

10

100.0% 100.0% 

0.72% 8.04% 7 External filter 

reduces 

environmental 

release 

10DF10 0.07% 0.80% 5 

10DF100 0.007% 0.08% 3 

10DF1000 0.0007% 0.008% 1 
4-16

boundaries are closer together nearer to the site to enable higher resolution at areas of greater 

interest. Boundaries are selected to be consistent with certain areas of interest for the analysis.

For example, the radial boundary at 1.33 miles from the site is used to approximate the area 

within a mile of the site boundary, 10 miles is used to approximate the plume exposure EPZ, 15 

miles is used to approximate the outer distance for shadow evacuation (see Section 3.6.2 for 

discussion), and 50 miles is used to approximate the ingestion exposure EPZ.

The use of 64 angular sectors and 26 radial rings enables results to be reported over 1,664 grid 

elements. A discussion on which grid elements are used for the output metrics is provided later in 

Section 4.

Table 4-7 Radial boundaries used in the polar calculational grid

4.2.3 Site Data

The SecPop preprocessor code (version 4.3) was used to generate site data that is needed for 

the consequence calculations. SecPop accesses external population, land use, and economic 

databases to obtain the data and then uses various algorithms to map the data to each of the 

1,664 individual grid elements. The data types and sources are summarized in Table 4-8. More 

detail on the data sources and the algorithms used are provided in the SecPop reference manual, 

NUREG/CR-6525, Rev. 1 [5].

Radial 

Boundary 

Number 

Radial 

Boundary 

Distance (mi) 

Radial 

Boundary 

Distance (km)

Radial 

Boundary 

Number 

Radial 

Boundary 

Distance (mi) 

Radial 

Boundary 

Distance (km)

1 0.1 0.16 14 15 24.1 

2 0.33 0.53 15 20 32.2 

3 0.75 1.21 16 25 40.2 

4 1 1.61 17 30 48.3 

5 1.33 2.1 18 40 64.4 

6 2 3.2 19 50 80.5 

7 2.5 4.0 20 70 113 

8 3 4.8 21 100 161 

9 3.5 5.6 22 150 241 

10 5 8.0 23 200 322 

11 7.5 12.1 24 350 563 

12 10 16.1 25 500 805 

13 12.5 20.1 26 1000 1610 
4-17

Table 4-8 Site data types and sources

Population data was scaled forward from 2010 to the year of interest based on state level 

population growth data available from the U.S. Census Bureau. The year of interest was selected 

as 2013 because it was the most recent year for which population projection and consumer price 

index data were available. For sites such as Peach Bottom and Limerick, where the 50-mile radial 

area includes multiple states, an approximate land fraction was used to weight the state-specific 

population growth rates. This is presented in Table 4-9 and shows a weighted average value of 

1.016 for Peach Bottom and 1.009 for Limerick. SecPop reads population data at the census 

block level and uses various algorithms to map the data to the user-specified MACCS polar grid.

Table 4-9 Population growth multipliers for Peach Bottom and Limerick

Economic values are based on 2007 data and are scaled to 2013 using the consumer price index 

for all urban consumers (CPI-U) [23]. Based on a national CPI-U of 207.342 in 2007 and a 

national CPI-U of 232.957 in 2013, a multiplier of 1.124 is used for all economic values.

State Census 

2010

Census 2013 

Est. 

2010 to 

2013

Multiplier 

Peach Bottom 

Approximate 

50-mile Area

Fraction

Limerick 

Approximate 

50-mile Area

Fraction

PA 12,702,379 12,773,801 1.006 0.5 0.7 

MD 5,773,552 5,928,814 1.027 0.35 0.05 

DE 897,934 925,749 1.031 0.1 0.05 

NJ 8,791,894 8,899,339 1.012 0.05 0.2 

Population Multiplier for 2010 to 2013: 1.016 1.009 

Data Type Source 

Population  U.S. Census Bureau, 2010 Census

Land fraction  U.S. Census Bureau, 2010 Census

Fraction of land 

used for farming  U.S. Department of Agriculture, 2007 Census of Agriculture

Farm value per 

hectare  U.S. Department of Agriculture, 2007 Census of Agriculture

Annual farm sales 

per hectare  U.S. Department of Agriculture, 2007 Census of Agriculture

Fraction of annual 

farm sales from 

dairy products 

 U.S. Department of Agriculture, 2007 Census of Agriculture

Non-farm wealth 

per capita 

 U.S. Census Bureau, 2010 Census

 U.S. Census Bureau, 2007 American Housing Survey

 U.S. Department of Agriculture, Economic Research Service

 U.S. Department of Agriculture, Natural Resources Conservation

Service, 2007 National Resources Inventory Report

 U.S. Department of Commerce, Bureau of Economic Analysis
4-18

4.2.4 Meteorological Data

The atmospheric transport and dispersion model in MACCS relies on the following types of 

meteorological data for the following purposes. Data for wind direction, wind speed, precipitation, 

and stability class can be provided hourly or every 15 or 30 minutes and should span an entire 

year to capture daily and seasonal variations. 

• Hourly wind direction data is used to identify the direction each plume segment will travel

• Hourly wind speed data is used to characterize the speed of plume travel away from the

site and the plume meander factor. Plume speed is adjusted each hour for each plume

segment.

• Hourly precipitation data is used to determine the timing and magnitude of wet

deposition and the timing of reduced evacuation travel speeds

• Hourly Pasquill-Gifford stability class data is used to characterize the dispersion of the

plume in the vertical and cross-wind directions and the plume meander factor. Stability

class is adjusted each hour for each plume segment.

• Diurnal (morning and afternoon) seasonal mixing layer height data is used to determine

the upper boundary of the region in which each plume may expand.

4.2.4.1 Hourly Meteorological Data

The hourly data was developed through an analysis of the raw weather data from site 

meteorological towers provided by Exelon, the licensee of Peach Bottom and Limerick. Exelon 

provided raw weather data for Peach Bottom for 2005 and 2006 and for Limerick for 2012 and 

2013. MACCS requires meteorological data to be provided for each time point of the entire year 

(8,760 hourly data points for a non-leap year). Therefore missing data was reviewed and was 

filled in by NRC meteorologists and in accordance with EPA’s “Meteorological Monitoring 

Guidance for Regulatory Applications” [24]. For Peach Bottom and Limerick, the 2006 and 2013 

years, respectively, were chosen because they had the higher data recovery rates of the two 

years available. A summary of meteorological statistics for the data sets is provided in Table 4-10. 

The wind rose for each data set is provided in Figure 4-1 on a 16 sector grid. These show that for 

both sets, the most common wind direction is toward the southeast. 

Table 4-10 Summary of meteorological data sets used for Peach Bottom and Limerick

Peach Bottom Year 

2006 

Limerick 

Year 2013 

Average Wind Speed (m/s) 2.12 2.36 

Precipitation 

Total (in) 44.42 44.92 

Hours 602 650 

Frequency (%) 6.87% 7.42% 

Stability Class 

Frequency (%) 

Unstable 17.75% 7.33% 

Neutral 24.57% 47.91% 

Stable 57.68% 44.76% 

Joint Data Recovery (%) 99.25% 95.19% 
4-19

Figure 4-1 Wind Rose for Peach Bottom 2006 and Limerick 2013

The methodology described in NUREG-0917, “Nuclear Regulatory Commission Staff Computer 

Programs for Use with Meteorological Data” [25] was used to perform quality assurance 

evaluations of all meteorological data. In accordance with Regulatory Guide 1.23, “Meteorological 

Monitoring Programs for Nuclear Power Plants,” [26], a data recovery rate greater than 90% was 

achieved for the wind speed, wind direction, and atmospheric stability parameters. In addition, 

atmospheric stability data was reviewed to determine if the time of occurrence and duration of 

reported stability conditions were generally consistent with expected meteorological conditions 

(e.g., neutral and slightly stable conditions predominating during the year with stable and neutral 

conditions occurring at night and unstable and neutral conditions occurring during the day).

The weather data was also reviewed against data from other years to assess how representative 

the chosen year seemed for the site’s climate. Data from other sources, such as annual effluent 

reports and dose assessment reports, for ~10 other years were analyzed and compared to the 

selected year and showed that they were indeed representative of the climate conditions.

4.2.4.2 Mixing Height Data

In addition to hourly observation data, the MACCS meteorological file requires morning and 

afternoon mixing height data for four meteorological seasons, for a total of eight entries. The 

morning mixing height is the minimum mixing height used in the code, and the afternoon mixing 

height is the maximum mixing height. MACCS uses the site longitude and latitude coordinates to 

determine the time of sunrise and sunset and it estimates the mixing height by linear interpolation 

between the minimum and the maximum, based on the time of day. Mixing height data is based 

on upper air measurements that are only available at selected locations across the United States.

The mixing height data used for this analysis came from the U.S. Environmental Protection 

Agency’s (EPA’s) SCRAM database [27] for the nearest weather stations to Peach Bottom and 

Limerick at Pittsburgh, PA, and Sterling, VA, for the three most recent years available, 1989-1991.

Data from January through March was used for winter, April-June for spring, July-September for 

summer, and October-December for autumn. In addition, the EPA report, “Mixing Heights, Wind 

Speeds, and Potential for Urban Air Pollution throughout the Contiguous United States,” [28], 

0%

5%

10%

15%

N

NNE

NE

ENE

E

ESE

SE

SSE

S

SSW

SW

WSW

W

WNW

NW

NNW

Peach Bottom Wind Rose 2006 

(Toward)

0%

5%

10%

15%

20%

N

NNE

NE

ENE

E

ESE

SE

SSE

S

SSW

SW

WSW

W

WNW

NW

NNW

Limerick Wind Rose 2013 (Toward)
4-20

which provides mixing height contour maps across the U.S., was used to provide an estimate of 

the mixing heights. Data from these sources were averaged to yield the mixing heights used in the 

analysis. The values are rounded to the nearest tens of meters to comply with the MACCS 

formatting requirements and these are provided in Table 4-11. 

Table 4-11 Seasonal diurnal mixing heights (m) used for Peach Bottom and Limerick

4.2.4.3 Weather Bin Sampling

In MACCS, a weather trial is defined by the starting hour of weather data from the meteorological 

data file and each trial uses as many of the following hours as are required to transport plume 

segments through and out of the computational grid. Rather than using all 8,760 available weather 

trials (corresponding to each starting hour of the year) in an offsite consequence analysis, weather 

bins are defined to categorize similar sets of weather conditions, and a sampling approach is used 

to randomly select the weather trials within each bin. This analysis used the nonuniform weather 

bin sampling approach and yielded approximately 1,000 weather samples based on 36 weather 

bins. The 36 weather bins are based on wind speed, stability class, and the occurrence of 

precipitation. Sixteen bins were defined based on combinations of stability class and wind speed 

and the remaining 20 bins were defined based on rain occurrence and intensity at various 

downwind locations within 20 miles of the site. The parameters used to define the bins are the 

same as those used in previous studies such as SOARCA [9] and are documented in 

NUREG/CR-7009 [13]. 

The nonuniform sampling approach allows the user to specify a different number of random 

samples for each bin. Consistent with the SOARCA project, the number of trials selected from 

each bin is the maximum of 12 trials and 10% of the number of trials in the bin. For bins 

containing fewer than 12 trials, all of the trials within the bin are used for sampling. This strategy 

resulted in 984 weather trials for Peach Bottom and 963 for Limerick. 

4.2.4.4 Boundary Weather

Boundary weather refers to the use of artificial weather conditions for the outermost ring of the 

calculational grid. Continuous rain is specified in the ring spanning 500-1000 miles as a way to 

ensure that all of the radionuclides released into the atmosphere will be deposited within the 

computational domain. This is clearly unrealistic and therefore consequences for the 500-1000-

mile ring are not presented in this report.

4.2.5 Atmospheric Transport and Dispersion

MACCS models dispersion of radioactive materials released into the atmosphere using the 

straight-line Gaussian plume segment model with provisions for meander and surface roughness 

effects. The ATD model treats the following: plume rise resulting from the sensible heat content 

(i.e., buoyancy), initial plume size caused by building wake effects, release of up to 200 plume 

segments, dispersion under statistically representative meteorological conditions, deposition 

under dry and wet (precipitation) conditions, and decay and ingrowths of up to 150 radionuclides 

Winter Spring Summer Fall 

Morning 760 650 500 570 

Afternoon 1000 1700 1680 1130 
4-21

and a maximum of six generations. The model does not treat in detail irregular terrain and spatial 

variations in the wind field. The model considers temporal variations in wind direction at the time 

each plume segment is released. Once a plume segment moves in the direction of the initial wind, 

the plume segment is modeled to continue in that same direction even if the wind field changes.

However the next plume segment released at the next time step (typically 1 hour) would move in a 

different direction if the wind field changed.

4.2.5.1 Dispersion Parameters

The Gaussian plume segment model uses two spatially dependent dispersion parameters to estimate 

the atmospheric dispersion, σy for the horizontal, cross-wind dispersion and σz for the vertical 

dispersion. These parameters are specified for each stability condition as a function of downwind 

distance and can be modeled in two ways, either using power-law functions or a lookup table.

The growth of plume dimensions during downwind transport to short distances (1 km) has been

experimentally determined [29] over flat terrain covered by prairie grass for short release 

durations (10 min) during stable, neutral, and unstable atmospheric conditions. Pasquill used this 

data to develop curves that depict the increase of plume dimensions (σy andσz values) with 

downwind distance for each of the six Pasquill-Gifford stability classes used in MACCS, A-F.

Although measurements had only been made to 1 km, Pasquill extrapolated the curves to 100 km 

and they have subsequently been extrapolated farther. Tadmor and Gur [30] developed a powerlaw correlation to fit to the original experimental data. Eimutis and Konicek developed a separate 

correlation to better represent the different power-law fit coefficients for different distance ranges 

[31]. The Eimutis and Konicek formulation was recently converted into a MACCS lookup table by 

Bixler, Napier, and Rishel [32]. 

For horizontal, cross-wind dispersion, the coefficient has a relatively constant slope on a log-log 

plot of distance and both the power-law and lookup table approaches approximate the prairie 

grass experimental data similarly well. However, the dispersion coefficient has a more nonlinear 

slope as a function of distance and stability class on a log-log plot and therefore the lookup table 

approach can approximate vertical dispersion better than power law functions. Despite this, power 

law functions have the advantage of requiring far fewer values to specify dispersion, and therefore 

uncertainty analysis sampling is made much more efficient. Power-law functions based on 

Tadmor and Gur have been commonly used in past studies such as SOARCA for this reason.

Uncertainty in these parameters, along with a variety of others, was propagated in the SOARCA 

Peach Bottom Uncertainty Analysis [14]. Because this project is not aimed at re-characterizing 

dispersion parameter uncertainty, the lookup table approach based on the Bixler et al. conversion 

of the Eimutis and Konicek formulation was used because it best approximates vertical dispersion 

in addition to horizontal dispersion.

4.2.5.2 Surface Roughness

The surface roughness of a land area characterizes the amount of interaction a plume would have 

with the ground based on topographical features such as row crops, trees, and houses. MACCS 

requires a single surface roughness value for each site studied. To select a single value, the 

USDA CropScape database [33] was accessed, which provides land use data for the EPA land 

use categories [34] for any user-specified area in the continental U.S. The land use area fractions 

were used to weight various typical land use surface roughness values to yield one representative 

value for each site. This process is summarized in Table 4-12, which led to a value of 26 cm for 

Peach Bottom and 33 cm for Limerick. Land use data for a circular area of approximately 30-mile 

radius from each site was chosen because the 30-mile distance is considered a reasonable 
4-22

estimate of the distance plumes travel before becoming well-mixed between the ground and the 

mixing height layer. The typical surface roughness values for the 10 land types are from 

DOE/RL/87-09, “The Remedial Action Priority System (RAPS): Mathematical Formulations, 1987 

[35] and NUREG/CR-7110 [10].

Table 4-12 Surface roughness calculation for Peach Bottom and Limerick

In MACCS, surface roughness affects both vertical dispersion and dry deposition velocities. The 

effect on vertical dispersion has traditionally been modeled by means of a multiplicative factor.

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 [29], 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 for Peach Bottom corresponding to a 26 cm surface roughness is 

(26 / 3)0.2 = 1.54 and for Limerick corresponding to a 33 cm surface roughness is 1.62. The effect 

of surface roughness on dry deposition velocities is described in the following section.

4.2.5.3 Dry Deposition

The dry deposition velocity of an aerosol particle is a function of particle size and of the degree of 

turbulence in the atmosphere, which is affected by wind speed and surface roughness. The effect 

of surface roughness on deposition velocity has been characterized by Bixler et al. in NUREG/CR7161 [36] based on expert elicitation data in NUREG/CR-6545 [37]. 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. This correlation is 

valid for aerosol diameters up to about 20 μm and surface roughness up to about 60 cm. For the 

largest particle size bin of 41.2 μm, the correlation is no longer valid so the effect of gravitation 

settling alone is considered. The mean wind speed used in the correlation was taken from the 

year of weather data selected, 2006 for Peach Bottom (2.12 m/s) and 2013 for Limerick (2.36 

Land Use Type Typical Surface 

Roughness (cm) 

Peach Bottom 

Land Fraction 

Limerick Land 

Fraction 

Open Water 0.03 2.12% 0.77% 

Barren 1 0.15% 0.13% 

Grass/Pasture 3 23.46% 17.38% 

Developed/Open Space and 

Developed/Low Intensity 5 17.39% 27.50% 

Shrubland 5 1.45% 2.65% 

Wetlands 5 2.41% 0.40% 

Farmland 14 20.73% 12.76% 

Forest 60 29.40% 31.12% 

Developed/Medium Intensity 70 2.06% 4.96% 

Developed/High Intensity 300 0.84% 2.32% 

Weighted Average Site Surface Roughness: 26 cm 33 cm 
4-23

m/s). Table 4-13 provides the set of dry deposition velocities used in the analysis for Peach 

Bottom and Limerick for the 10 aerosol particle size bins calculated by MELCOR.

Table 4-13 Dry deposition velocities for Peach Bottom and Limerick

4.2.5.4 Wet Deposition

Wet deposition is an important phenomenon that strongly affects atmospheric transport. Under 

heavy rains, wet deposition rapidly depletes the plume. Even under light rains, the plume is 

depleted much faster than by dry deposition alone. The wet deposition process can produce 

concentrated deposits on the ground and create what is often referred to as a hot spot (i.e., an 

area of higher radioactivity than the surrounding areas). While rain occurs less than 10% of the 

time for most of the U.S., it can significantly affect consequence calculations when it does occur.

The wet deposition model predicts how much radioactive material is deposited on the ground by 

rainfall. Wet deposition model parameters were derived based on expert elicitation data [36]. The 

linear washout coefficient was selected to be 1.89E-5 per second and the exponential washout 

coefficient was selected to be 0.664.

4.2.5.5 Plume Meander

Plume meander refers to the broadening of the plume in the crosswind (σy) direction as a result of 

wind direction fluctuations. MACCS provides three options for plume meander modeling. The first 

option is the original MACCS plume meander model, which considers plume segment duration 

and derives a meander scaling factor by comparing to the Pasquill-Gifford experimental data. The 

second option is based on NRC Regulatory Guide 1.145, “Atmospheric Dispersion Models for 

Potential Accident Consequence Assessments at Nuclear Power Plants,” [38] and described in 

more detail in NUREG/CR-2260 [39]. This option is based on plume segments of one hour 

duration, which is the same as the plume segment durations used in this study. The third option is 

to not model plume meander. The Regulatory Guide 1.145 model considers stability class and 

wind speed for the meander factor and was selected for this analysis because it represents the 

NRC’s most recent evaluation of plume meander for consequence analysis. This model requires 

user-specified values of wind speed, distance, and meander factor; all selected values are taken 

Particle Size Bin Particle 

Diameter 

(μm) 

Dry Deposition Velocities (cm/s) 

Peach Bottom Limerick 

1 0.15 0.095 0.115 

2 0.29 0.087 0.105 

3 0.53 0.114 0.138 

4 0.99 0.192 0.233 

5 1.84 0.375 0.456 

6 3.43 0.767 0.932 

7 6.38 1.480 1.798 

8 11.88 2.427 2.948 

9 22.12 3.005 3.650 

10 41.18 5.151 5.151 
4-24

directly from Regulatory Guide 1.145. The effect of plume meander using this model is greatest 

around 800 m from the site and diminishes farther from the site.

4.2.6 Early Phase Exposure Pathways, Protective Actions, and Costs

The early phase model in MACCS assesses the time period immediately following a radioactive 

release. The EPA PAG Manual describes this phase as “the beginning of a radiological incident 

when immediate decisions for effective use of protective actions are required and must therefore 

be based primarily on the status of the radiological incident and the prognosis for worsening 

conditions”. This period is commonly referred to as the emergency phase and it can last for days 

to weeks following the start of the accident. Early exposures account for emergency planning (i.e., 

sheltering, evacuation, relocation of the population, and ingestion of potassium iodide). MACCS 

models sheltering, evacuation, and KI ingestion for user-specified population cohorts. Different 

shielding factors for the different exposure pathways (cloudshine, groundshine, inhalation, and 

deposition on the skin) are associated with three types of activities: normal activity, sheltering, and 

evacuation. 

4.2.6.1 Emergency Response Timeline

Emergency response programs for nuclear power plants are designed to protect public health and 

safety in the unlikely event of a radiological accident. These emergency response programs are 

developed, tested, and evaluated and are in place as an element of the NRC’s defense-in-depth 

policy. Detailed plans for onsite and offsite response are approved by the NRC and the Federal 

Emergency Management Agency respectively. Offsite response organization (ORO) emergency 

plans are required to include detailed evacuation plans for the 10-mile plume exposure 

emergency planning zone (EPZ) [40]. Site-specific information was obtained from ORO 

emergency response plans to support development of timelines for protective action 

implementation. Site-specific planning elements were modeled, for example whether evacuation 

of schools follows declaration of a site area emergency (SAE) or a general emergency (GE). 

One emergency response timeline was developed for each reference site for all accident 

scenarios using information from the MELCOR analyses, expected timing of emergency 

classification declarations, and information from the evacuation time estimate (ETE) reports. The 

timeline identifies points at which population 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. In contrast, it is 

expected that nuclear power plants in most if not all other states would implement a keyholeshaped evacuation consisting of an inner circular region and a region extending outward based on 

the expected direction of the prevailing winds.

One of the objectives of the SOARCA project was to model emergency response in a more 

detailed and realistic manner than past studies, using site-specific emergency planning 

information. Therefore, for each scenario at each SOARCA pilot plant, a unique emergency 

response timeline was developed and was reviewed with the licensee for accuracy. This project

seeks to model a significantly larger number of accident progression scenarios compared to 

SOARCA. Therefore, to develop the timeline for implementation of protective actions, the accident 

initiator and expected accident progression conditions and timing were reviewed for similarity to 

the SOARCA scenarios. The initiating event in this analysis is an extended loss of alternating 

current (AC) power (ELAP) postulated to be caused by an unspecified internal event or seismic 

event. In the majority of the MELCOR simulations, the reactor core isolation cooling (RCIC) 

system is assumed to be available for at least 4 hours and therefore the series of events closely 
4-25

aligns with the accident timeline of the SOARCA Peach Bottom unmitigated long-term station 

blackout (LTSBO). Thus the SOARCA Peach Bottom unmitigated LTSBO emergency response 

timeline was adopted for this project and is used for Peach Bottom and Limerick because they 

would both follow the same Pennsylvania-specific guidance. The timeline is described in detail in 

Table 4-14. 

Table 4-14 Emergency response timeline

4.2.6.2 Evacuation Modeling

MACCS now enables two different types of evacuation regions. The first type is an evacuation of 

the full 360° EPZ region surrounding the plant. The second type is a recent advancement to 

MACCS and involves a keyhole-shaped evacuation region consisting of an inner circular region 

and a region extending outward in the direction of the prevailing winds. Most states are expected 

to implement a keyhole-shaped evacuation region; however, Pennsylvania would likely implement 

a full 360° EPZ region according to officials with the Pennsylvania Emergency Management 

Agency (PEMA). Because both reference plants for this project, Peach Bottom and Limerick, are 

in Pennsylvania, the full EPZ evacuation region approach is used.

MACCS input parameters related to evacuation modeling were developed primarily from the sitespecific ETE reports, which are required to be developed and updated by the licensee under 10 

CFR 50.47 (b)(10), “Emergency Plans.” ETEs provide the time required to evacuate various 

sectors and distances within the EPZ for transient and permanent residents, and these times are 

used to develop response timing and travel speeds for evacuating cohorts in MACCS. The ETEs 

provided the mobilization times and travel times of different segments of the population as well as 

evacuation routing information. The guidance in NUREG/CR-7002 [41] describes the detail 

included in an ETE study. Important information in an ETE report includes demographic and 

response data for the following four population segments and may be readily converted into 

cohorts, if appropriate: (1) permanent residents and transient population, (2) transit-dependent 

permanent residents—people who do not have access to a vehicle or are dependent upon help 

from outside the home to evacuate, (3) special facility residents—people in nursing homes, 

Time Event 

0:00 ELAP 

0:15 SAE is declared via MS1 of the SAE Emergency Action Level (EAL) based 

on loss of all AC power. 

0:45 

GE is declared based on EAL MG1, 45 minutes into the event (coincidentally 

15 minutes before the issue of the first Emergency Alert System (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 which would include instructions for 

implementing protective actions. 

1:00 

Sirens for SAE sound about 45 minutes after SAE declaration. 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. 

1:30 Sirens for GE sound again 45 minutes after the GE declaration, which is 30 

minutes after the siren and initial EAS message for the SAE. 
4-26

assisted living centers, hospitals, jails, prisons, etc., and (4) schools, including all public and 

private educational facilities within the EPZ.

The ETE typically includes about 10 scenarios that vary by season, day of the week, time of day, 

and weather conditions, as well as other EPZ-specific situations such as special events. The 

ETEs do not consider a large seismic event and its impact on road infrastructure, which could be 

substantial. This study considers the impact of a seismic event slowing evacuation in a set of 

sensitivity calculations described in Section 4.4.4. Consistent with past analyses such as 

SOARCA for Peach Bottom and Surry, the ETE scenario for a winter weekday, mid-day, good/fair 

weather accident was used for the development of MACCS evacuation model parameters. This 

scenario includes residents at work and children at school at the time of declaration of the 

emergency.

ETEs compute the vehicle demand for each scenario based on population information from the 

licensee, census reports, and telephone surveys of local residents. These telephone surveys are 

also used to estimate the mobilization times for each population group, which is the time to learn 

of the event and prepare to evacuate. ETE studies use detailed link-node representations of the 

road network to estimate all the routing pathways and the travel speeds of the public through the 

road network. These studies consider that the public mobilizes and evacuates over a period of 

time as a distribution of data. In contrast, MACCS models the evacuation process as a series of 

discrete events for each population cohort. The use of multiple cohorts allows a more realistic 

modeling of the evacuation process and helps to better represent the evacuating public as a 

distribution.

The initial response parameter in MACCS is OALARM, which can be used to uniformly adjust 

protective action timing for all cohorts. Consistent with the SOARCA analysis of Peach Bottom, 

the OALARM time is set to zero, the time of accident initiation and reactor scram. For each 

population cohort, MACCS requires the following duration and travel speed parameters:

• Delay to shelter (DLTSHL) represents the duration of time from the accident initiation

until the population learns of the event and begins sheltering. During this period

shielding parameters are applied assuming normal activity.

• Delay to evacuate (DLTEVA) represents the duration of time from the start of sheltering

to the start of evacuation. During this period, shielding parameters are applied assuming

sheltering.

• Duration of beginning phase (DURBEG) of evacuation is the first of three time periods

and is used to represent the time of travel from one’s starting point until they are in the

evacuation queue. This period begins when the sheltering period ends. During this

phase a travel speed (ESPEED1) is applied for each cohort. For each of the three

evacuation phases, shielding parameters are applied assuming evacuation.

• Duration of middle phase (DURMID) of evacuation is used to represent the time of travel

after DURBEG to exit the EPZ. This period begins when DURBEG ends. During this

phase a second travel speed (ESPEED2) is applied for each cohort.

• The third and final phase of evacuation is defined as the period of time from the end of

DURMID to the end of the early phase (7 days from the time of the first radiation

release). During this phase a third travel speed (ESPEED3) is used to represent the

travel speed outside the EPZ on large roads such as interstate highways. The population
4-27

travels at ESPEED3 until they have reached a distance of 50 miles from the nuclear 

power plant, at which point they are modeled to receive no further early phase radiation 

exposure.

For both Peach Bottom and Limerick, the following cohorts are used:

1. Schools within 0-10 miles. This cohort includes elementary, middle, and high school

student populations within the EPZ. Based on Pennsylvania protocols, schools would

receive early and direct warning from OROs and have response plans in place to

support busing of students out of the EPZ. The ETE provides considerable detail

regarding schools, including the number and location of schools, student and staff

population, number of buses required to evacuate the students, and the ETE, which

considers whether return trips are required. Students evacuated by bus would be

brought to pre-designated reception centers where parents would pick them up.

Preschools and daycare facilities are described in the ETE but are not included in this

cohort because, unlike elementary school students, younger children would be expected

to be picked up by parents or caregivers and evacuated with the family and not

evacuated on buses. College students are also not included in this cohort because they

would be expected to evacuate largely via private cars rather than by buses.

2. Special Facilities within 0-10 miles. The special facilities population includes residents of

hospitals, nursing homes, prisons, etc. This cohort was modeled differently for Peach

Bottom and Limerick and is therefore discussed in the following sections of site-specific

evacuation characteristics.

3. Transit-Dependent Residents within 0-10 miles. This population refers to evacuees who

do not have access to transportation and confined persons who require special

transportation assistance. This population is specifically described and modeled in the

ETE reports. Population estimates are based upon U.S. Census data, results of a

telephone survey, and for Peach Bottom on data provided by the Lancaster County

Geographic Information System (GIS) Department. The ETEs assume a fraction of this

population will evacuate with a neighbor or friend and the rest will be evacuated by bus

to a reception center outside the EPZ.

4. Early Evacuees within 0-10 miles. Pennsylvania guidance is for sirens to be sounded at

SAE. This is a local decision implemented by the OROs. This evacuation is considered

early because it is before the OROs officially instruct the public to evacuate. A

population fraction of 20% was selected based on data from a national telephone survey

of residents of EPZs [42].

5. General Public within 0-10 miles. This cohort represents the population that is not

included in any of the other 0-10-mile cohorts. The population fraction is calculated by

subtracting the population fractions for all of the other 0-10-mile cohorts.

6. Tail within 0-10 miles. The 0-10-mile tail is defined as the last 10% of the public to

evacuate [43] from the 10-mile EPZ. The evacuation tail takes longer to evacuate for

valid reasons, such as shutting down farming or manufacturing operations, performing

other time consuming actions prior to evacuating, or they may have missed the initial

notification.
4-28

7. Nonevacuating Public within 0-10 miles. This cohort represents a portion of the public

residing in the EPZ who would either not receive any notification of the event or would

choose not to evacuate. SOARCA and other studies have assumed this group to be

0.5% of the population. This percentage is consistent with research on large-scale

evacuations that has shown a small percentage of the public refuses to evacuate [44].

This value is used in the Limerick evacuation model; however, a higher value is used for

Peach Bottom reflecting the expected response of the Pennsylvania Dutch (Amish)

community. This is described further in the following section.

8. Shadow Evacuees within 10-15 miles. A shadow evacuation occurs when members of the

public evacuate from areas that are not under official evacuation orders and typically begins

when a large-scale evacuation is ordered [44]. This study assumed that 20% of the

residents in the area between 10-15 miles of the nuclear power plant would evacuate

without being ordered to do so. The 20% value was obtained from results of a telephone

survey of residents of EPZs conducted by the NRC in 2008 [42]. The location of this cohort,

10-15 miles from the plant, was selected based on guidance in NUREG/CR-7002 [41].

A normal weather winter weekday ETE scenario was selected to develop the evacuation delay 

and travel speed parameters. However, real weather data is used in the calculation, so to 

represent the impact of adverse weather, the travel speed of each cohort is reduced when 

adverse weather occurs. Adverse weather is typically defined as rain, ice, or snow that affects the 

response of the public during an emergency. The meteorological data file includes hourly 

precipitation but does not distinguish between rain, ice, snow, etc. MACCS uses the evacuation 

speed multiplier (ESPMUL) parameter for each cohort to reduce the defined travel speed when 

precipitation is occurring, as indicated from the meteorological data file. The ESPMUL factor was 

set at 0.7, which effectively slows down the evacuating public to 70% of the established travel 

speed during the time precipitation occurs. This value was based on existing ETE guidance [41] 

and is consistent with SOARCA.

In addition to specifying evacuation travel speeds for each cohort, MACCS includes a network 

evacuation model that allows the user to select the direction of travel in each spatial grid element 

in the polar calculational grid. This feature was used in this analysis for Peach Bottom and 

Limerick. Maps of each site were displayed against the spatial grid, allowing selection of travel 

directions to represent the likely flow of traffic based on the road infrastructure.

4.2.6.2.1 Peach Bottom Site-Specific Evacuation Model Considerations

A summary of evacuation data for Peach Bottom is provided in Table 4-15. Eight evacuation 

cohorts were modeled based on the ETE report submitted by Exelon to NRC in June 2014 [22].

The eight cohorts and their general characteristics were described in the previous section. All 

eight cohorts are modeled as if the population exists uniformly across the entire EPZ. As 

described above, the nonevacuating cohort is used to represent a very small fraction of the public 

that would either not receive notification of the event or would choose not to evacuate. A value of 

0.5% was used in the SOARCA Peach Bottom analyses; however, this study uses a higher value 

based on information in the recently submitted ETE, which considered the expected response of 

the Pennsylvania Dutch (Amish) community. Based on discussions with Lancaster County 

emergency management officials, the Amish men (considered age 15 or older) would remain on 

their land while the women and children would evacuate in the event of an incident at Peach 

Bottom. The number of Amish men is approximately 1.5% of the EPZ population, so therefore 
4-29

2.0% was selected as the nonevacuating cohort fraction. The Amish women and children are 

modeled with the transit-dependent population. This is clearly a Peach Bottom site-specific 

consideration and the use of a larger than typical nonevacuating cohort fraction is discussed later 

in the results section.

Figure 4-2 shows the emergency response timeline for Peach Bottom in a chart to better illustrate 

the components of the timeline for each cohort and the differences among the cohorts. The time 

duration parameters (DLTSHL, DLTEVA, DURBEG, and DURMID) for the schools, special 

facilities, and transit-dependent cohorts were selected by reviewing the ETE, which models the 

mobilization and travel time for each individual school, medical facility, etc. in the EPZ. The time 

duration parameters for the tail cohort were selected to align the response with the ETE 

mobilization time and evacuation time for 100% of the EPZ to be cleared. For the selected ETE 

scenario (winter, midweek, mid-day, good/fair weather), the ETE 100% time was 3:55 from the 

time of notification. Assuming this cohort is notified via the GE siren and corresponding 

emergency alert system (EAS) messaging at 1:30, the total evacuation time is approximately 5:30 

as shown in the figure below. The time duration parameters for the rest of the cohorts were 

selected by approximating the mobilization and evacuation distribution curves in the ETE report. 

The evacuation travel speeds are shown in the figure below for the first two evacuation phases in 

units of miles per hour. They were selected considering the DURBEG and DURMID evacuation 

travel durations to make the travel distance approximately 10 miles. In reality, the centroid of 

population within the EPZ would not be very close to the plant, and thus the average distance to 

the EPZ boundary might only be about 5 miles (or less). However, the travel distance of 10 miles 

is approximated because the path of travel must follow the existing road network and is assumed 

to not be directly outward. For example, the fastest travel path may involve driving on a small local 

road toward the plant to then enter a limited-access highway that goes generally away from the 

plant at a higher speed. The evacuation travel speeds were also selected so that travel for all of 

the cohorts would be about the same at the same time. The first few cohorts to enter the road 

network travel at about 15 mph. Shortly after, a larger fraction of the population enters the roads 

and all speeds drop to about 5 mph on average. These travel speeds include the delays and 

queue times that occur at intersections. The tail cohort is the last group in the EPZ so they travel 

at a faster average speed (20 mph) because the roads are much less crowded than before. The 

10-15-mile shadow evacuation cohort travels at speeds different from the others because it is on

different road sections that are less congested.


7DEOH 3HDFK%RWWRPHYDFXDWLRQFRKRUWVDQGSDUDPHWHUV

3RSXODWLRQ 5HVSRQVH

'XUDWLRQVKUV

(YDFXDWLRQ

3KDVH'XUDWLRQV

KUV

(YDFXDWLRQ7UDYHO

6SHHGVPSK

&RKRUW7LWOH

&RKRUW

/RFDWLRQ

PL

3RSXODWLRQ

/RFDWLRQ

3RSXODWLRQ

)UDFWLRQ

'HOD\WR

6KHOWHU

'HOD\WR

(YDFXDWH (DUO\ 0LGGOH (DUO\ 0LGGOH /DWH

 6FKRROV ±         

 6SHFLDO

)DFLOLWLHV ±         

 7UDQVLW

'HSHQGHQW ±         

 (DUO\ ±         

 *HQHUDO ±         

 7DLO ±         

 1RQHYDFXHHV ±         

 6KDGRZ ±         

)LJXUH 3HDFK%RWWRPHPHUJHQF\UHVSRQVHWLPHOLQHDQGWUDYHOVSHHGVPSK
4-31

4.2.6.2.2 Limerick Site-Specific Evacuation Model Considerations

A summary of evacuation data for Limerick is provided in Table 4-16. Nine evacuation cohorts 

were modeled based on the ETE report submitted by Exelon to NRC in January 2014 [45].

Compared to the Peach Bottom evacuation model, an additional cohort was used for Limerick to 

distinguish between two types of special facility populations, those at medical and assisted living 

facilities vs. those at correctional facilities. (Peach Bottom has no correctional facilities in the 

EPZ.) According to the Limerick ETE, the 6,037 inmates (3,957 at the State Correctional Institute 

at Graterford in Graterford, PA and 2,080 at the Montgomery County Correctional Facility in 

Eagleville, PA) would shelter in place rather than evacuate, unlike the medical facility residents 

who would evacuate in the transit-dependent cohort. Because this significant fraction of the 

population would not evacuate and it exists in two specific locations, effort was taken to model 

them in the specific calculational grid elements where they exist, rather than applying the 

population uniformly across the EPZ as is done with the rest of the Limerick cohorts. The 

Graterford, PA prison inmates are all modeled in the 7.5-10-mile ring in compass sector 16 and 

the Eagleville, PA prison inmates are all modeled in the 7.5-10-mile ring in compass sector 22 

(compass sector 1 is centered on due north of the plant, compass sector 17 is centered on due 

east of the plant, etc.). The ability to model cohorts in specific grid elements is a recent 

advancement to MACCS and improves realism. Unlike in the Peach Bottom EPZ, the Amish 

community represents a negligible fraction of the population and is therefore not considered as 

part of the nonevacuating cohort in the Limerick evacuation model.

Table 4-16 Limerick evacuation cohorts and parameters

Population Response 

Durations (hrs)

Evacuation 

Phase 

Durations 

(hrs)

Evacuation Travel 

Speeds (mph)

Cohort Title

Cohort 

Location 

(mi)

Population

Location 

Population 

Fraction

Delay 

to 

Shelter

Delay to 

Evacuate Early Middle Early Middle Late

1 Schools 0 – 10 49,321 0.19 0.25 1.75 0.50 3.0 10 2 5

2 

Special 

Facilities 

(Medical)

0 – 10 2,741 0.011 0.25 3.00 0.50 2.0 2 2 5 

3 

Special 

Facilities 

(Correctional)

Individual 

Grid 

Elements

6,037 0.023 

(of EPZ) - - - - - - - 

4 TransitDependent 0 – 10 3,764 0.015 0.25 2.75 0.50 3.75 2 2 5 

5 Early 0 – 10 51,949 0.2 1.00 1.00 0.50 3.0 10 2 5

6 General 0 – 10 118,602 0.457 1.50 1.00 0.75 3.25 2 2 5

7 Tail 0 – 10 25,974 0.1 3.00 2.50 1.50 1.50 2 5 5

8 Nonevacuees 0 – 10 1,299 0.005 - - - - - - - 

9 Shadow 10 – 15 64,018 0.2 1.00 3.00 0.50 2.00 5 5 5

Figure 4-3 shows the emergency response timeline for Limerick in a chart to better illustrate the 

components of the timeline for each cohort and the differences among the cohorts. The time 

duration parameters and evacuation travel speeds were developed using the same approach as 

was described for Peach Bottom. Limerick has an ETE population more than five times higher 

than Peach Bottom; however, the ETE for 100% of the EPZ to be cleared for the same (winter, 

midweek, mid-day, good/fair weather) scenario is 6:50 from the notification time, just ~3 hrs longer 


WKDQ3HDFK%RWWRP7KH(7(IRU(3=FOHDUDQFHLVDOVRFRQVLGHUHGDXVHIXOPHWULFIRU

HPHUJHQF\UHVSRQVHRIILFLDOVDQGLVIRU/LPHULFNDQGIRU3HDFK%RWWRPERWKIURPWKH

WLPHRIQRWLILFDWLRQ2QHRIWKHSULPDU\UHDVRQVIRUWKHKLJKO\VXEOLQHDULQFUHDVHLQHYDFXDWLRQ

WLPHDVDIXQFWLRQRISRSXODWLRQLVWKDWWKH/LPHULFNURDGQHWZRUNLVODUJHUDQGFDQDFFRPPRGDWH

DKLJKHUYHKLFOHIORZUDWHRQDYHUDJHFRPSDUHGWR3HDFK%RWWRP

)LJXUH /LPHULFNHPHUJHQF\UHVSRQVHWLPHOLQHDQGWUDYHOVSHHGVPSK

6KLHOGLQJ

,Q0$&&6VKLHOGLQJDQGSURWHFWLRQIDFWRUVDUHVSHFLILHGIRUHDFKGRVHSDWKZD\DQGGLUHFWO\

DIIHFWWKHGRVHVUHFHLYHGE\LQGLYLGXDOVDWHDFKORFDWLRQ7KHVKLHOGLQJDQGSURWHFWLRQIDFWRUVDUH

XVHGDVPXOWLSOLHUVRQWKHGRVHWKDWDSHUVRQZRXOGUHFHLYHLIWKHUHZHUHQRVKLHOGLQJ7KXVD

VKLHOGLQJIDFWRURIRQHUHSUHVHQWVWKHOLPLWLQJFDVHRIDSHUVRQUHFHLYLQJWKHIXOOGRVHLH

VWDQGLQJRXWGRRUVDQGFRPSOHWHO\XQSURWHFWHGIURPWKHH[SRVXUHDVKLHOGLQJIDFWRURI]HUR

UHSUHVHQWVWKHOLPLWLQJFDVHRIFRPSOHWHVKLHOGLQJIURPWKHH[SRVXUH

7KUHHW\SHVRIDFWLYLW\QRUPDOVKHOWHULQJDQGHYDFXDWLRQDUHHYDOXDWHGIRUHDFKGRVHSDWKZD\

IRUHDFKSRSXODWLRQFRKRUWZKLFKHYDFXDWHVZKLOHMXVWRQHW\SHRIDFWLYLW\LVHYDOXDWHGIRUWKH

QRQHYDFXDWLQJSXEOLF,QWKLVFRQWH[WQRUPDODFWLYLW\UHIHUVWRDFRPELQDWLRQRIDFWLYLWLHVWKDWDUH

DYHUDJHGRYHUDW\SLFDOZHHNDQGRYHUWKHSRSXODWLRQLQFOXGLQJEHLQJLQGRRUVDWKRPH

FRPPXWLQJEHLQJLQGRRUVDWZRUNDQGEHLQJRXWGRRUV7KHVHYDOXHVDUHDSSOLHGWRWKHWLPH

SHULRGVLQ7DEOHDQG7DEOHDQGFRQWLQXHIRUHYDFXHHVXQWLOWKH\KDYHWUDYHOHGWRDSRLQW

PLOHVIURPWKHQXFOHDUSODQWIRUQRQHYDFXHHVWKH\DUHDSSOLHGXQWLOWKHHDUO\SKDVHFRQFOXGHV

GD\VIURPWKHVWDUWRIWKHUDGLRQXFOLGHUHOHDVHWRWKHHQYLURQPHQWRUXQWLOWKH\UHORFDWH7DEOH

DQG7DEOHVKRZWKHFORXGVKLQHVKLHOGLQJIDFWRUJURXQGVKLQHVKLHOGLQJIDFWRULQKDODWLRQ
4-33

protection factor, and skin protection factor values used in this analysis. The values in Table 4-17

are consistent with those used for Peach Bottom in SOARCA [10] and were developed for 

NUREG-1150 [46] considering region-specific housing stock. The development of shielding 

parameters for NUREG-1150 is described in greater detail in NUREG/CR-4551, Vol. 2, Rev. 1, 

Part 7, “Evaluation of Severe Accident Risks: Quantification of Major Input Parameters: MACCS 

Input” [47]. Special facilities are typically larger and more robust structures than residential 

housing stock and therefore shielding factors are lower than for other population cohorts. 

Table 4-17 Shielding and protection factors for evacuating cohorts

Population 

Facility 

Type

Cloudshine Shielding Inhalation and Skin Protection Groundshine Shielding

Normal 

Activity Sheltering Evacuating Normal Activity Sheltering Evacuating Normal Activity Sheltering Evacuating

Regular 

Facilities 0.60 0.50 1.00 0.46 0.33 0.98 0.18 0.10 0.50

Special 

Facilities 

(Medical)

0.31 0.31 1.00 0.33 0.33 0.98 0.05 0.05 0.50

Table 4-18 Shielding and protection factors for nonevacuating population

Population and Facility Type Cloudshine 

Shielding

Inhalation and 

Skin Protection

Groundshine 

Shielding

Limerick Special Facilities (Correctional) 0.31 0.33 0.05

Nonevacuees Within EPZ 0.60 0.46 0.18

Remaining Population Beyond EPZ 0.50 0.33 0.10

For the nonevacuating population groups, only one shielding value for each exposure pathway is 

defined. For the Limerick special facilities cohort corresponding to the two correctional facilities in 

Graterford, PA and Eagleville, PA, shielding factors were selected to be consistent with the normal 

activity/sheltering shielding for the medical facilities. For the nonevacuees within the EPZ, 

shielding factors were selected to be consistent with the normal activity values for regular facility 

populations. Finally, for the remaining population outside the EPZ (80% of the population between 

10-15 miles and everyone beyond 15 miles), shielding factors were selected to be consistent with

the sheltering values for regular facilities. This is based on the expected protective action

recommendation of the OROs for the remaining affected population to shelter in place and

monitor the situation as it evolves based on television, phone, radio, or other means.

4.2.6.4 Potassium Iodide

Some States have distributed potassium iodide (KI) tablets to people who live near commercial 

NPPs. KI has been distributed within the Peach Bottom and Limerick EPZ according to officials with 

PEMA. The purpose of the KI is to saturate the thyroid gland with stable iodine so that further uptake 

of radioiodine by the thyroid is diminished. If taken at the right time, KI can nearly eliminate doses to 

the thyroid gland from inhaled radioiodine. Ingestion of KI is modeled for half of the EPZ population 

for each site, applied uniformly across all population cohorts. A further assumption is that most 

residents do not take KI at the optimal time (from shortly before to immediately after plume arrival), 

so the efficacy is only 70% (i.e., the thyroid dose from inhaled radioiodine is reduced by 70%). KI 

ingestion is not modeled for any of the population beyond 10 miles from each site.
4-34

4.2.6.5 Early Phase Relocation

In addition to evacuation, sheltering, and KI ingestion, OROs would implement the additional 

protective action of relocation. In the event of an accident at the NPP, OROs would use State, 

utility, and Federal agency computer models and field measurements to project dose levels in the 

area considering available source term information and current and projected meteorology. Based 

on these dose projections, OROs would inform the affected population directly and through 

available communication channels and instruct them to relocate. The U.S. Environmental 

Protection Agency (EPA) Protective Action Guides (PAGs) [48] provide guidance to OROs for 

relocation based on a range of 1 to 5 rem dose projected over a 4-day period. In reality, relocation 

would be instructed on a gradual scale, starting with those most at risk and then moving to 

populations at progressively lower risk levels. In MACCS, emergency phase relocation is modeled 

with two user-specified dose criteria to trigger the action, and a relocation time for the population 

affected by each dose. The higher dose level, used first, is set to the maximum point of the range 

in the EPA PAGs (5 rem projected over a 4-day period) and is referred to as the hotspot relocation 

dose criterion. The lower dose level, used next, is set to the lower end of the range in the EPA 

PAGs (1 rem projected over a 4-day period) and is referred to as the normal relocation dose 

criterion. The projected dose in MACCS would include the cloudshine, groundshine, direct 

inhalation, and resuspension inhalation exposure pathways and also considers shielding. In 

MACCS, individuals who reside in a grid element where either relocation dose criterion is 

exceeded, and who are modeled to have not already evacuated, are exposed for a user-specified 

time duration and then are removed from the computational grid so that they receive no further 

early phase dose. This time duration depends on the source term and on the size of the 

population affected. Larger source terms would trigger relocation farther from the plant than 

smaller source terms and areas with larger population would take longer to notify and to evacuate 

than those with a smaller population.

Table 4-19 provides the relocation dose criteria and exposure times for Peach Bottom and 

Limerick used in this project. The dose criteria were selected based on the upper and lower end of 

the range of dose levels specified in the EPA PAGs. The exposure times were selected 

considering the range of areas and population that might reach either dose criterion. The hotspot 

relocation exposure time duration was selected to account for relocation that would apply to the 

closest group of people that would not have already evacuated, those in the 10-20-mile rings. The 

normal relocation exposure time duration was selected to account for relocation that would apply 

to the population farther from the site, approximately 20-30 miles and potentially farther.

Table 4-19 Early phase relocation parameters

NPP Site Relocation Type Dose Criterion Exposure Time Duration 

After Plume Arrival

Peach Bottom Hotspot 5 rem 3 hrs

Normal 1 rem 24 hrs

Limerick Hotspot 5 rem 21 hrs

Normal 1 rem 48 hrs

The hotspot relocation exposure time duration considers the time when the 10-20-mile population 

would be advised to relocate, the time this population would take to mobilize and travel past 20 

miles, the time of the start of the release of radionuclides to the environment, and the travel time 

for the radionuclide plume to reach the 10-20-mile rings. There is a large degree of uncertainty 
4-35

with all of these times and they depend on the accident, the ORO decisionmaking process, the 

population affected, and the meteorological conditions. For Peach Bottom, and for the quickest 

environmental release timing (~ 7 hours corresponding to Mark I MELCOR case 49, which 

assumes a short-term station blackout with no batteries available and no RCIC), notification is 

estimated to align with the GE siren (1.5 hours), plume travel is estimated based on average 

annual wind speed (10 mi / ~5 mph = ~2 hours), and the mobilization and travel time is estimated 

at about 10 hours for 450,000 people. This leads to an exposure duration of about 3 hours. For 

Limerick, the same source term timing is assumed (7 hours) and the same plume travel time is

assumed (2 hours). Because Limerick has a much larger EPZ population, ORO officials would 

likely wait to advise relocation for the 10-20-mile population until the EPZ population would have 

time to leave first, since they would be at higher risk. The EPZ evacuation is modeled to conclude 

around 8.5 hours (1.5 hours for GE siren and 7 hours for the 100% ETE); therefore, the relocation 

notification is assumed at 9 hours. The mobilization and travel time for the 10-20-mile population 

is about 21 hours for 930,000 people, a linear extrapolation from the ETE of 7 hours for 300,000 

people. This leads to an exposure duration of about 21 hours for Limerick.

The normal relocation exposure durations are designed to simulate the potential for evacuation of 

the 20-30-mile ring and potentially farther. For Limerick, the population from 20 miles to 30 miles 

is approximately 2.8 million. To approximate the time to evacuate such a large population, the 

times for large-scale evacuations for hurricanes Rita and Ivan were reviewed. These hurricanes 

included populations in this range and as described in NUREG/CR-6981, “Assessment of 

Emergency Response Planning and Implementation for Large Scale Evacuations,” [49] the largest 

numbers of evacuees took on the order of 36 to 48 hours to complete their evacuation. Therefore, 

48 hours was selected as an approximate time to evacuate 2.8 million people from the 20-30-mile 

area. The notification timing for this population is even more uncertain than for the 10-20-mile 

area and thus 48 hours is estimated as the normal relocation exposure duration. The population 

for this ring around Peach Bottom is about 1 million, and therefore 24 hours is estimated for the 

Peach Bottom normal relocation based on extrapolation of 48 hours for Limerick for about three 

times higher population.

In addition to advising relocation for populations projected to exceed the EPA PAGs, OROs would 

advise the public to shelter in place. This protective action is captured in MACCS using a shielding 

factor for this group based on sheltering rather than normal activity.

4.2.6.6 Early Phase Costs

MACCS calculates the cost of the early (emergency) phase of the accident by multiplying a userspecified daily per person compensation cost by the number of people affected by the number of 

days each person is affected. The population affected includes those who evacuate in any of the 

defined evacuation population cohorts as well as those who would be relocated by exceeding 

either the hotspot or normal relocation dose criterion. The number of people and number of days 

affected is computed at the grid element level.

The daily compensation cost, defined in MACCS as EVACST, was selected considering the cost 

of lodging, food, and lost income. Studies show that the majority of people who evacuate their 

homes do not use shelter facilities for overnight stay [50]. People most often stay with relatives 

and friends or in hotels. For those who must leave their homes during the early phase, the costs 

of hotels and meals are additional costs that would not have been incurred under normal 

conditions. A typical federal per diem lodging rate is about $150 per night and considering the 

median household size of 2.58 persons per house from the U.S. Census, the lodging cost would 

be $58/person. This cost varies regionally and seasonally. For displaced people who stay with 
4-36

family or friends, there are other tangible and intangible costs associated with these stays. Food 

costs can vary considerably depending on the restaurant category. For this estimate, typical meal

prices for diner type restaurants were reviewed. An estimate of $35 per person per day includes 

the meals, beverage, tax, and tip. Restaurant costs also vary between rural and metropolitan 

areas, but the difference is not as significant as lodging. The lost income was selected based on 

the median annual household income and dividing by the number of working days in a year and 

by the median number of people per household. The US Census identifies the 2012 median 

annual household income as approximately $51,000. There are 365 days in a year, minus 104 

weekend days and 11 holidays, leaving 250 working days. Income divided by working days 

equals $204/per day per household and dividing this by 2.58 persons per household equals $79 

per person per day. Therefore the total evacuation compensation cost per person per day is $172 

($58 + $35 + $79). 

4.2.7 Intermediate Phase Exposure Pathways, Protective Actions, and Costs

The intermediate phase begins after the source and releases have been brought under control 

and accounts for the time needed to plan the long-term restoration and cleanup activities before 

they can begin. These activities include the following:

1. defining the areas of interest,

2. characterizing the contamination using dose data and field surveys,

3. identifying the types of materials to be decontaminated and the equipment and

personnel needed,

4. developing a waste management plan including estimating waste volumes, storage

requirements, storage locations, and acquiring storage materials,

5. acquiring decontamination equipment and bringing it onsite, and

6. training personnel and bringing them onsite.

The length of time to conduct these processes would depend on a variety of factors including the 

extent and location of the contamination, cleanup criteria, material types, and the state and local 

decision processes. As the distance from the plant increases, the level of contamination may 

decrease; however, other influences may increase because a wider variety of materials may be 

involved and additional towns, counties, and stakeholders would become involved.

Based on the number and types of planning processes that are needed prior to starting 

decontamination, 3 months was selected as an average intermediate phase duration and was 

used for all source terms in this analysis. Some decontamination would be expected to start prior 

to 3 months; however, much would likely take longer to begin. The first documented 

decontamination efforts after the March 11, 2011, Fukushima accident began with removal of 

contaminated soils from schools in May 2011, about 10 weeks after the accident [51]. In contrast, 

there is still much area for which decontamination has not begun years after the accident.

According to the Japanese Ministry of the Environment in an April 2014 presentation, “Progress 

on Off-site Cleanup Efforts in Japan,” seven of the eleven affected municipalities had not started 

decontamination work as of April 2012, over a year after the accident. In addition, one of the 

municipalities, Futaba, had not started decontamination work as of April 2014, over three years 

after the accident [52]. 

The only protective action modeled in the intermediate phase is relocation. If the projected dose to 

a population exceeds a user-specified threshold projected over the duration of the intermediate 
4-37

phase, the population is assumed to be relocated to an uncontaminated area for the entire 

duration of this phase. A corresponding per-capita per diem compensation cost is defined by the 

user. If the projected dose does not reach the user specified threshold, exposure pathways for 

groundshine and inhalation of resuspended material are modeled.

The intermediate phase relocation dose criterion was selected based on linear extrapolation of 

Pennsylvania’s annual habitability criterion of 500 mrem to the duration of 3 months, equaling 125 

mrem. MACCS calculates the cost of the intermediate phase of the accident very similar to the 

process for the early phase. A user-specified daily per person compensation cost is multiplied by 

the number of people affected by the number of days each person is affected (91 days). The 

intermediate phase daily per person compensation cost, defined in MACCS as RELCST, was 

selected to be identical to the early phase compensation cost, $172/person-day, because the 

costs incurred are similar. 

4.2.8 Long-Term Phase Exposure Pathways, Protective Actions, and Costs

In MACCS, the long-term phase starts at the conclusion of the intermediate phase and lasts for a 

user-specified duration, which was selected in this project as 50 years, consistent with SOARCA.

This period accounts for the time needed complete all recovery and restoration actions. Protective 

actions are implemented to minimize the dose to an individual by external (e.g., groundshine) and 

internal (e.g., food and water ingestion and resuspension inhalation) exposure pathways.

Protective actions are based on decisions relating to whether farmland, at a specific location (grid 

element) and time, is suitable for agricultural production (farmability) and whether nonfarmland, at 

a specific location and time, is suitable for human habitation (habitability). Habitability is defined by 

a user-specified maximum dose and a user-specified exposure period to receive that dose.

Habitability decision making for nonfarmland can result in four possible outcomes: 

1. land is immediately habitable,

2. land is habitable after decontamination,

3. land is habitable after decontamination and interdiction, or

4. land is not deemed habitable after decontamination plus 30 years of interdiction and is

therefore condemned.

Land is also condemned if the cost of decontamination exceeds the value of the land. The dose 

criterion for the MACCS modeling of individuals returning back to the affected (i.e., contaminated) 

area is based on a Pennsylvania state-specific dose criterion of 500 mrem-per-year for each year; 

whereas most other states would likely follow the guidance in the EPA PAGs [18] of 2 rem in the 

first year and 500 mrem-per-year each year thereafter.

The decision on whether land is suitable for farming is first based on prior evaluation of its 

suitability for human habitation—land cannot be used for agriculture unless it is habitable.

Furthermore, farmland must be able to grow crops or produce meat or dairy products that meet 

the user-specified farmability criterion, which is 500 mrem-per-year effective dose from ingestion.

This dose criterion was selected to be consistent with the Pennsylvania state-specific habitability 

criterion. If farmland is habitable and farmable, a food chain model is used to determine doses 

that would result from consuming the food grown or produced on this land. The COMIDA2 food 

chain model is the latest model developed for use in MACCS. This model contains data on 

expected radionuclide uptake in nine foodstuff types for different seasons of the year for different 

contamination levels and food category consumption rates for an average adult. 
4-38

Decisions on decontamination are made using a decision tree. The first decision is whether land is 

habitable. If it is, then no further actions are needed. The population returns to their homes and 

receive a dose not to exceed 500 mrem-per-year from any deposited radionuclides for the entire 

long-term phase. If land is not habitable, the first option considered is to decontaminate at the lower 

level of dose reduction, which is also cheaper to implement. If this level is sufficient to restore the 

land to habitability, then it is performed. Following the decontamination, the population returns to 

their homes and receives a dose not to exceed 500 mrem-per-year based on the residual 

contamination for the duration of the long-term phase. If the first level of decontamination is 

insufficient to restore habitability, then a higher and more expensive level is considered. If the higher 

level of decontamination is insufficient, then interdiction for up to 30 years is considered following the 

decontamination. During the interdiction period, radioactive decay and weathering reduce the dose 

rates that would be received by the returning population. If the higher level of decontamination 

followed by interdiction is sufficient to restore habitability, then it is implemented and the population 

is allowed to return. Doses are accrued for the duration of the long-term phase. If habitability cannot 

be restored by any of these actions, then the land is condemned. The land is also condemned if the 

cost of the required action to restore habitability is greater than the value of property.

To support the decision process of whether decontamination is cost-effective for a land use type 

for a grid element, MACCS requires one parameter to characterize the value of farm wealth 

(VALWF in $ per hectare) and one parameter to characterize the value of nonfarm wealth 

(VALWNF in $ per person). If the cost to decontaminate farmland in a grid element exceeds 

VALWF, then the farmland in the grid element is condemned. If the per person cost to 

decontaminate nonfarmland in a grid element exceeds the cost of condemning the property 

(VALWNF plus the cost to permanently relocate), then the nonfarmland in the grid element is 

condemned. If condemned, the actual cost of condemnation is taken from the site data file, 

discussed previously in Section 3.3, which includes a unique value of these parameters for each 

grid element. To identify one value for VALWF and one value for VALWNF that would represent 

the potential range of affected land, the individual values from the site file for each grid element 

out to 50 miles were weighted by the number of people and the area of farmland they contain.

This resulted in a weighted average farm wealth value of $24,400 per hectare and a weighted 

average nonfarm wealth value of $518,000 per person for Peach Bottom and $28,600 per hectare 

and $528,000 per person for Limerick.

MACCS values of total long-term population dose and health effects account for exposures received 

by workers performing decontamination. While engaged in cleanup efforts, workers are assumed to 

wear respiratory protection devices; therefore, they only accumulate doses from groundshine. Table 

20 provides the decontamination plan data used in this analysis for both Peach Bottom and 

Limerick. The farm costs ($ per hectare) and nonfarm costs ($ per person) are based on the 

NUREG-1150 values and are scaled from the NUREG-1150 target year, 1986, to this project’s 

target year, 2013, using the CPI-U inflator of 2.126. The MACCS decontamination model also 

requires a decontamination worker cost which is used to convert the total decontamination cost to a 

number of worker-years, which is then used to compute the total population dose to the 

decontamination workers. The decontamination worker cost was estimated at $76,000 per worker 

per year. This is based on the median annual wage for hazardous materials removal workers of 

$37,590 according to the U.S. BLS Occupational Outlook Handbook [53] and adding a factor of 

100% to cover overhead and all other non-direct labor costs.
4-39

Table 4-20 Decontamination plan data for Peach Bottom and Limerick

For population displaced from their home in the long-term phase, a one-time relocation cost is 

applied (POPCST). This cost is assessed if decontamination alone, decontamination followed by 

interdiction, or condemnation is required. The value is intended to account for personal and 

corporate income losses for a transitional period, as well as moving expenses. The value used in 

this analysis is $10,600 per person which is based on the same CPI-U inflator of 2.126 from the 

NUREG-1150 study, based on 1986, to this project’s target year, 2013. There is no analogous 

cost for farmland.

The cost of interdiction for nonfarmland includes a component to capture the loss of use of the 

property. For farmland, this is the only cost for interdiction. This equation considers the per capita 

or per hectare value (from the site file), a typical inflation-adjusted annual rate of return on 

investments, a typical depreciation rate, and a fraction of the per capita and per hectare value that 

is due to improvements (which are subject to depreciation). The depreciation rate (DPRATE) used 

is 20% per year, consistent with SOARCA and previous studies. Past studies used a 12% return 

on investments (DSRATE), which is not representative of the current investment climate.

Therefore a current DSRATE value was computed by a simple average of inflation-adjusted 

annual returns on two investment indexes over a period of greater than 20 years: the Case-Shiller 

U.S. National Home Price Index [54] for residential property and the National Council of Real 

Estate Investment Fiduciaries (NCREIF) Property Index [55] for commercial property. The 

average of the average annual inflation-adjusted return on each index resulted in a DSRATE of 

about 3%. The fraction of farm wealth in the region due to improvements (FRFIM, considered 

everything other than the value of the land itself) used was 0.25, consistent with SOARCA and 

past studies. The fraction of nonfarm wealth in the region due to improvements (FRNFIM) was 

selected to be 0.62 based on a study by Davis and Heathcote [56], which concluded that from 

1970 to 2003, the nominal value of land accounts for 38% of the nominal market value of homes. 

Finally, for farmland that exceeds the farmability criterion and is modeled to be temporarily unable 

to produce crops, meat, and dairy products that are suitable for human consumption, this food for 

the current growing season must be disposed. MACCS calculates these costs by multiplying the 

farm area of each affected grid element by the annual agricultural sales per hectare and summing 

this over all grid elements. The site data file includes a parameter to estimate the fraction of 

annual farm sales for dairy products and thus the disposal cost for dairy products is reported 

separately from the disposal cost for all other agricultural products.

4.2.9 Dosimetry

The approach used for dosimetry modeling in this analysis is identical to that used in the SECY12-0157 offsite consequence analysis [16]. The dosimetric quantities computed by MACCS for

use in modeling protective action decisionmaking or health effects are based on a dose

conversion factor approach. In general, the radiological dose to a receptor in a given spatial

element is the product of (1) the integrated air concentration or total ground deposition of a

radionuclide, (2) the exposure duration for an exposure pathway, (3) the shielding factor for an

Dose Reduction Factor 

(Reduction Percent) 

Farm 

Decontamination 

Cost (per hectare) 

Nonfarm 

Decontamination 

Cost (per 

person)

Decontamination 

Time (days) 

3 (67%) $1,200 $6,400 60

15 (93%) $2,700 $17,000 120
4-40

exposure pathway, (4) the dose conversion factor for a radionuclide and pathway, and (5) the 

usage factor for an exposure pathway. The total dose to an organ or the whole body used for 

modeling of health effects or protective action decisionmaking is then summed across all 

radionuclides and the relevant exposure pathways.

The detailed model formulation for each exposure pathway is discussed in Chapter 3 of 

NUREG/CR-4691 [2]. The exposure pathways considered during the emergency phase include 

cloudshine, groundshine, cloud inhalation, and inhalation of resuspended radionuclides. The 

exposure pathways for the intermediate phase include groundshine and inhalation of 

resuspended radionuclides. The long-term phase includes groundshine, inhalation of 

resuspended radionuclides, food ingestion, and water ingestion. For the early phase, two kinds of 

doses are calculated: (1) acute doses used for calculating early fatalities and injuries and (2) 

lifetime dose commitment used for calculating cancers resulting from the early exposure. For the 

long-term phase, only lifetime dose commitments are calculated. 

The quantities used in the dose equations depend on the exposure pathway and are either user 

inputs or are computed internally by MACCS. The radionuclide concentrations are calculated by 

the ATD code module at ground level along the plume centerline. In order to calculate the doses 

at different locations within a grid element, a correction factor (discussed in Section 3.1.1 and 

3.2.1 of NUREG/CR-4691, Volume 2 [2]) is derived to adjust the concentration to account for the 

reduction in concentration away from the plume centerline. The duration of exposure depends on 

the exposure pathway and the protective actions in a grid element, and is either calculated by 

MACCS or specified by the user. The shielding factor is a dimensionless quantity used to reduce 

the radiation dose as a result of shielding protection provided by a given protective action for a 

given exposure pathway. Shielding factors for the early phase were discussed in Section 4.2.6.3.

Shielding factors for the long-term phase use the early phase normal activity value for the entire 

population. The dose conversion factors for all exposure pathways are provided via the MACCS 

dose conversion factor file and have the following units for each exposure pathway:

• Cloudshine: Sv/s per Bq/m3

• Groundshine: Sv/s per Bq/m2

• Acute Inhalation: Sv/Bq

• Chronic Inhalation: Sv/Bq

• Ingestion: Sv/Bq

The dose conversion factors used in this analysis are identical to those used in SOARCA. They

are based on a methodology that considers the updated dosimetry and health effects models from 

FGR-13 [19], as well as the instantaneous dose rate values provided in the supplemental files 

provided with FGR-13. This allows both a consideration of the acute effects due to short-term 

exposure, as well as the ability to consider annual doses as well as committed doses. The dose 

conversion factor file set used in this analysis, “FGR13GyEquivDCF.INP,” together with its annual 

dose files, contains dose conversion factors based on FGR-13 for 825 radionuclides, 26 tissues, 

organs, the whole body effective dose, and four exposure pathways. MACCS contains a more 

limited set of organ dose quantities than are available in the DCF file based on FGR-13. MACCS 

considers nine organs (including whole body) for stochastic effects from chronic exposures and 

six organs for deterministic effects from acute exposures. Due to a current limitation of eight 

cancer sites (organs), MACCS calculates the dose to seven specific cancer sites and one residual 

cancer site. To estimate residual cancers, the dose coefficients for the pancreas are used as a 
4-41

surrogate for dose to soft tissue, following recommendations of the Oak Ridge National 

Laboratory letter report [58].

The dosimetry calculation also considers a usage factor for the inhalation and ingestion exposure 

pathways. For cloudshine and groundshine, the value in the calculation is one. For inhalation, the 

usage factor is a volumetric breathing rate and is specified by the user. Consistent with past 

studies, one value of this parameter is used for all populations and time periods: 2.66E-4 m3

/s.

This value was derived in NUREG/CR-4551, Vol. 2, Rev. 1, Part 7 [47] for an adult man who 

sleeps 8 hours per day and engages in light activity when awake. This value is slightly higher than 

one that considered men, women, and children. The usage factor for the ingestion pathway has 

two components, both of which are specified in the COMIDA2 food chain file that is an input to 

MACCS. The first describes the uptake of different radionuclides into different foodstuff types for 

different seasons of the year. The second component describes the consumption rate of each 

foodstuff type for an average adult.

The effective dose (ICRP60ED) is used internally by MACCS for simulating protective action 

decisions based on dosimetric quantities computed under a system of radiation protection. This 

requires a specification of the tissue weighting factors to compute an effective dose. The tissue 

weighting factors used in the computation of the dose conversion factor for the ICRP60ED 

effective dose are taken from ICRP-60 [60] and are identical to those used in SOARCA.

4.2.10 Health Effects

The approach used for health effects modeling in this analysis is identical to that used in the 

SECY-12-0157 offsite consequence analysis [16].

MACCS considers two types of health effects: deterministic health effects arising from acute 

exposures during the early phase of an accident, and stochastic health effects arising from acute 

exposures during the early phase of an accident and chronic exposures during the intermediate and 

long-term phases. The health effects models in MACCS are based on the models described in the 

NUREG/CR-4214 series of reports [57], as reflected in NUREG/CR-4691 [2] and NUREG/CR-6613 

[3]. The models presented in these reports provide estimates of the likelihood that an exposed 

individual may experience a specific health effect (e.g., lung impairment, breast cancer). Depending 

upon the exposure pathway, MACCS considers three types of populations: (1) individuals residing in 

the area surrounding the accident site who are directly exposed to contaminated media, (2) 

individuals who reside in unspecified locations that consume food grown in, or drink water 

originating in, the spatial elements surrounding the accident site, who are therefore indirectly 

exposed to contaminated media, and (3) decontamination workers who reside in unspecified 

locations. Health effects for all three populations are attributed to the grid element in which the 

contamination exists that created the exposure, even if the exposed population resides elsewhere. 

The distinction between the population types is used in the reporting of individual vs. collective 

health effects as shown in Table 4-21. After average individual risks have been estimated using 

the individual risk models, total cases of health effects are calculated in MACCS by multiplying the 

average individual risk of experiencing an effect by the number of people who receive the same 

dose that leads to the risk. However, health effect cases also account for the ingestion pathway, 

whereas individual risk does not. Furthermore, health effect cases also count doses to 

decontamination workers, whereas, individual risk does not. Measures of individual risk of health 

effects are based only on the direct pathways. Doses from these direct pathways are attributed to 

individual grid elements, and because the exposed population is confined to that grid element, 

estimates of individual risk can be made. Quantitative output for collective measures such as 
4-42

population dose includes all three types of populations. Although doses from indirect pathways 

(such as food or water ingestion or decontamination worker doses) are attributed to individual grid 

elements, they are not included in individual risk measures because the exposure population for 

these pathways may be very different from the population residing in that grid element.

Table 4-21 Exposure pathways and population types used to compute individual and 

collective health effects

The early health effect risk model implemented in MACCS is described in NUREG/CR-4691 [2] 

and in NUREG/CR-6613 [3]. This model is based on NUREG/CR-4214 [57] and has a sigmoidal 

dependence of individual risk on dose to the target organ in an exposed individual. 

The model considers a shape parameter that determines the steepness of the sigmoidal dose 

response curve, a dose equivalent delivered to the target organ, a dose equivalent threshold 

below which the effect is not expected to occur and thus the risk is zero, and a dose equivalent 

that would induce the effect in half the exposed population (D50). 

Parameters representing acute health effects are derived from expert elicitation data documented 

in NUREG/CR-6545 [37] and in NUREG/CR-7161 [36]. The most sensitive organ is typically the 

red bone marrow, which has a dose threshold of 2.32 Sv and a D50 of 5.6 Sv for causing fatal 

hematopoietic syndrome. Other fatal acute health effects include pulmonary syndrome from 

sufficiently large acute lung dose and gastrointestinal syndrome from sufficiently large acute 

stomach dose.

For stochastic health effects, NRC uses the linear-no-threshold (LNT) model for analyses serving 

regulatory purposes and this model is adopted in MACCS here. The LNT dose-response 

relationship suggests that any increase in dose, no matter how small, results in an incremental 

increase in risk. To support the SOARCA project, an Oak Ridge National Laboratory (ORNL) letter 

report “Radiation Dose and Health Risk Estimation: Technical Basis for the State-of-the-Art 

Reactor Consequence Analysis Project,” [58] was written to provide a basis for the risk factors 

which are taken from FGR-13 [19] and have their origin in the BEIR V report [59]. The cancer risk 

factors are used to convert doses to 8 different organs to the risk of an individual contracting 8 

Exposure Pathway and 

 Population Type 

Individual Health Effects 

(Early Fatality Risk and Latent 

Cancer Fatality Risk) 

Collective Health Effects

(Population Dose) 

Cloud Inhalation Early 

Cloudshine Early 

Groundshine to Population 

Residing on Spatial Grid Early, Intermediate, and Long-Term 

Groundshine to 

Decontamination Workers - Long-Term 

Inhalation of Resuspended 

Materials to Population 

Residing on Spatial Grid 

Early, Intermediate, and Long-Term 

Skin Dose Early 

Food and Water Ingestion - Long-Term 
4-43

different fatal cancers: leukemia, bone, breast, lung, thyroid, liver, colon, and residual. Residual 

cancers represent all types of cancer that are not explicitly treated.

The BEIR V report also specified the dose and dose rate effectiveness factor (DDREF), which is 

used to reduce the health impact of low doses and dose rates. The DDREF is applied to all doses 

in the intermediate and long-term phases and to those doses in the early phase that are less than 

a threshold value of 20 rem (0.2 Sv) to the target organ. This guidance for the application of the 

DDREF is identical to the recommendations provided in ICRP 60 [60]. The DDREF is given a 

value of 2 for central estimates of most cancer types with the exception of breast and thyroid 

cancers, for which the DDREF is assigned a value of 1.

4.3 Offsite Consequence Results

The summary results for the 18 Mark I source term bins and 9 Mark II source term bins are 

provided in Table 4-22 and Table 4-23, respectively so that they can be mapped to the release 

categories developed in the project’s accident sequence analysis. Individual early fatality risk is 

presented for the area within 1.3 miles of the site because this ring most closely approximates the 

area within 1 mile from the site boundary, the area for which NRC’s early fatality quantitative 

health objective (QHO) applies [61]. Individual latent cancer fatality risk is presented for the areas 

within 10, 50 and 100 mi from the site. The 10-mile area is presented because it corresponds to 

the QHO for cancer fatality risk [61] and to the plume exposure EPZ. Results are displayed for the 

50-mile area because that region corresponds to the ingestion exposure EPZ and is used in

NRC’s regulatory analyses. Results are also displayed for the 100-mile area for sensitivity

calculations in the estimates of relative public health risk reduction. Population dose, offsite cost,

land contamination, and population subject to long-term protective actions are provided for the 50-

mile and 100-mile areas; the 50-mile area is used for the base case regulatory analysis

calculations and the 100-mile area is used for sensitivity calculations in the regulatory analysis.

Results for the 100-mile area are also provided to serve as an indication of the extent/fraction of

offsite consequences captured within the 50-mile area.
4-44

Table 4-22 MACCS results for 18 Mark I source term bins

∗ Note: For the purpose of quantifying the time signature of a source term release, an hourly plume 

segment is considered “significant” if it contributes at least 0.5% of that source term’s total cumulative 

cesium release to the environment. Cesium, rather than iodine, was selected here because all of the 

resulting offsite consequences are driven by long-term phase exposures.

Individiual Early 

Fatality Risk

0-1.3 mi and beyond 0-10 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi

1 28DF1000 0.0006% 0.006% 14.9 7 0 4.65E-07 4.57E-08 2.06E-08 1,620 2,380

2 48DF100 0.002% 0.02% 11.4 8 0 1.90E-06 1.90E-07 8.69E-08 5,480 8,260

3 10DF100 0.01% 0.08% 16.3 6 0 6.25E-06 7.16E-07 3.21E-07 16,500 27,300

4 7DF1000 0.02% 0.26% 14.9 20 0 1.72E-05 2.35E-06 1.01E-06 48,400 77,600

5 11DF10 0.06% 0.78% 14.4 4 0 2.03E-05 3.36E-06 1.62E-06 71,200 127,000

6 48 0.23% 1.69% 11.4 8 0 7.95E-05 1.61E-05 7.79E-06 253,000 450,000

7 15 0.60% 5.85% 14.9 7 0 1.21E-04 3.28E-05 1.64E-05 524,000 932,000

8 46 0.98% 11.01% 14.8 17 0 1.53E-04 4.59E-05 2.34E-05 790,000 1,410,000

9 5DF10 1.05% 2.89% 24.2 34 0 3.55E-04 7.50E-05 3.35E-05 1,040,000 1,720,000

10 5 1.39% 6.46% 24.2 41 0 4.06E-04 9.78E-05 4.51E-05 1,360,000 2,290,000

11 8 1.49% 19.25% 14.9 5 0 1.35E-04 6.41E-05 3.43E-05 1,110,000 2,030,000

12 1 1.93% 22.68% 14.9 22 0 2.91E-04 1.01E-04 5.23E-05 1,720,000 3,090,000

13 41DF1000 3.40% 7.65% 9.8 17 0 5.22E-04 1.49E-04 7.89E-05 1,900,000 3,610,000

14 22dw 2.82% 18.64% 14.9 27 0 4.27E-04 1.28E-04 6.57E-05 1,830,000 3,320,000

15 53 2.79% 29.05% 17.4 13 0 2.59E-04 1.19E-04 6.96E-05 1,740,000 3,520,000

16 41 4.54% 14.10% 9.8 16 0 5.57E-04 1.75E-04 9.82E-05 2,300,000 4,520,000

17 3DF10 8.85% 24.65% 9.8 63 0 7.10E-04 2.95E-04 1.68E-04 3,830,000 7,720,000

18 52 15.90% 34.32% 17.4 11 0 5.39E-04 2.23E-04 1.50E-04 3,080,000 6,870,000

0-50 mi 0-100 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi

1 28DF1000 0.0006% 0.006% 14.9 7 78,900,000 78,900,000 0 0 - -

2 48DF100 0.002% 0.02% 11.4 8 79,700,000 79,700,000 1 1 0 0

3 10DF100 0.01% 0.08% 16.3 6 98,100,000 98,700,000 10 11 1 1

4 7DF1000 0.02% 0.26% 14.9 20 141,000,000 141,000,000 23 23 7 7

5 11DF10 0.06% 0.78% 14.4 4 220,000,000 240,000,000 41 65 118 118

6 48 0.23% 1.69% 11.4 8 1,150,000,000 1,390,000,000 116 175 3,440 3,440

7 15 0.60% 5.85% 14.9 7 2,740,000,000 3,690,000,000 190 361 15,000 16,600

8 46 0.98% 11.01% 14.8 17 3,760,000,000 5,220,000,000 242 506 20,700 27,400

9 5DF10 1.05% 2.89% 24.2 34 7,290,000,000 8,600,000,000 351 429 35,200 35,200

10 5 1.39% 6.46% 24.2 41 9,900,000,000 12,000,000,000 479 715 51,400 51,500

11 8 1.49% 19.25% 14.9 5 5,960,000,000 9,720,000,000 286 673 40,500 55,800

12 1 1.93% 22.68% 14.9 22 13,000,000,000 17,400,000,000 549 1,040 64,500 79,700

13 41DF1000 3.40% 7.65% 9.8 17 19,400,000,000 24,700,000,000 783 1,170 168,000 190,000

14 22dw 2.82% 18.64% 14.9 27 12,900,000,000 18,300,000,000 544 1,010 93,700 114,000

15 53 2.79% 29.05% 17.4 13 15,700,000,000 26,500,000,000 573 1,290 111,000 142,000

16 41 4.54% 14.10% 9.8 16 25,500,000,000 35,400,000,000 904 1,500 235,000 281,000

17 3DF10 8.85% 24.65% 9.8 63 47,000,000,000 68,100,000,000 1,360 2,470 417,000 504,000

18 52 15.90% 34.32% 17.4 11 46,500,000,000 87,700,000,000 987 2,170 467,000 873,000

Rep Case 

I (%)

Start 

Time 

(hrs)

Land (sq mi) Exceeding 

Long-Term Habitability 

Criterion

Population Subject to 

Long-Term Protective 

Actions

Individual Latent Cancer Fatality Risk Population Dose

 (person-rem)

Offsite Cost ($ 2013)

# Hrs with 

Significant 

Cs 

Release*

Bin Rep Case Rep Case 

Cs (%)

Rep Case 

I (%)

Start 

Time 

(hrs)

# Hrs with 

Significant 

Cs 

Release*

Bin Rep Case Rep Case 

Cs (%)
4-45

Table 4-23 MACCS results for 9 Mark II source term bins

∗ Note: For the purpose of quantifying the time signature of a source term release, an hourly plume 

segment is considered “significant” if it contributes at least 0.5% of that source term’s total cumulative 

cesium release to the environment. Cesium, rather than iodine, was selected here because all of the 

resulting offsite consequences are driven by long-term phase exposures.

4.3.1 Individual Early Fatality Risk

Individual early fatality risk was calculated to be zero for all Mark I and Mark II source term bin 

cases within 1.3 miles of the site and beyond as shown in Table 4-22 and Table 4-23. This is 

because the source terms are not large enough to exceed the threshold for the acute dose to the 

red bone marrow, which is typically the most sensitive tissue for early fatalities. As described 

previously in Section 3.10, the dose threshold for the red bone marrow is 2.32 Sv (232 rem).

Among the 18 Mark I and 9 Mark II calculations, the largest peak dose to the red bone marrow, 

averaged over all weather trials, is about 35 rem, well below the dose threshold, for the closest 

populated ring between 0.33 and 0.75 miles of the site. The computed results depend on 

meteorological conditions and population distribution, which are clearly site-specific. Thus, had 

other sites been selected for offsite consequence modeling than Peach Bottom and Limerick, 

calculated individual early fatality risk might have been nonzero, however recent NRC 

consequence studies such as SOARCA [9] and SOARCA uncertainty analyes [14] suggest they 

could still be characterized as “essentially zero.”

4.3.2 Individual Latent Cancer Fatality Risk

Individual latent cancer fatality (LCF) risk is defined as the risk of an average individual in a given 

spatial grid element contracting a fatal cancer from the radiation exposure pathways that may be 

applied to the residents of the area. These include early phase cloudshine, groundshine, 

inhalation, and skin deposition; intermediate phase groundshine and inhalation of resuspended 

materials; and long-term phase groundshine and inhalation of resuspended materials. Individual 

LCF risk is a population-weighted metric and is calculated by dividing the expected number of 

Individual Early 

Fatality Risk

0-1.3 mi and beyond 0-10 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi

1 11DF1000 0.00004% 0.0005% 20.3 20 0 9.72E-08 1.03E-08 3.45E-09 282 345

2 5DF1000 0.0006% 0.005% 32.2 20 0 1.15E-06 1.81E-07 6.35E-08 4,340 5,440

3 42DF100 0.0043% 0.037% 14.3 13 0 6.58E-06 8.67E-07 3.02E-07 20,700 26,700

4 11 0.042% 0.45% 20.3 20 0 7.90E-05 9.68E-06 3.27E-06 202,000 261,000

5 51DF10 0.23% 2.01% 16.6 9 0 1.35E-04 3.39E-05 1.21E-05 689,000 888,000

6 5 0.55% 4.94% 32.2 20 0 2.29E-04 1.05E-04 4.01E-05 2,160,000 2,900,000

7 3 1.09% 10.26% 14.3 20 0 3.08E-04 1.88E-04 7.43E-05 4,140,000 5,580,000

8 1 2.46% 19.81% 22.8 25 0 4.70E-04 3.17E-04 1.25E-04 6,110,000 8,260,000

9 52 3.57% 28.67% 16.6 10 0 4.03E-04 2.46E-04 1.01E-04 5,430,000 7,440,000

0-50 mi 0-100 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi

1 11DF1000 0.00004% 0.0005% 20.3 20 381,000,000 381,000,000 - - - -

2 5DF1000 0.0006% 0.005% 32.2 20 381,000,000 381,000,000 0 0 - -

3 42DF100 0.0043% 0.037% 14.3 13 393,000,000 393,000,000 2 2 0 0

4 11 0.042% 0.45% 20.3 20 844,000,000 846,000,000 44 47 1,030 1,030

5 51DF10 0.23% 2.01% 16.6 9 4,250,000,000 4,380,000,000 130 221 15,400 15,400

6 5 0.55% 4.94% 32.2 20 24,000,000,000 28,000,000,000 303 551 62,400 62,400

7 3 1.09% 10.26% 14.3 20 80,800,000,000 105,400,000,000 698 1,200 619,000 649,000

8 1 2.46% 19.81% 22.8 25 85,500,000,000 109,300,000,000 854 1,680 721,000 741,000

9 52 3.57% 28.67% 16.6 10 53,600,000,000 63,800,000,000 618 1,400 414,000 449,000

Offsite Cost ($ 2013)

Population Dose

 (person-rem)

Land (sq mi) Exceeding 

Long-Term Habitability 

Criterion

Bin Rep Case Rep Case 

Cs (%)

Rep Case 

I (%)

Start 

Time 

(hrs)

Individual Latent Cancer Fatality Risk # Hrs with 

Significant 

Cs 

Release*

Population Subject to LongBin Rep Case Term Protective Actions Rep Case 

Cs (%)

Rep Case 

I (%)

Start 

Time 

(hrs)

# Hrs with 

Significant 

Cs 

Release*
4-46

fatal latent cancers in a spatial grid element by the population residing in that grid element. The 

expected number of fatal latent cancers considers the doses to each population in a grid element 

for the relevant pathways and multiplies them by the cancer risk factors specified for the 8 fatal 

cancer types considered. The risk to each organ is then added together and the risk from each 

phase is added together to arrive at an expected number of latent cancer deaths.

In general, individual LCF risk increases with source term magnitude for the 18 Mark I and 9 Mark 

II source term bins. This is shown in Figure 4-4 for the Mark I source terms and Figure 4-5 for the 

Mark II source terms. Larger source terms generally result in higher radionuclide concentrations 

across the spatial grid resulting in higher expected doses and therefore higher numbers of health 

effects. However, there are certain cases when a slightly larger source term can result in a lower 

individual LCF risk. For example, Mark I bin 10 has a slightly smaller cesium release (1.39%) than 

Mark I bin 11 (1.49%), yet bin 10 leads to a higher individual LCF risk for all three areas 

considered. (Mark I bin 10 causes a conditional (per event) individual LCF risk of 4E-04 within 10 

miles whereas Mark I bin 11 causes a conditional individual LCF risk of ~1E-04 for the same 

area.) This is likely explained by the different time signature of each source term. Mark I bin 10 

has a more gradual release; whereas, Mark I bin 11 is more of a pulse-type release. One way to 

quantify the time signature of a release is to identify the number of hours in which a significant 

portion of the source term is released. Because the different source term bins span many orders 

of magnitude, the determination of significant is based on a percent of the individual source term’s 

total release, rather than an activity level. A threshold of 0.5% would consider any hourly plume 

segment to be significant if it releases half a percent of that individual source term’s total cesium.

For example, if the total cumulative environmental release of a source term was 2% of the core 

inventory of cesium, an hourly plume segment would be considered in this quantification if it 

released at least 0.01% of the total core inventory. Using this threshold, one can quantify the 

difference in release profiles. Mark I bin 10 contains 41 hours in which at least 0.5% of the source 

term’s total Cs is released, while Mark I bin 11 contains just 5 hours in which at least 0.5% of the 

source term’s total Cs is released. Mark I bin 10 causes higher individual LCF risks because the 

longer release duration allows more time for the wind to change direction, resulting in 

radionuclides being transported across a larger portion of the grid. When radionuclides are spread 

in more directions across the grid, there is a greater chance that the plume will intersect with 

population centers leading to a higher individual LCF risk. The larger consequences of a longer, 

more gradual release compared with a shorter, more punctuated release are more pronounced at 

shorter distances and less apparent at longer distances. This trend results from the fact that the 

exposures from narrower, more concentrated plume patterns provide significant exposures at 

longer distances; whereas, less concentrated plumes that are spread over multiple compass 

sectors become depleted at shorter distances. 

Individual LCF risk generally decreases when larger areas are considered because contamination 

levels generally decrease with distance. Thus the 10-mile area risk is the largest for each source 

term followed by the 50-mile area and then the 100-mile area has the smallest average individual 

risk for each source term. Smaller source terms generally result in a larger difference in individual 

LCF risk between small (10-mi) and larger (50-, 100-mi) areas because smaller source terms 

generally lead to smaller exposures and therefore the expected number of fatal latent cancers 

drops faster as a function of distance from the site. For example, Mark II source term bin 1 has a 

10-mile individual LCF risk about 28 times higher than the 100-mile value; whereas, Mark II

source term bin 9 has a 10-mile individual LCF risk just 4 times higher than the 100-mile value.
4-47

Figure 4-4 Mark I source terms – individual latent cancer fatality risk per event

Figure 4-5 Mark II source terms – individual latent cancer fatality risk per event

1.0E-08

1.0E-07

1.0E-06

1.0E-05

1.0E-04

1.0E-03

1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18

Mark I Source Term Bin

0-10 miles

0-50 miles

0-100 miles

1.0E-09

1.0E-08

1.0E-07

1.0E-06

1.0E-05

1.0E-04

1.0E-03

1 2 3 4 5 6 7 8 9

Mark II Source Term Bin

0-10 miles

0-50 miles

0-100 miles


)LJXUHDQG)LJXUHVKRZWKHIUDFWLRQRIWKHDYHUDJHLQGLYLGXDO/&)ULVNWKDWUHVXOWVIURPWKH

HPHUJHQF\SKDVHDQGIURPWKHFRPELQDWLRQRIWKHLQWHUPHGLDWHDQGORQJWHUPSKDVHVIRUWKH

0DUN,DQG0DUN,,VRXUFHWHUPELQVUHVSHFWLYHO\IRUWKHDQGPLOHDUHDV7KHVHILJXUHV

VKRZWKDWWKHLQWHUPHGLDWHDQGORQJWHUPSKDVHVGRPLQDWHFRPSDUHGWRWKHHPHUJHQF\SKDVH

$OVRWKH\VKRZWKDWWKHHPHUJHQF\SKDVHKDVDVLJQLILFDQWO\VPDOOHUFRQWULEXWLRQWRLQGLYLGXDO

/&)ULVNIRUWKHPLOHDUHDFRPSDUHGWRWKHPLOHDUHDIRUDOO0DUN,DQG0DUN,,VRXUFHWHUPV

7KHHPHUJHQF\SKDVHFRQWULEXWHVRQDYHUDJHVLPSOHDYHUDJHQRWIUHTXHQF\ZHLJKWHGIRU

WKH0DUN,VRXUFHWHUPVDQGRQDYHUDJHIRUWKH0DUN,,VRXUFHWHUPVIRUWKHPLOHDUHD

ZKHUHDVWKHHPHUJHQF\SKDVHFRQWULEXWHVDQGRQDYHUDJHIRUWKH0DUN,DQG0DUN,,

VRXUFHWHUPELQVUHVSHFWLYHO\IRUWKHPLOHDUHD7KLVGLIIHUHQFHLVDUHVXOWRIWKHHQKDQFHG

SURWHFWLYHDFWLRQVIRUWKHPLOHDUHDUHODWLYHWRDUHDVEH\RQGPLOHVWKDWDUHH[SHFWHGLQD

UHDODFFLGHQWDQGPRGHOHGLQ0$&&67KH(3=SRSXODWLRQH[FOXGLQJDYHU\VPDOOQRQHYDFXDWLQJ

FRKRUWLVPRGHOHGWRUHFHLYHHDUO\QRWLILFDWLRQVKHOWHUDQGHYDFXDWHDQGKDOIRIWKH(3=

SRSXODWLRQLVPRGHOHGWRFRQVXPH.,

2EVHUYDWLRQRI)LJXUHDQG)LJXUHDOVRVKRZVWKDWWKHUHODWLYHFRQWULEXWLRQVRIWKHGLIIHUHQW

DFFLGHQWSKDVHVWRLQGLYLGXDO/&)ULVNYDU\ZLWKWKHVRXUFHWHUP7KLVLVODUJHO\DIXQFWLRQRIWKH

UHODWLYHFHVLXPDQGLRGLQHVRXUFHWHUPVL]HVIRUHDFKELQ,RGLQHKDVJUHDWHUSRWHQWLDOIRUHDUO\

KHDOWKHIIHFWVZKHUHDVFHVLXPKDVJUHDWHUSRWHQWLDOIRUORQJWHUPKHDOWKHIIHFWV6RXUFHWHUPV

ZLWKDUHODWLYHO\KLJKLRGLQHUHOHDVHOHDGWRUHODWLYHO\ODUJHUFRQWULEXWLRQVIURPWKHHPHUJHQF\

SKDVH)RUH[DPSOH0DUN,ELQDQGELQKDYHDQLRGLQHUHOHDVHDERXWWLPHVODUJHULQ

WHUPVRISHUFHQWRIFRUHLQYHQWRU\WKDQWKHFHVLXPUHOHDVHDQGWKHVHOHDGWRWKHKLJKHVW

FRQWULEXWLRQIURPWKHHPHUJHQF\SKDVHIRUWKHPLOHDUHDIRUELQIRUELQ,Q

FRQWUDVW0DUN,ELQKDVDQLRGLQHUHOHDVHIUDFWLRQMXVWDERXWWKUHHWLPHVKLJKHUWKDQLWVFHVLXP

UHOHDVHIUDFWLRQDQGWKLVOHDGVWRWKHVPDOOHVWFRQWULEXWLRQIURPWKHHPHUJHQF\SKDVHIRUWKH

PLOHDUHDFRPSDUHGWRIURPWKHLQWHUPHGLDWHDQGORQJWHUPSKDVHV

)LJXUH 0DUN,VRXUFHWHUPV±LQGLYLGXDOODWHQWFDQFHUIDWDOLW\ULVNSHUHYHQWZLWK

EUHDNGRZQE\DFFLGHQWSKDVH

0.0E+00

1.0E-04

2.0E-04

3.0E-04

4.0E-04

5.0E-04

6.0E-04

7.0E-04

8.0E-04

0-10 mi0-50 mi0-10 mi0-50 mi0-10 mi0-50 mi0-10 mi0-50 mi0-10 mi0-50 mi0-10 mi0-50 mi0-10 mi0-50 mi0-10 mi0-50 mi0-10 mi0-50 mi0-10 mi0-50 mi0-10 mi0-50 mi0-10 mi0-50 mi0-10 mi0-50 mi0-10 mi0-50 mi0-10 mi0-50 mi0-10 mi0-50 mi0-10 mi0-50 mi0-10 mi0-50 mi

1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18

Mark I Source Term Bin

Intermediate and Long-Term Phase

Emergency Phase

mi


)LJXUH 0DUN,,VRXUFHWHUPV±LQGLYLGXDOODWHQWFDQFHUIDWDOLW\ULVNSHUHYHQWZLWK

EUHDNGRZQE\DFFLGHQWSKDVH

3RSXODWLRQ'RVH

3RSXODWLRQGRVHSURYLGHVDZD\WRFKDUDFWHUL]HWKHVRFLHWDOFRQVHTXHQFHVRIDQDFFLGHQW7KLV

PHWULFVXPVWKHGRVHVIURPDOOH[SRVXUHSDWKZD\VWKRVHXVHGLQWKHODWHQWFDQFHUIDWDOLW\ULVN

FDOFXODWLRQDVZHOODVIRRGDQGZDWHULQJHVWLRQDQGJURXQGVKLQHWRGHFRQWDPLQDWLRQZRUNHUVDQG

PXOWLSOLHVWKHPE\WKHVL]HRIWKHSRSXODWLRQWKDWZRXOGEHH[SHFWHGWRUHFHLYHWKHPIRUWKH

FDOFXODWHGWLPHGXUDWLRQ7KHSRSXODWLRQGRVHPHDVXUHGLQSHUVRQUHPLVVKRZQLQ)LJXUH

DQG)LJXUHIRUWKH0DUN,DQG0DUN,,VRXUFHWHUPVUHVSHFWLYHO\(DFKILJXUHVKRZVERWKWKH

PLOHDQGPLOHDUHD$VZLWKLQGLYLGXDO/&)ULVNWKHSRSXODWLRQGRVHJHQHUDOO\LQFUHDVHV

ZLWKVRXUFHWHUPPDJQLWXGH$VDVHFRQGDU\FRQVLGHUDWLRQWKHSRSXODWLRQGRVHDOVRJHQHUDOO\

LQFUHDVHVIRUVRXUFHWHUPVWKDWVSDQDORQJHUGXUDWLRQ)RUH[DPSOH0DUN,ELQKDVDFHVLXP

UHOHDVHDOPRVWGRXEOH0DUN,ELQKRZHYHU0DUN,ELQUHVXOWVLQDORZHUZHDWKHUDYHUDJHG

SRSXODWLRQGRVHaSHUVRQUHPWKDQ0DUN,ELQaSHUVRQUHP0DUN,ELQ

OLNHO\KDVDKLJKHUSRSXODWLRQGRVHEHFDXVHLWKDVDPXFKORQJHUUHOHDVHGXUDWLRQZLWKPRUH

KRXUVRIVLJQLILFDQWFHVLXPUHOHDVH0DUN,ELQKDVKRXUSHULRGVLQZKLFKDWOHDVWRI

WKHWRWDOVRXUFHWHUPLVEHLQJUHOHDVHGZKHUHDV0DUN,ELQKDVMXVWKRXUSHULRGVLQZKLFK

DWOHDVWRIWKHWRWDOVRXUFHWHUPLVEHLQJUHOHDVHG$ORQJHUUHOHDVHGXUDWLRQLQFUHDVHVWKH

SUREDELOLW\WKDWWKHSOXPHVLQWHUVHFWZLWKSRSXODWLRQFHQWHUVZKLFKGULYHVXSWKHSRSXODWLRQGRVH

0.0E+00

5.0E-05

1.0E-04

1.5E-04

2.0E-04

2.5E-04

3.0E-04

3.5E-04

4.0E-04

4.5E-04

5.0E-04

0-10

mi

0-50

mi

0-10

mi

0-50

mi

0-10

mi

0-50

mi

0-10

mi

0-50

mi

0-10

mi

0-50

mi

0-10

mi

0-50

mi

0-10

mi

0-50

mi

0-10

mi

0-50

mi

0-10

mi

0-50

mi

123456789

Mark II Source Term Bin
4-50

Figure 4-8 Mark I source terms – population dose per event (person-rem)

Figure 4-9 Mark II source terms – population dose per event (person-rem)

1.0E+02

1.0E+03

1.0E+04

1.0E+05

1.0E+06

1.0E+07

1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18

Population Dose (person-rem)

Mark I Source Term Bin

0-50 miles

0-100 miles

1.0E+02

1.0E+03

1.0E+04

1.0E+05

1.0E+06

1.0E+07

1 2 3 7 8 9

Population Dose (person-rem)

4 5 6

Mark II Source Term Bin

0-50 miles

0-100 miles
4-51

Figure 4-10 and Figure 4-11 show the population dose fraction that results from the different 

accident phases and exposure pathways for the area within 50 miles of each site. Each figure 

includes an inset chart to rescale the population dose for the smaller source term bins so they are 

legible. The fraction from the emergency phase ranges from about 15% to 60% of the total 

population dose. This fraction depends on the size of the source term’s iodine release relative to 

its cesium release. Source terms with relatively high iodine releases (Mark I bin 11 and Mark I bin 

12) have a larger fraction of the population dose from the emergency phase. For the Mark II

source terms, the emergency phase fraction generally increases with source term magnitude. The

ratio of iodine release to cesium release is fairly consistent, around an order of magnitude, so the

trend is more consistent than for the Mark I source term bins.

The groundshine exposure pathway, considering both the intermediate and long-term phases, is 

also a major contributor to population dose, ranging from about 20% to 70%. This pathway’s 

contribution generally increases with the magnitude of the cesium release because this creates 

more offsite contamination for a longer period of time. The groundshine exposure pathway doses 

are accumulated for the population that live on land that has been contaminated but at doses that 

are under the intermediate phase relocation criterion and the long-term phase habitability criterion.

The long-term phase food and water ingestion exposure pathway is a significant contributor to 

population dose for the smaller source terms, but this pathway’s relative contribution (fraction of 

dose) drops as the source term increases. This is because larger source terms create 

progressively more groundshine exposure dose so the relative contribution from ingestion 

decreases as a result. 

Intermediate and long-term phase resuspension dose is a very small contributor to population 

dose for all source terms because this pathway is less effective in causing doses to humans than 

others, like groundshine. 

The long-term phase decontamination exposure doses, both to farmland and populated land, are 

also very small contributors to population dose. These increase for larger cesium source terms 

because more cesium release contaminates more land and thus requires more decontamination.

Research efforts are underway to evaluate newly emerging information from the Fukushima 

accident recovery experience, and in particular develop MACCS decontamination plan input 

parameters based on Fukushima. The decontamination plan input parameters include the costs to 

decontaminate, the dose reductions achieved, and the times required to perform decontamination.

These research efforts were not completed in time for this analysis, but preliminary information, 

had it been used in this project, may have shown a higher contribution to total population dose 

from the decontamination exposure pathway. Additional information from the Fukushima accident 

related to total offsite accident cost is provided in the following section. 


)LJXUH 0DUN,VRXUFHWHUPV±SRSXODWLRQGRVHSHUHYHQWSHUVRQUHPZLWKEUHDNGRZQ

E\DFFLGHQWSKDVHDQGH[SRVXUHSDWKZD\PLOHV

)LJXUH 0DUN,,VRXUFHWHUPV±SRSXODWLRQGRVHSHUHYHQWSHUVRQUHPZLWK

EUHDNGRZQE\DFFLGHQWSKDVHDQGH[SRVXUHSDWKZD\PLOHV

0.0E+00

5.0E+05

1.0E+06

1.5E+06

2.0E+06

2.5E+06

3.0E+06

3.5E+06

4.0E+06

1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18

Mark I Source Term Bin

Emergency Phase

Intermediate and LongTerm Phase Groundshine

Long-Term Phase Ingestion

Intermediate and LongTerm Phase Resuspension

Population Dependent

Decontamination

Farm Dependent

Decontamination

0.0E+00

1.0E+06

2.0E+06

3.0E+06

4.0E+06

5.0E+06

6.0E+06

7.0E+06

1 2 3 4 5 6 7 8 9

Mark II Source Term Bin

Emergency Phase

Intermediate and LongTerm Phase Groundshine

Ingestion

Intermediate and LongTerm Phase Resuspension

Population Dependent

Decontamination

Farm Dependent

Decontamination
4-53

4.3.4 Offsite Economic Cost

As with total population dose, offsite economic cost is a way to characterize the societal 

consequences of an accident. This metric sums the costs of the protective actions that need to be 

taken to reduce offsite exposures to avoid and minimize health effects and to restore land to 

usability and habitability. These costs include evacuation and relocation costs, moving expenses 

for people displaced, decontamination costs, costs due to loss of use of property, disposal of 

contaminated food, and the costs of condemning land. These cost categories are described in 

more detail in Section 2.3.3.

The total offsite cost, measured in 2013 dollars, is shown in Figure 4-12 and Figure 4-13 for the 

Mark I and Mark II source terms, respectively. Each figure shows both the 50-mile and 100-mile 

area. Like population dose, the total offsite cost generally increases with source term magnitude.

As a secondary consideration, the offsite cost also generally increases for source terms that span 

a longer duration. For example, Mark II bin 9 has a larger cesium release than Mark II bin 8 but 

results in a lower offsite cost ($ ~54B) than Mark II bin 8 ($ ~86B). Mark II bin 8 has a higher 

offsite cost because it has a significantly longer release duration. Mark II bin 8 has 25 hours in 

which a “significant” portion of the source term is released (for the purpose of quantifying the time 

signature of a source term release, an hourly plume segment is considered “significant” if it 

contributes at least 0.5% of that source term’s total cumulative cesium release to the 

environment); whereas, Mark II bin 9 has just 10 hours. A longer release duration increases the 

probability that the plumes intersect with population centers and drives up the economic cost. The 

larger consequences of a longer, more gradual release compared with a shorter, more punctuated 

release are more pronounced at shorter distances and less apparent at longer distances. This 

trend results from the fact that the exposures from narrower, more concentrated plume patterns 

provide significant exposures at longer distances; whereas, less concentrated plumes that are 

spread over multiple compass sectors become depleted at shorter distances.

For the smallest Mark I and Mark II source terms, the radionuclide release is too small to 

necessitate intermediate phase and long-term phase protective actions. Therefore the total offsite 

cost is essentially just the fixed cost of evacuating the EPZ population for the duration of the 

emergency phase, 1 week. For example, Mark I bin 1 has an offsite cost of ~ $79M. This cost 

represents evacuation of the EPZ population and is equal to the population times the per diem 

cost times the number of days (7). The total offsite cost for the smallest Mark II source term, bin 1, 

is $381M. This value is larger than for Mark I bin 1 solely as a result of the Mark II reference 

MACCS model having a higher population density.

Figure 4-12 and Figure 4-13 also show that for the smaller source terms, the total offsite cost is 

attributed essentially entirely to the area within 50 miles of the site. Using the total offsite cost values 

from Table 4-22 and Table 4-23, rounding to one significant figure shows that a cesium release of at 

least about 0.5% is needed to see any difference in the 50-mile and 100-mile area values.
4-54

Figure 4-12 Mark I source terms – offsite economic cost ($ 2013)

Figure 4-13 Mark II source terms – offsite economic cost ($ 2013)

Figure 4-14 and Figure 4-15 show the total offsite cost fraction that results from the different cost 

components for the area within 50 miles of each site. For the smaller source terms, the 

emergency phase cost dominates the total offsite cost because offsite contamination is minimal 

and therefore long-term phase protective actions are not warranted. As the source terms increase 

in magnitude, the intermediate phase and population-dependent interdiction costs represent an 

increasingly dominant share of the total offsite cost. As the source term magnitudes increase 

further, the farm-based costs start contributing to the total offsite costs (Mark I bins 3-5 and Mark II 

1.0E+07

1.0E+08

1.0E+09

1.0E+10

1.0E+11

1.0E+12

1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18

Offsite Economic Cost ($ 2013)

Mark I Source Term Bin

0-50 miles

0-100 miles

1.0E+07

1.0E+08

1.0E+09

1.0E+10

1.0E+11

1.0E+12

1 2 3 4 5 6 7 8 9

Offsite Economic Cost ($ 2013)

Mark II Source Term Bin

0-50 miles

0-100 miles


ELQ)DUPODQGFRVWVLQFOXGHWKHYDOXHRIPLONDQGFURSVWKDWFDQQRWEHFRQVXPHGDVZHOODV

IDUPODQGLQWHUGLFWLRQ+RZHYHUWKHIDUPEDVHGFRVWVTXLFNO\GHFUHDVHLQVKDUHRIWKHWRWDOFRVW

EHFDXVHWKHSRSXODWLRQEDVHGFRVWVJURZPXFKIDVWHUIRUODUJHUVRXUFHWHUPV

7KHFRQWULEXWLRQRISRSXODWLRQGHSHQGHQWGHFRQWDPLQDWLRQFRVWVLQFUHDVHVZLWKVRXUFHWHUP

PDJQLWXGHEXWGRHVQ¶WDFFRXQWIRUPRUHWKDQRIWKHWRWDOPLOHRIIVLWHFRVWHYHQIRUWKH

ODUJHVWVRXUFHWHUPV$VGHVFULEHGLQWKHSUHYLRXVVHFWLRQUHVHDUFKHIIRUWVDUHXQGHUZD\WR

HYDOXDWHLQIRUPDWLRQIURPWKH)XNXVKLPDDFFLGHQWUHFRYHU\H[SHULHQFHDQGLQSDUWLFXODUGHYHORS

0$&&6GHFRQWDPLQDWLRQSODQLQSXWSDUDPHWHUVEDVHGRQ)XNXVKLPD7KHGHFRQWDPLQDWLRQSODQ

LQSXWSDUDPHWHUVLQFOXGHWKHFRVWVWRGHFRQWDPLQDWHWKHGRVHUHGXFWLRQVDFKLHYHGDQGWKHWLPHV

UHTXLUHGWRSHUIRUPGHFRQWDPLQDWLRQ7KHVHUHVHDUFKHIIRUWVZHUHQRWFRPSOHWHGLQWLPHIRUWKLV

DQDO\VLVEXWSUHOLPLQDU\LQIRUPDWLRQKDGLWEHHQXVHGLQWKLVSURMHFWPD\KDYHVKRZQKLJKHU

GHFRQWDPLQDWLRQFRVWVDQGDKLJKHUFRQWULEXWLRQWRWRWDORIIVLWHFRVWIURPWKHGHFRQWDPLQDWLRQFRVW

FRPSRQHQW

)LJXUH 0DUN,VRXUFHWHUPV±RIIVLWHHFRQRPLFFRVWZLWKEUHDNGRZQE\

DFFLGHQWSKDVHDQGSURWHFWLYHDFWLRQW\SHPLOHV

0.0E+00

5.0E+09

1.0E+10

1.5E+10

2.0E+10

2.5E+10

3.0E+10

3.5E+10

4.0E+10

4.5E+10

5.0E+10

1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18

Mark I Source Term Bin

Population Dependent

Decontamination

Farm Dependent

Decontamination

Population Dependent

Interdiction

Farm Dependent Interdiction

Population Dependent

Condemnation

Farm Dependent Condemnation

Emergency Phase

Intermediate Phase

Milk Disposal

Crop Disposal


)LJXUH 0DUN,,VRXUFHWHUPV±RIIVLWHHFRQRPLFFRVWZLWKEUHDNGRZQE\

DFFLGHQWSKDVHDQGSURWHFWLYHDFWLRQW\SHPLOHV

/DQG&RQWDPLQDWLRQDQG3RSXODWLRQ6XEMHFWWR/RQJ7HUP3URWHFWLYH$FWLRQV

%H\RQGWRWDOSRSXODWLRQGRVHDQGWRWDORIIVLWHHFRQRPLFFRVWH[WHQWRIODQGFRQWDPLQDWLRQDQG

SRSXODWLRQVXEMHFWWRORQJWHUPSURWHFWLYHDFWLRQVUHSUHVHQWWZRDGGLWLRQDOPHDVXUHVRIWKH

VRFLHWDOFRVWRIDQXFOHDUDFFLGHQW7KHVHKDYHWKHDGYDQWDJHWKDWWKH\DUHPHDVXUHGLQWHUPV

DUHDQXPEHURISHRSOHWKDWPD\EHPRUHPHDQLQJIXOWRVRPHVWDNHKROGHUVFRPSDUHGWR

SHUVRQUHPRUGROODUV

/DQGFRQWDPLQDWLRQLVPHDVXUHGDVWKHDUHDRIODQGWKDWH[FHHGVWKHORQJWHUPSKDVHKDELWDELOLW\

FULWHULRQZKLFKLVPRGHOHGKHUHDVPUHPSHU\HDUVWDUWLQJLQWKHILUVW\HDURIWKHORQJWHUP

SKDVH7KLVLVWKHODQGWKDWLVHLWKHUWHPSRUDULO\LQWHUGLFWHGZLWKRUZLWKRXWGHFRQWDPLQDWLRQRULV

FRQGHPQHGDQGFRQVLGHUVERWKIDUPODQGDQGSRSXODWHGODQG7KHORQJWHUPSKDVHEHJLQVDIWHUWKH

HQGRIWKHWKUHHPRQWKLQWHUPHGLDWHSKDVHZKLFKEHJLQVDIWHUWKHHQGRIWKHZHHNORQJHPHUJHQF\

SKDVH7KLVPHWULFGRHVQRWFRQVLGHUWKHGXUDWLRQRIWLPHIRUZKLFKWKHODQGH[FHHGVWKHKDELWDELOLW\

FULWHULRQDQGLVFRQVLGHUHGWRRFRQWDPLQDWHGWREHKDELWDEOH

)LJXUHDQG)LJXUHVKRZWKHDUHDVTXDUHPLOHVRIODQGH[FHHGLQJWKHORQJWHUPSKDVH

KDELWDELOLW\FULWHULRQIRUWKHPLOHDQGPLOHDUHDVDURXQGHDFKVLWHIRUWKH0DUN,DQG0DUN,,

VRXUFHWHUPV7KHDUHDRIODQGFRQWDPLQDWHGLQFUHDVHVZLWKVRXUFHWHUPPDJQLWXGHDQGZLWKUHOHDVH

GXUDWLRQEHFDXVHDORQJHUUHOHDVHDOORZVPRUHWLPHIRUWKHZLQGWRFKDQJHGLUHFWLRQDQGWKXV

LQFUHDVHVWKHSUREDELOLW\RIVSUHDGLQJFRQWDPLQDWLRQRYHUDODUJHUDUHD7KHZHDWKHUDYHUDJHGDUHDRI

ODQGFRQWDPLQDWLRQUDQJHVIURPHVVHQWLDOO\]HURIRUWKHVPDOOHVW0DUN,,VRXUFHWHUPELQVWRDERXW

0.0E+00

1.0E+10

2.0E+10

3.0E+10

4.0E+10

5.0E+10

6.0E+10

7.0E+10

8.0E+10

9.0E+10

1 2 3 4 5 6 7 8 9

Mark II Source Term Bin

Population Dependent

Decontamination

Farm Dependent

Decontamination

Population Dependent

Interdiction

Farm Dependent Interdiction

Population Dependent

Condemnation

Farm Dependent Condemnation

Emergency Phase

Intermediate Phase

Milk Disposal

Crop Disposal
4-57

1,400 square miles for Mark I bin 17 within 50 miles of the site. This maximum represents about 18% 

of the land within 50 miles.

Examination of the blue bars representing the 50-mile area and the orange bars representing the 100-

mile area shows that the contaminated land is concentrated in the inner 50-mile circle for the small 

source terms. However for the largest source terms, more than half of the contaminated land is 

beyond 50 miles from the site. For example, 2,170 square miles are contaminated within 100 miles for 

Mark I bin 18; 987 square miles (~45%) of which are within 50 miles, and 1,183 square miles (~55%) 

of which are between 50 and 100 miles.

Figure 4-16 Mark I source terms – land exceeding long-term phase habitability criterion 

(square miles) per event

1.0E-01

1.0E+00

1.0E+01

1.0E+02

1.0E+03

1.0E+04

1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18

Land (sq. mi) Exceeding Long-Term 

Phase Habitability Criterion Per Event

Mark I Source Term Bin

0-50 miles

0-100 miles

1.0E-01

1.0E+00

1.0E+01

1.0E+02

1.0E+03

1.0E+04

1 2 3 4 5 6 7 8 9

Land (sq. mi) Exceeding Long-Term 

Phase Habitability Criterion Per Event

Mark II Source Term Bin

0-50 miles

0-100 miles

Figure 4-17 Mark II source terms – land exceeding long-term phase habitability criterion 

(square miles) per event
4-58

The weather-averaged number of individuals displaced from their land for long-term phase 

interdiction or condemnation is shown in Figure 4-18 and Figure 4-19 for the 50-mile and 100-mile 

areas around each site. This metric does not consider the duration of time for which the land 

exceeds the habitability criterion and is considered contaminated. Some of the population might 

be able to return within one year following decontamination while others might need to stay away 

for a decade or more, or permanently.

The number of people displaced from interdicted and condemned land increases with source term 

magnitude and with release duration because a longer release allows more time for the wind to 

change direction and thus increases the probability of spreading contamination over population 

centers. The weather-averaged population displaced ranges from essentially zero for the smallest 

source terms to about 720,000 people within 50 miles for Mark II bin 8. This maximum represents 

about 9% of the total population within 50 miles of the Mark II reference site, Limerick. As with 

land contamination area, the people displaced from interdicted and condemned land is 

concentrated in the 50-mile area around each site. Even for the largest source terms, the 50-mile 

area dominates compared to the 50-100-mile area. For example, Mark II bin 8, which causes the 

highest population displaced, affects 721,000 people within 50 miles (~97% of the 100-mile 

population total) compared to just 20,000 people between 50 and 100 miles (~3% of the 100-mile 

population total). 

Figure 4-18 Mark I source terms – population displaced from interdicted and condemned 

land per event

1.0E-01

1.0E+00

1.0E+01

1.0E+02

1.0E+03

1.0E+04

1.0E+05

1.0E+06

1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18

Population Displaced from Interdicted and 

Condemned Land Per Event

Mark I Source Term Bin

0-50 miles

0-100 miles
4-59

Figure 4-19 Mark II source terms – population displaced from interdicted and condemned 

land per event

4.4 Sensitivity Analyses

4.4.1 Approach

The overall modeling approach for the offsite consequence analysis component of the CPRR 

rulemaking effort was to develop and use site-specific MACCS models of a reference BWR Mark I 

site and of a reference BWR Mark II site. Site-specific features of the model include meteorology, 

population distribution, land use, economic values, evacuation characteristics and timing, use of 

KI, and intermediate and long-term phase habitability criteria. However, this rulemaking is 

designed to apply to the entire fleet of U.S. BWR Mark I and Mark II plants and their various sites 

have many different features. There are 15 different Mark I sites and 5 different Mark II sites. In 

order to better capture some of the effects of different site features on the project’s results, a 

number of parameters and features were varied in sensitivity studies. These sensitivities help 

gauge the impact of different population sizes, different evacuation timelines, different 

nonevacuating cohort sizes, different intermediate phase durations, and different long-term phase 

habitability criteria on the calculated offsite consequence metrics. The following sections discuss 

the approach to each of these.

4.4.1.1 Population

A SecPop site file was created for each Mark I and Mark II site and the 50-mile population value 

was ranked for each. High, medium, and low population sites were selected to approximate the 

90th, 50th, and 10th percentiles, respectively. The population is based on the 2010 U.S. Census 

and is scaled to 2013 using state level population growth projections from the Census Bureau

using the same process described earlier in Section 3.3. Peach Bottom and Limerick were already 

selected as the reference Mark I and Mark II sites for the baseline calculations and these are used 

1.0E-01

1.0E+00

1.0E+01

1.0E+02

1.0E+03

1.0E+04

1.0E+05

1.0E+06

1 2 3 4 5 6 7 8 9

Population Displaced from Interdicted and 

Condemned Land Per Event

Mark II Source Term Bin

0-50 miles

0-100 miles
4-60

as the high population sites for the sensitivity calculations. Vermont Yankee and Hatch were 

selected as the medium and low population Mark I sites. Susquehanna and Columbia were 

selected as the medium and low population Mark II sites. Population and site data is summarized 

in Table 4-24. Population data for each Mark I site is presented in Table 4-25 and for each Mark II 

site in Table 4-26. Population growth multipliers are provided in Table 4-27 for Hatch, Vermont 

Yankee, Columbia, and Susquehanna.

The SecPop site file contains one value of farmland ($ per hectare) and one value of populated 

land ($ per person) for each individual spatial grid element across the entire modeling domain.

These values for the 50-mile area around each plant were weighted by the quantity of farmland in 

each and by the population in each to develop a weighted average value for the 50-mile area.

These values are also presented in Table 26 and are used in the population sensitivities.

The sensitivity calculations that use Vermont Yankee, Hatch, Susquehanna, or Columbia site data 

do not use any other site-specific modeling features that would pertain to these four sites. For 

example, the Mark II population sensitivity calculations using the low population site (Columbia) 

still use all other parameters and features for the Mark II reference site, Limerick. Thus the Mark II 

calculations with the Columbia site data still use Limerick meteorology, Limerick evacuation 

characteristics, and the Pennsylvania state-specific habitability criterion. These sensitivities are 

included simply to assess how the consequence metrics might change for a less populated area 

with lower weighted average economic values.

Table 4-24 Sites selected for population sensitivity calculations

Site Population 

within 50 miles 

Weighted Average 

Value of Farmland 

($ 2013 per 

hectare) 

Weighted Average 

Value of 

Populated Land ($ 

2013 per person) 

Mark I 

High Peach Bottom 5,645,811 $24,400 $518,000 

Medium Vermont Yankee 1,536,793 $20,000 $475,000 

Low Hatch 453,404 $7,800 $285,000 

Mark 

II 

High Limerick 8,108,436 $28,600 $528,000 

Medium Susquehanna 1,790,924 $14,800 $399,000 

Low Columbia 464,310 $5,800 $359,000 
4-61

Table 4-25 Mark I site population information

Table 4-26 Mark II site population information

Mark I Site Population 

within 50 mi 

Population 

Density 

(per sq mi) 

Rank Percentile 

Dresden 7,374,320 939 1 1.00 

HIGH Peach Bottom 5,645,811 719 2 0.93 

Hope Creek 5,633,411 717 3 0.86 

Pilgrim 4,851,642 618 4 0.79 

Fermi 4,808,370 612 5 0.71 

Oyster Creek 4,567,689 582 6 0.64 

Monticello 3,052,698 389 7 0.57 

MEDIUM Vermont Yankee 1,536,793 196 8 0.50 

Browns Ferry 997,194 127 9 0.44 

Fitzpatrick & Nine Mile Point 923,614 118 10 0.36 

Duane Arnold 673,752 86 11 0.29 

Quad Cities 653,780 83 12 0.21 

Brunswick 479,743 61 13 0.14 

LOW Hatch 453,404 58 14 0.07 

Cooper 159,946 20 15 0.00 

Mark II Site 

Population 

within 50 mi 

Population 

Density 

(per sq mi) 

Rank Percentile 

HIGH Limerick 8,108,436 1,032 1 1.00 

LaSalle 1,909,500 243 2 0.75 

MEDIUM Susquehanna 1,790,924 228 3 0.50 

Nine Mile Point 923,614 118 4 0.25 

LOW Columbia 464,310 59 5 0.00 
4-62

Table 4-27 Population growth multipliers for selected sensitivity sites

4.4.1.2 Selection of Source Terms for Use in Sensitivity Analyses

The source term bins used in the Mark I and Mark II baseline analyses span over 4 orders of 

magnitude for cesium release, and the results clearly depend on the size of the release. Rather 

than run sensitivity calculations for all 18 Mark I source term bins and all 9 Mark II source term 

bins, a low, medium, and high source term was selected for each containment type. Similar to the 

population values, the high source term was selected to approximate the 90th percentile, the 

medium source term was selected to represent the median value, and the low source term was 

selected to approximate the 10th percentile. Note that these percentiles are computed from 

considering the spread of bin results; these percentiles do not represent the distribution of 

frequency-weighted releases.

As shown in Table 4-28, bins 3, 10, and 17 were selected as the low, medium, and high source 

terms for the Mark I sensitivity calculations. Similarly, Table 4-29 shows that bins 2, 5, and 8 were 

selected as the low, medium, and high source terms for the Mark II sensitivity calculations.

Table 4-28 Selection of source terms for Mark I sensitivity calculations

Bin Rep Case Rep Case 

Cs (%)

Rep Case 

I (%)

Start 

Time 

(hrs)

# Hrs with 

Significant 

Cs 

Release*

Percentile

1 28DF1000 0.0006% 0.006% 14.9 7 0.000

2 48DF100 0.002% 0.02% 11.4 8 0.058

LOW 3 10DF100 0.0073% 0.08% 16.3 6 0.117

4 7DF1000 0.02% 0.26% 14.9 20 0.176

5 11DF10 0.06% 0.78% 14.4 4 0.235

6 48 0.23% 1.69% 11.4 8 0.294

7 15 0.60% 5.85% 14.9 7 0.352

Census 

2010 

Census 

2013 Est. 

2010 to 

2013 

Multiplier 

Hatch Vermont 

Yankee Columbia Susquehanna 

Approximate 50-mile Area Fraction 

GA 9,687,653 9,992,167 1.031 1 - - - 

MA 6,547,629 6,692,824 1.022 - 0.4 - - 

NH 1,316,470 1,323,459 1.005 - 0.3 - - 

NY 19,378,102 19,651,127 1.014 - 0.05 - - 

OR 3,831,074 3,930,065 1.026 - - 0.1 - 

PA 12,702,379 12,773,801 1.006 - - - 1 

VT 625,741 626,630 1.001 - 0.25 - - 

WA 6,724,540 6,971,406 1.037 - - 0.9 - 

Population Multiplier for 2010 to 2013: 1.031 1.012 1.036 1.006 
4-63

Bin Rep Case Rep Case 

Cs (%)

Rep Case 

I (%)

Start 

Time 

(hrs)

# Hrs with 

Significant 

Cs 

Release*

Percentile

8 46 0.98% 11.01% 14.8 17 0.411

9 5DF10 1.05% 2.89% 24.2 34 0.470

MEDIUM 10 5 1.39% 6.46% 24.2 41 0.529

11 8 1.49% 19.25% 14.9 5 0.588

12 1 1.93% 22.68% 14.9 22 0.647

13 41DF1000 3.40% 7.65% 9.8 17 0.823

14 22dw 2.82% 18.64% 14.9 27 0.764

15 53 2.79% 29.05% 17.4 13 0.705

16 41 4.54% 14.10% 9.8 16 0.882

HIGH 17 3DF10 8.85% 24.65% 9.8 63 0.941

18 52 15.90% 34.32% 17.4 11 1.000

Note: For the purpose of quantifying the time signature of a source term release, an hourly plume 

segment is considered “significant” if it contributes at least 0.5% of that source term’s total cumulative 

cesium release to the environment

Table 4-29 Selection of source terms for Mark II sensitivity calculations

Note: For the purpose of quantifying the time signature of a source term release, an hourly plume 

segment is considered “significant” if it contributes at least 0.5% of that source term’s total cumulative 

cesium release to the environment

Bin Rep Case Rep Case 

Cs (%) 

Rep 

Case I 

(%) 

Start 

Time 

(hrs) 

# Hrs with 

Significant 

Cs 

Release* 

Percentile 

1 11DF1000 0.00004% 0.0005% 20.3 20 0 

LOW 2 5DF1000 0.00055% 0.005% 32.2 20 0.125 

3 42DF100 0.0043% 0.037% 14.3 13 0.25 

4 11 0.042% 0.45% 20.3 20 0.375 

MEDIUM 5 51DF10 0.23% 2.01% 16.6 9 0.5 

6 5 0.55% 4.94% 32.2 20 0.625 

7 3 1.09% 10.26% 14.3 20 0.75 

HIGH 8 1 2.46% 19.81% 22.8 25 0.875 

9 52 3.57% 28.67% 16.6 10 1 
4-64

4.4.1.3 Evacuation Delay

Each accident progression and source term scenario modeled in this rulemaking technical basis 

begins with an ELAP. The ELAP is postulated to come from both internal and external events. For 

certain external events that could cause the ELAP, e.g., earthquake or flood, they could also slow 

the EPZ evacuation effort. For example, an earthquake could damage roads or bridges and cut 

power to traffic signals. A flood could obstruct access to certain road sections and also cut power 

to traffic signals. These would result in a slower evacuation because people would need to travel 

farther to avoid obstructed areas and traffic signals might take longer to pass through. Therefore a 

sensitivity calculation was included that delays the start of evacuation for all population cohorts by 

1 hour to assess the impact of this delay on the offsite consequence metrics. The uniform delay of 

1 hour was modeled by changing the MACCS OALARM parameter from 0 to 1 hour.

4.4.1.4 Nonevacuating Cohort

The EPZ population around the reference Mark I site, Peach Bottom, and the reference Mark II 

site, Limerick, were studied and segmented into population cohorts with distinct evacuation delay 

and travel characteristics. For each evacuation model, there is a cohort that is modeled to not 

evacuate, for any number of reasons including refusing to respond to EAS messaging and sirens 

or not receiving emergency alert communication. SOARCA and other studies have assumed this 

group to be 0.5% of the EPZ population. This percentage is consistent with research on largescale evacuations that has shown a small percentage of the public refuses to evacuate [44]. The 

MACCS model for the reference Mark II site uses 0.5% for the nonevacuating cohort. The 

MACCS model for the reference Mark I site, however, uses 2% for the nonevacuating cohort to 

represent the expected emergency response of the Amish community within the Peach Bottom 

EPZ. According to the recently submitted ETE for Peach Bottom, Amish men and boys aged 15 

and older would not evacuate and would stay on their land [22].

The nonevacuating cohort size is commonly perceived to influence consequences related to early 

phase exposures and is therefore included as a sensitivity case. For each set of Mark I and Mark 

II sensitivity calculations, a sensitivity run is included that increases the nonevacuating cohort to 

5% of the EPZ population. The 5% value was selected based on discussion with senior NRC 

emergency preparedness staff. To accommodate the increase in population of the nonevacuating 

cohort, the “general” cohort for each site was reduced by the same percentage. Sensitivity 

calculations that use a 5% nonevacuating cohort do not also use the 1-hour evacuation delay 

described in the previous section because the causes of these hypothetical situations are 

considered independent.

4.4.1.5 Intermediate Phase Duration

The intermediate phase begins after the reactor and releases have been brought under control 

and accounts for the time needed to plan the long-term restoration and cleanup activities before 

they can begin. These activities include the following:

1. defining the areas of interest,

2. characterizing the contamination using dose data and field surveys,

3. identifying the types of materials to be decontaminated and the equipment and

personnel needed,

4. developing a waste management plan including estimating waste volumes, storage

requirements, storage locations, and acquiring storage materials,
4-65

5. acquiring decontamination equipment and bringing it onsite, and

6. training personnel and bringing them onsite.

The length of time to conduct these processes would depend on a variety of factors, including the 

extent and location of the contamination, cleanup criteria, material types, and the state and local 

decision processes. As the distance from the plant increases, the level of contamination is 

expected to decrease; however, other influences may increase because a wider variety of 

materials may be involved and additional towns, counties, and stakeholders would become 

involved. 

Based on the number and types of planning processes that are needed prior to starting 

decontamination, 3 months was selected as an average intermediate phase duration and was 

used for all baseline calculations. Some decontamination would be expected to start prior to 3 

months; however, much could take longer to begin. To assess the impact of the intermediate 

phase duration, sensitivity calculations are included using a low value (zero) and a high value (1 

year). The dose criterion applied to the intermediate phase uses the long-term phase criterion 

scaled to the applicable time duration. 

4.4.1.6 Long-Term Phase Habitability Criterion

The long-term phase habitability criterion refers to a projected dose accumulated over a specific 

time and is used to determine whether people can reside on land and whether people can use 

farmland for agricultural production (assuming the crops meet the farmability criterion). The U.S. 

EPA provides guidance in its PAG Manual [18] for a habitability criterion of 2 rem in the first year 

followed by 500 mrem each year thereafter. Pennsylvania, however, uses a stricter habitability 

criterion of 500 mrem-per-year, starting in the first year. Because both Mark I and Mark II 

reference sites are located in Pennsylvania, the 500 mrem-per-year long-term phase habitability 

criterion was used in all baseline calculations. Most BWR Mark I/II sites are not in Pennsylvania 

and would likely implement the EPA’s recommended habitability criterion. To assess the impact of 

this parameter on the consequence metrics, the EPA’s recommended value was used in 

sensitivity calculations. MACCS requires one dose and one time period, so to accommodate this, 

2 rem-per-year was used for each year. The specific intermediate phase and long-term phase 

habitability parameters are provided in Table 4-30 below. The second and third columns show the 

sensitivity parameter choices. The next four columns show the implementation in MACCS, which 

in some cases is slightly different. For variations 4 and 6 in which 2 rem is applied in year 1 of the 

long-term phase, some or all of this 2 rem is applied to the intermediate phase. In variation 6, the 

intermediate phase is a full year, so all of the 2 rem is applied to the intermediate phase and then 

the long-term phase uses 500 mrem each year. In variation 4, the intermediate phase is 3 months 

so 1/4th of the 2 rem is applied to the intermediate phase (500 mrem) and the remaining 1.5 rem is 

applied to the long-term phase. Because this 1.5 rem would apply to the first 9 months of the longterm phase year 1, an additional 0.125 rem is included (1/4th of the second year’s 500 mrem) 

totaling 1.625 rem. 
4-66

Table 4-30 Intermediate and long-term phase sensitivity calculation parameters

4.4.2 Sensitivity Calculations Supporting Regulatory Analysis: Results and Discussion

The results of the 144 sensitivity calculations are provided in Table I-1 through Table I-6 in 

Appendix I. Each table contains the results of 24 calculations corresponding to a base MACCS 

model (either the reference site for Mark I, Peach Bottom, or the reference site for Mark II, 

Limerick) and a low, medium, or high population site file. Each table contains 8 calculations for 

each of the three source terms used—low, medium, and high. The 8 calculations for each base 

model, site file, and source term are as follows: 

Run 1) Baseline parameters

Run 2) Baseline parameters but with 1 hour evacuation delay

Run 3) Baseline parameters but with no intermediate phase

Run 4) Baseline parameters but with 1 year intermediate phase

Run 5) Baseline parameters but with EPA PAG recommended long-term phase habitability 

criterion

Run 6) Baseline parameters but with no intermediate phase and EPA PAG recommended 

long-term phase habitability criterion

Run 7) Baseline parameters but with 1 year intermediate phase and EPA PAG 

recommended long-term phase habitability criterion

Run 8) Baseline parameters but with 5% nonevacuating cohort fraction

4.4.2.1 Impact of Evacuation Delay

Comparing the Run 2 sensitivity to the Run 1 baseline sensitivity for each of the 18 groupings (2 

reference models × 3 site files × 3 source terms) shows zero change to all of the offsite 

consequences as a result of the 1-hour evacuation delay. All of the source terms used in the 

sensitivity calculations have an environmental release that is delayed sufficiently to allow time for 

evacuation of the EPZ. Table 4-31 shows the ratio of each consequence for the evacuation delay 

case to the baseline case for each of the 18 groupings.

Implementation in MACCS 

Intermediate Phase Long-Term Phase 

Variation 

Intermediate 

Phase 

Duration (yrs) 

Long-Term 

Phase 

Habitability 

Criterion 

Duration 

and Dose 

Projection 

Period 

(yrs) 

Dose 

Criterion 

Dose 

Projection 

Period 

(yrs) 

Dose 

Criterion 

1 0.25 500 mrem 0.25 125 mrem 1 500 mrem 

2 0 500 mrem 0 0 1 500 mrem 

3 1 500 mrem 1 500 mrem 1 500 mrem 

4 0.25 2 rem 0.25 500 mrem 1 1.625 rem 

5 0 2 rem 0 0 1 2 rem 

6 1 2 rem 1 2 rem 1 500 mrem 
4-67

Table 4-31 Ratio of consequences for evacuation delay sensitivity cases to baseline 

cases

* Indicates that both the numerator and denominator in the ratio are zero

4.4.2.2 Impact of Larger Nonevacuating Cohort

Comparing the Run 8 sensitivity to the Run 1 baseline sensitivity for each of the 18 groupings (2 

reference models × 3 site files × 3 source terms) shows either small or zero change to the offsite 

consequences as a result of the larger nonevacuating cohort. Table 4-32 shows the ratio of each 

consequence for the nonevacuating cohort sensitivity case to the baseline case for each of the 18

groupings. This table uses colored formatting to help visualize trends. Higher ratios appear in 

green while lower ratios appear in red.

Individual early fatality risk is essentially zero for all cases analyzed because even the largest 

source terms combined with meteorological conditions are not able to cause sufficiently large 

acute doses. The individual LCF risk within 10 miles shows the largest difference among the 

consequence metrics. For the Mark II sensitivities, in which the nonevacuating cohort size

increased from 0.5%to 5%, the 10-mile area LCF risk increased by 3-4% for the small and 

medium source terms and by 10-20% for the high source term. For the Mark I sensitivities, in 

which the nonevacuating cohort is increased from 2-5%, the increase in individual LCF risk is 

smaller, about 1-2%for all three source terms. The increase is smaller for the Mark I sensitivities 

because the increase in fraction of the nonevacuating cohort is smaller. Overall, the impact of a 

larger nonevacuating cohort is fairly small because early phase relocation is still projected and 

modeled, which would ultimately help this population avoid or minimize exposures. 

The 50-mile LCF risk and the 50-mile total population dose show an increase of between 0 and 

1% and the increase drops when larger areas (0-100 miles) are considered. The land area 

exceeding the long-term habitability criterion and the population subject to long-term protective 

actions show no change because these are not affected by the early phase. The total offsite cost 

decreases slightly because when fewer people evacuate, fewer people would hypothetically be 

paid the per diem cost. This would decrease the evacuation cost of the accident. The evacuation 

cost is a large contributor to the total offsite cost for the low source term, so the low source terms 

Individual Early 

Fatality Risk

0-1.3 mi and 

beyond 0-10 mi 0-50 mi 0-100 

mi 0-50 mi 0-100 

mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi 0-50 mi 0-100 

mi

Mark I - Low (Bin 3) 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

Mark I - Med (Bin 10) 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

Mark I - High (Bin 17) 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

Mark I - Low (Bin 3) 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

Mark I - Med (Bin 10) 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

Mark I - High (Bin 17) 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

Mark I - Low (Bin 3) 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

Mark I - Med (Bin 10) 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

Mark I - High (Bin 17) 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

Mark II - Low (Bin 2) 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 * *

Mark II - Med (Bin 5) 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

Mark II - High (Bin 8) 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

Mark II - Low (Bin 2) 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 * *

Mark II - Med (Bin 5) 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

Mark II - High (Bin 8) 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

Mark II - Low (Bin 2) 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 * *

Mark II - Med (Bin 5) 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

Mark II - High (Bin 8) 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

Mark II - Limerick

Low - 

Columbia

Medium - 

Susquehanna

High - 

Limerick

Land (sq mi) 

Exceeding LongTerm 

Habitability 

Criterion

Individual Latent Cancer

Fatality Risk

 Population Dose

 (person-rem) 

Offsite Cost

 ($ 2013)

Population 

Subject to LongTerm Protective 

Actions

Mark I - Peach Bottom

Low - Hatch

Individual early 

fatality risk is 

zero for all 

baseline and 

sensitivity cases.

Medium - 

Vermont 

Yankee

High - Peach 

Bottom

Base Model

Site File Source Term
4-68

show a decrease in total offsite cost by about 1-3%. For the medium and large source terms, the 

evacuation cost is a very small component of the total offsite cost, so these show no impact. 

Table 4-32 Ratio of consequences for larger nonevacuating cohort sensitivity cases to 

baseline cases

* Indicates that both the numerator and denominator in the ratio are zero

4.4.2.3 Impact of Intermediate Phase Duration

The impacts of changing the intermediate phase duration to zero and to one year can be seen by 

comparing the Run 3 and 4 sensitivities to the Run 1 baseline sensitivity for each of the 18 

groupings. Table 4-33 and Table 4-34 show the ratio of each consequence for the one-year and 

zero intermediate phase duration sensitivities, respectively, compared to the baseline cases using 

a 0.25 year intermediate phase duration. These tables use colored formatting; larger ratios appear 

in green while smaller ratios appear in red. 

For individual LCF risk, using zero intermediate phase increases the results for the small source 

term (~7-14%) and decreases the results for the medium and large source terms (~4-17%). Using 

a 1 year intermediate phase generally shows the opposite trend: individual LCF risk decreases for 

the small source term (~10%) and generally increases for the large source terms (2-18%). For the 

medium source term the results are within +/- 10%.

Using no intermediate phase generally increases the total population dose, whereas using a 1 

year intermediate phase generally decreases the total population dose. The total offsite cost 

varies between about +60% for the high source term for a 1 year intermediate phase to about -

40% for the medium and large source term for the cases with no intermediate phase.

The use of a longer intermediate phase (1 year) allows more time for radionuclides to decay and 

be removed by natural weathering and therefore less land (~4-40% reduction) is contaminated 

when the long-term phase begins. Therefore fewer people are subject to long-term phase 

protective actions (~60-90% reduction) assuming the source term is sufficiently large (medium 

and high source terms). The low source term is not high enough to cause offsite contamination so 

land contamination is unaffected. The opposite is true for the use of a shorter (zero) intermediate 

Individual 

Early 

Fatality Risk

0-1.3 mi and 

beyond 0-10 mi 0-50 mi 0-100 

mi 0-50 mi 0-100 

mi 0-50 mi 0-100 

mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi

Mark I - Low (Bin 3) 1.02 1.00 1.00 1.00 1.00 0.98 0.98 1.00 1.00 1.00 1.00

Mark I - Med (Bin 10) 1.01 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

Mark I - High (Bin 17) 1.01 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

Mark I - Low (Bin 3) 1.02 1.00 1.00 1.00 1.00 0.98 0.98 1.00 1.00 1.00 1.00

Mark I - Med (Bin 10) 1.01 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

Mark I - High (Bin 17) 1.01 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

Mark I - Low (Bin 3) 1.02 1.00 1.00 1.00 1.00 0.98 0.98 1.00 1.00 1.00 1.00

Mark I - Med (Bin 10) 1.01 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

Mark I - High (Bin 17) 1.01 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

Mark II - Low (Bin 2) 1.04 1.00 1.00 1.00 1.00 0.99 0.99 1.00 1.00 * *

Mark II - Med (Bin 5) 1.03 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

Mark II - High (Bin 8) 1.13 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

Mark II - Low (Bin 2) 1.04 1.01 1.00 1.00 1.00 0.97 0.97 1.00 1.00 * *

Mark II - Med (Bin 5) 1.03 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00

Mark II - High (Bin 8) 1.17 1.01 1.00 1.01 1.00 1.00 1.00 1.00 1.00 1.00 1.00

Mark II - Low (Bin 2) 1.04 1.01 1.01 1.01 1.00 0.96 0.96 1.00 1.00 * *

Mark II - Med (Bin 5) 1.04 1.00 1.00 1.01 1.00 1.00 1.00 1.00 1.00 1.00 1.00

Mark II - High (Bin 8) 1.20 1.01 1.01 1.01 1.00 1.00 1.00 1.00 1.00 1.00 1.00

Individual 

early fatality 

risk is zero 

for all 

baseline and 

sensitivity 

cases.

Land (sq mi) 

Exceeding 

Long-Term 

Habitability 

Criterion

High - 

Limerick

Mark II - Limerick

Low - 

Columbia

Medium - 

Susquehanna

Mark I - Peach Bottom

Low - Hatch

Medium - 

Vermont 

Yankee

High - Peach 

Bottom

Individual Latent Cancer

Fatality Risk

 Population Dose

 (person-rem) 

Offsite Cost

 ($ 2013)

Population 

Subject to 

Long-Term 

Protective 

Actions

Base Model

Site File Source Term
4-69

phase. The long-term phase starts sooner and because there is less time for radionuclide decay 

and weathering, more land is contaminated and more people are displaced. The population 

subject to long-term protective actions increases by up to 550% but is more commonly in the 

range of 200-300%. The area of land exceeding the long-term phase habitability criterion 

increases in the range of 2-80%.

The total offsite cost generally increases with the longer intermediate phase for the large source 

term and decreases with zero intermediate phase. Even though a longer intermediate phase allows 

more time for decay and weathering, the cost of intermediate phase relocation also increases (50-

60% within 50 miles) because the per diem cost is applied for each relocatee for 4 times longer. The 

small source term is not large enough to necessitate nearly as much, if any, intermediate phase

relocation, so there is no effect on offsite cost. The medium source term sensitivity case with the 

longer intermediate phase results in an increased cost of about 0-40%. The total offsite cost 

generally decreases by about 20-% for the sensitivities with no intermediate phase for the medium 

and large source terms and is less affected for the small source term (0-6%).

Table 4-33 Ratio of consequences for 1-year intermediate phase duration sensitivity 

cases to baseline cases

* Indicates that both the numerator and denominator in the ratio are zero

Individual 

Early 

Fatality 

Risk

0-1.3 mi

and beyond 0-10 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi

Mark I - Low (Bin 3) 0.88 0.89 0.88 0.98 0.99 0.98 0.98 1.00 1.00 0.00 0.00

Mark I - Med (Bin 10) 1.07 0.93 0.91 0.97 0.97 1.38 1.18 0.86 0.92 0.48 0.48

Mark I - High (Bin 17) 1.04 0.98 0.93 0.98 0.96 1.61 1.39 0.80 0.87 0.60 0.53

Mark I - Low (Bin 3) 0.88 0.88 0.88 0.94 0.95 0.96 0.96 1.00 1.00 0.14 0.14

Mark I - Med (Bin 10) 1.06 0.92 0.89 0.93 0.92 1.39 1.04 0.73 0.86 0.57 0.57

Mark I - High (Bin 17) 1.02 0.97 0.91 0.97 0.92 1.58 1.33 0.71 0.82 0.59 0.46

Mark I - Low (Bin 3) 0.88 0.89 0.88 0.95 0.95 0.97 0.97 1.00 1.00 0.16 0.16

Mark I - Med (Bin 10) 1.07 0.92 0.90 0.93 0.92 1.31 1.16 0.91 0.94 0.39 0.39

Mark I - High (Bin 17) 1.04 0.97 0.93 0.97 0.94 1.60 1.46 0.86 0.89 0.55 0.51

Mark II - Low (Bin 2) 0.90 0.93 0.93 0.99 0.99 1.00 1.00 1.00 1.00 * *

Mark II - Med (Bin 5) 0.96 0.92 0.92 0.98 0.98 1.00 1.00 0.99 1.00 0.29 0.29

Mark II - High (Bin 8) 1.18 0.98 0.98 0.98 0.98 1.50 1.49 0.86 0.90 0.20 0.19

Mark II - Low (Bin 2) 0.90 0.93 0.93 0.96 0.96 1.00 1.00 1.00 1.00 * *

Mark II - Med (Bin 5) 0.98 0.93 0.90 0.95 0.93 1.18 1.11 0.94 0.97 0.44 0.44

Mark II - High (Bin 8) 1.18 0.98 0.98 0.97 0.97 1.63 1.49 0.62 0.81 0.26 0.21

Mark II - Low (Bin 2) 0.90 0.93 0.93 0.93 0.94 1.00 1.00 1.00 1.00 * *

Mark II - Med (Bin 5) 1.00 0.92 0.91 0.94 0.93 1.08 1.06 0.96 0.97 0.45 0.45

Mark II - High (Bin 8) 1.17 0.97 0.98 0.95 0.96 1.57 1.48 0.68 0.81 0.21 0.20

Individual 

early fatality 

risk is zero 

for all 

baseline 

and 

sensitivity 

cases.

Land (sq mi) 

Exceeding LongTerm Habitability 

Criterion

Mark II - Limerick

Low - 

Columbia

Medium - 

Susquehanna

High - 

Limerick

High - Peach 

Bottom

Offsite Cost

 ($ 2013)

Population 

Subject to LongTerm Protective 

Actions

Mark I - Peach Bottom

Low - Hatch

Medium - 

Vermont 

Yankee

Individual Latent Cancer

Fatality Risk

 Population Dose

 (person-rem) 

Base Model

Site Source Term
4-70

Table 4-34 Ratio of consequences for zero intermediate phase sensitivity cases to 

baseline cases

* Indicates that both the numerator and denominator in the ratio are zero

∗∗ Indicates that the denominator in the ratio is zero but the numerator is nonzero (~0.16 persons subject to

long-term protective actions for Mark II – High Site File – Low Source Term

4.4.2.4 Impact of Long-Term Phase Habitability Criterion

Comparing the Run 5 sensitivity to the Run 1 baseline sensitivity for each of the 18 groupings 

shows a much larger impact on the results than the early phase model sensitivities. Table 4-35

shows the ratio of each consequence metric for the 2 rem-per-year habitability criterion sensitivity 

case to the baseline case using 500 mrem-per-year. This table uses colored formatting to facilitate 

observation of trends. Larger ratios appear in green, while smaller ratios appear in red.

The 2 rem-per-year habitability criterion allows up to 4 times greater annual dose in the first year 

than the Pennsylvania criterion of 500 mrem-per-year and therefore the population is exposed to 

more radiation. Because a higher exposure level is tolerated, less land is considered 

contaminated, fewer people need to be displaced from land, and there is a lower cost reflecting 

less decontamination needed, less property loss-of-use costs, and less compensation costs.

The two health effect consequence metrics, individual LCF risk and total population dose, 

increase when using the 2 rem-per-year habitability criterion. Individual LCF risk within 10 miles 

increases by 6-7% for the small source term but increases by about 50-250% for the medium and 

large source terms. The largest change is for the Mark I reference model with the Peach Bottom 

site file and the large source term: individual LCF risk increases from 7.1E-04 to 1.74E-03, an 

increase of about 246%. The increases become smaller when larger areas are considered. The 

total population dose for the 50-mile area shows a negligible increase for the small source term 

but increases by about 4-48% for the medium and large source terms. The increase to the 50-mile 

population dose is clearly larger than the increase to the 10-mile individual LCF risk but is very 

similar to the increase in individual LCF risk for the 50-mile area.

Offsite cost and population and land exceeding the long-term phase habitability criterion decrease 

when using the 2-rem-per-year habitability criterion. There is either no change or a very small 

Individual 

Early 

Fatality 

Risk

0-1.3 mi

and beyond 0-10 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi

Mark I - Low (Bin 3) 1.14 1.09 1.09 1.01 1.01 0.97 0.97 1.00 1.00 11.27 11.27

Mark I - Med (Bin 10) 0.87 0.99 1.05 1.00 1.02 0.63 0.60 1.44 1.27 2.45 2.56

Mark I - High (Bin 17) 0.94 0.92 0.99 0.98 1.00 0.66 0.62 1.37 1.29 1.72 2.13

Mark I - Low (Bin 3) 1.12 1.09 1.09 1.05 1.04 0.94 0.95 1.03 1.02 16.29 16.29

Mark I - Med (Bin 10) 0.88 1.01 1.08 1.02 1.06 0.61 0.48 1.77 1.42 2.58 2.92

Mark I - High (Bin 17) 0.96 0.93 1.00 0.97 1.01 0.65 0.58 1.52 1.40 1.87 2.63

Mark I - Low (Bin 3) 1.14 1.09 1.09 1.04 1.04 0.97 0.97 1.00 1.00 20.08 20.08

Mark I - Med (Bin 10) 0.86 1.00 1.05 1.01 1.04 0.62 0.56 1.29 1.21 3.12 3.34

Mark I - High (Bin 17) 0.94 0.92 0.98 0.96 0.99 0.65 0.62 1.27 1.27 1.85 2.17

Mark II - Low (Bin 2) 1.07 1.05 1.05 1.01 1.00 1.00 1.00 1.00 1.00 * *

Mark II - Med (Bin 5) 0.87 1.03 1.04 1.01 1.01 0.70 0.73 1.02 1.02 5.51 5.51

Mark II - High (Bin 8) 0.83 0.91 0.94 1.04 1.03 0.83 0.83 1.29 1.45 2.22 2.69

Mark II - Low (Bin 2) 1.07 1.05 1.05 1.03 1.03 1.00 1.00 1.02 1.02 * *

Mark II - Med (Bin 5) 0.84 1.02 1.06 1.02 1.04 0.58 0.58 1.24 1.12 3.32 3.32

Mark II - High (Bin 8) 0.85 0.93 1.05 1.09 1.11 0.77 0.70 1.78 1.73 2.62 6.04

Mark II - Low (Bin 2) 1.07 1.05 1.05 1.05 1.05 1.00 1.00 1.00 1.00 ** **

Mark II - Med (Bin 5) 0.84 1.04 1.05 1.04 1.04 0.55 0.55 1.18 1.10 3.33 3.33

Mark II - High (Bin 8) 0.87 0.90 0.96 1.10 1.11 0.83 0.78 1.65 1.67 2.57 3.09

Individual 

early fatality 

risk is zero 

for all 

baseline 

and 

sensitivity 

cases.

Land (sq mi) 

Exceeding LongTerm Habitability 

Criterion

Mark II - Limerick

Low - 

Columbia

Medium - 

Susquehanna

High - 

Limerick

Medium - 

Vermont 

Yankee

High - Peach 

Bottom

Individual Latent Cancer

Fatality Risk

 Population Dose

 (person-rem) 

Offsite Cost 

($ 2013)

Population 

Subject to LongTerm Protective 

Actions

Mark I - Peach Bottom

Low - Hatch

Base Model

Site Source Term
4-71

change for the cases using the small source term. For the 50-mile area and for the medium and 

large source terms, the offsite cost decreases by about 30-50%, the land contaminated decreases 

by about 10-50%, and the population subject to long-term phase protective actions decreases by 

about 70-95%. 

Table 4-35 Ratio of consequences for 2 rem-per-year long-term phase habitability 

criterion sensitivity cases to baseline cases

* Indicates that both the numerator and denominator in the ratio are zero

4.4.2.5 Impact of Population Density and Site File Data 

For each source term and containment base model, the high, medium, and low site file cases can 

be examined to assess the impact of population and economic values on the consequence 

results. Table 4-36 provides ratios of each consequence for the medium and high site files 

compared to the low site file for each of the 18 groupings.

Individual LCF risk is relatively insensitive to site file data. Comparing the medium and high 

population cases to the low population case, individual LCF risk within 10 miles increases by up to 

60% but also decreases by up to about 10%. When considering larger areas, individual LCF risk 

generally decreases as population increases.

Population dose is directly related to population size, so the sensitivity cases show a strong 

increase in population dose for larger population site files. For example, for the Mark II high 

source term, the high site file case has a population dose about 11 times higher than the low site 

file case. There are some cases in which a higher population site file does not lead to a higher 

population dose. For the Mark II low source term cases, the medium site file (Susquehanna) has a 

lower 50-mile population dose than the low site file (Columbia). This is likely due to the population 

distribution across the spatial grid in relation to the wind rose. Susquehanna has a 50-mile 

population almost 4 times larger than Columbia, yet Columbia may happen to have more people 

distributed in spatial grid elements that intersect with the prevailing wind directions characterized 

in the weather file.

Individual Early 

Fatality Risk

0-1.3 mi and 

beyond 0-10 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi

Mark I - Low (Bin 3) 1.06 1.01 1.01 1.00 1.00 0.97 0.97 1.00 1.00 0.00 0.00

Mark I - Med (Bin 10) 2.29 1.32 1.19 1.15 1.06 0.38 0.40 0.79 0.88 0.21 0.21

Mark I - High (Bin 17) 2.41 1.63 1.35 1.33 1.16 0.42 0.37 0.67 0.80 0.32 0.27

Mark I - Low (Bin 3) 1.07 1.02 1.01 1.01 1.00 0.93 0.93 1.00 1.00 0.01 0.01

Mark I - Med (Bin 10) 2.38 1.29 1.12 1.24 1.09 0.31 0.25 0.59 0.78 0.32 0.32

Mark I - High (Bin 17) 2.42 1.56 1.25 1.46 1.21 0.34 0.24 0.51 0.71 0.31 0.21

Mark I - Low (Bin 3) 1.06 1.00 1.00 1.00 1.00 0.97 0.97 1.00 1.00 0.03 0.03

Mark I - Med (Bin 10) 2.29 1.24 1.16 1.19 1.12 0.27 0.25 0.87 0.91 0.14 0.14

Mark I - High (Bin 17) 2.46 1.57 1.36 1.48 1.30 0.33 0.28 0.77 0.83 0.26 0.23

Mark II - Low (Bin 2) 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 * *

Mark II - Med (Bin 5) 1.49 1.14 1.12 1.04 1.02 0.52 0.57 0.99 0.99 0.05 0.05

Mark II - High (Bin 8) 2.24 1.35 1.32 1.18 1.13 0.53 0.53 0.89 0.92 0.32 0.30

Mark II - Low (Bin 2) 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 * *

Mark II - Med (Bin 5) 1.68 1.20 1.09 1.14 1.06 0.34 0.37 0.91 0.96 0.11 0.11

Mark II - High (Bin 8) 2.35 1.35 1.24 1.28 1.19 0.46 0.37 0.66 0.82 0.32 0.26

Mark II - Low (Bin 2) 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 1.00 * *

Mark II - Med (Bin 5) 1.74 1.17 1.14 1.14 1.11 0.29 0.30 0.93 0.96 0.16 0.16

Mark II - High (Bin 8) 2.32 1.34 1.30 1.29 1.25 0.46 0.41 0.72 0.83 0.30 0.29

Land (sq mi) 

Exceeding LongTerm Habitability 

Criterion

Individual early 

fatality risk is zero 

for all baseline and 

sensitivity cases.

Medium - 

Susquehanna

High - 

Limerick

High - Peach 

Bottom

Mark II - Limerick

Low - 

Columbia

Population Subject 

to Long-Term 

Protective Actions

Mark I - Peach Bottom

Low - Hatch

Medium - 

Vermont 

Yankee

Individual Latent Cancer

Fatality Risk

 Population Dose

 (person-rem) 

Offsite Cost

 ($ 2013)

Base Model

Site Source Term
4-72

For a given source term, the total offsite cost always increases with higher population site files.

Higher population site files also have more valuable farmland and populated land, so both 

population and economic values drive the costs up. Comparing the medium to low site file 

baseline sensitivity calculations, the medium site file calculations have about 2-7 times higher 

costs. Comparing the high to low site file baseline sensitivity calculations, the high site file 

calculations have about 4-20 times higher costs.

The population subject to long-term protective actions increases with higher population site files 

as expected. As expected, the area of land exceeding the long-term phase habitability criterion 

does not show any clear trends with the site file population and economic values. Land areas 

depend primarily on land fraction (land versus water), which is another type of information 

contained in the site file.

Table 4-36 Results for baseline cases with different site files

* Indicates that both the numerator and denominator in the ratio are zero

4.4.3 Environmental Release Timing Sensitivity

A separate sensitivity calculation was performed to examine the effect of the start time of 

environmental release on the consequence metrics. The representative source term case for each 

source term bin was selected by choosing the one with a cesium and iodine release closest to the 

average among the source terms in that bin. The selection of the representative source term case 

did not consider the time at which release to the environment begins because all of them provide 

sufficient time for evacuation of the EPZ population prior to the environmental release. To confirm 

this, a sensitivity calculation was run to compare the offsite consequence results for the source 

term case with the fastest environmental release of all the cases to the representative source term 

case for that source term’s bin. The fastest environmental release among all considered was 7.3 

hours for case 50 and is compared to case 48, the representative case for bin 6, which has a 

release to the environment beginning at 11.4 hours. The cesium release is very similar for each of 

these but they differ in many other ways beyond just the environmental release start time. The 

iodine release is larger in case 48 (1.69% vs. 1.09%). The release profile is also somewhat 

different: case 50 releases 75% of the source term’s cesium in the first hour whereas case 48’s 

hour-long plume segment with the largest cesium release is only about 43% of the source term’s 

total cesium.

Table 4-37 shows the offsite consequence metrics for Mark I case 48 and case 50. Consistent 

with all other calculations, individual early fatality risk is essentially zero for both cases. For 

individual LCF risk, population dose, offsite cost, and land exceeding the long-term phase 

habitability criterion, the consequences are about 10-30% higher for case 48. However, the 

Individual 

Early Fatality 

Risk

0-1.3 mi and 

beyond 0-10 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi

Med (VT Yankee) / Low (Hatch) 1.52 0.98 0.90 0.92 1.19 2.79 2.75 0.39 0.43 6.20 6.20

High (Peach Bottom) / Low (Hatch) 0.94 0.74 0.96 2.82 2.07 4.65 4.57 1.53 1.45 2.07 2.07

Med (VT Yankee) / Low (Hatch) 1.25 0.98 0.97 1.88 2.37 3.08 3.60 0.67 0.72 2.91 2.92

High (Peach Bottom) / Low (Hatch) 1.02 0.83 1.02 5.83 4.00 8.84 8.22 1.28 1.08 7.15 7.15

Med (VT Yankee) / Low (Hatch) 1.23 1.05 1.08 2.26 3.33 3.58 4.95 0.82 0.82 3.11 4.16

High (Peach Bottom) / Low (Hatch) 1.00 0.89 1.00 6.78 5.04 11.11 9.33 1.11 0.98 9.96 9.59

Med (Susquehanna) / Low (Columbia) 1.20 0.93 0.49 0.70 1.00 4.90 4.90 3.93 3.93 * *

High (Limerick) / Low (Columbia) 1.63 1.10 0.69 2.33 2.25 20.48 20.48 12.79 12.79 * *

Med (Susquehanna) / Low (Columbia) 0.94 0.86 0.49 1.38 1.96 2.32 2.33 0.40 0.56 6.35 6.35

High (Limerick) / Low (Columbia) 1.17 1.03 0.65 6.53 4.82 11.71 10.63 0.52 0.61 28.96 28.96

Med (Susquehanna) / Low (Columbia) 0.89 0.85 0.59 2.06 3.71 3.07 6.60 0.61 0.76 3.00 3.42

High (Limerick) / Low (Columbia) 1.07 1.04 0.68 10.82 9.32 18.49 17.97 0.69 0.75 17.87 17.09

Land (sq mi) 

Exceeding LongTerm Habitability 

Criterion

Mark I - Peach 

Bottom

Mark I - Low (Bin 3)

Mark I - Med (Bin 10)

Mark I - High (Bin 17)

Mark II - Limerick

Mark II - Low (Bin 2)

Mark II - Med (Bin 5)

Mark II - High (Bin 8)

Individual 

early fatality 

risk is zero 

for all 

baseline and 

sensitivity 

cases.

Individual Latent Cancer

Fatality Risk

 Population Dose

 (person-rem) 

Offsite Cost

 ($ 2013)

Population Subject 

to Long-Term 

Protective Actions

Base Model

Source Term Site File
4-73

population subject to long-term protective actions is 33% larger for case 50. The consequences 

are generally but not always higher for case 48 because of its slightly larger source term 

magnitude and slightly longer release duration. 

Table 4-37 Comparison of source term environmental release timing for Mark I Case 48 

and Mark I Case 50

4.4.4 Evacuation Sensitivity Calculations for Fastest Release (Mark I Case 49)

As discussed above, an evacuation delay of 1 hour applied uniformly to all evacuation cohorts has 

zero effect on the results because the releases begin late enough that the surrounding population 

is expected to still have time to evacuate. Additional calculations were performed to assess the 

sensitivity of extended delays in evacuation assumptions on the individual early fatality and ILCF 

risk measures. The calculations used the accident progression and source term data from Mark I 

MELCOR case 49, which had the fastest environmental release of all scenarios analyzed. This 

scenario assumes a short-term station blackout with no batteries available and no RCIC at a BWR 

with a Mark I containment. In this case, the release to the environment starts at 7.3 hours and 

releases 0.5% of the cesium and 1.7% of the iodine. The MACCS calculations were run for four 

variations of the scenario, each with and without an external filter. The cases with an external filter 

assumed a decontamination factor of 10 for all particle sizes. The four variations are as follows:

• Base Case: This uses the expected evacuation timing for Peach Bottom based on the

ETE report submitted to NRC in 2014. Based on the ETE data and emergency

declaration and notification assumptions, the EPZ would be cleared of evacuees in

about 5.5 hours.

• 3 Hour Delay: The evacuation timeline is delayed by 3 hours and is applied uniformly to

all evacuation cohorts.

• 6 Hour Delay: The evacuation timeline is delayed by 6 hours and is applied uniformly to

all evacuation cohorts.

• No Evacuation: A hypothetical situation in which the EPZ population does not evacuate

and instead shelters in place. The MACCS evacuation model was turned off and

shielding parameters were adjusted to simulate sheltering.

These sensitivity calculations are designed to simulate intentionally unrealistic emergency 

response situations. Emergency response programs are developed, tested, and evaluated by the 

NRC as an element of defense-in-depth. Detailed plans for onsite and offsite responses are 

approved by NRC and FEMA, and it is expected that the plans will be implemented as written. 

Individual Early 

Fatality Risk

0-1.3 mi and beyond 0-10 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi

6 48 0.23% 1.69% 11.4 8 0 7.95E-05 1.61E-05 7.79E-06 253,000 450,000

Test 50 0.21% 1.09% 7.3 6 0 7.11E-05 1.42E-05 6.75E-06 204,000 354,000

0-50 mi 0-100 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi

6 48 0.23% 1.69% 11.4 8 1,150,000,000 1,390,000,000 116 175 3,440 3,440

Test 50 0.21% 1.09% 7.3 6 1,000,000,000 1,190,000,000 91 135 4,590 4,590

Individual Latent Cancer Fatality Risk Population Dose

Bin (person-rem) Rep 

Case

Rep 

Case Cs 

(%)

Rep Case I 

(%)

Start Time 

(hrs)

# Hrs with 

Significant 

Cs 

Release*

Offsite Cost ($ 2013)

Land (sq mi) 

Exceeding LongTerm Habitability 

Criterion

Population Subject to 

Long-Term Protective 

Actions Bin Rep 

Case

Rep 

Case Cs 

(%)

Rep Case I 

(%)

Start Time 

(hrs)

# Hrs with 

Significant 

Cs 

Release*
4-74

In addition, for all of the MACCS calculations, the protective actions for the intermediate phase 

and long-term phase were kept in full effect, so the model changes only affect the emergency 

phase which was modeled to last for one week.

The summary results for the conditional ILCF risk for the 0-10 mile area are shown in 

Figure 4-20, “CRF for Conditional ILCF Risk (0-10 mi) for Evacuation Sensitivity Calculations for 

BWR Mark I MELCOR Case 49.” For the base case in Figure 4-20, a scenario with an external 

filter results in a consequence reduction factor (CRF) of about 3 for the ILCF risk, conditional on 

the accident occurring. For the 3 hour evacuation delay in Figure 4-20, there is no change to the 

ILCF risk total and no change to the CRF for the external filter. In this case the release would start 

before the EPZ is evacuated; however, by the time the plume travels out to the more populated 

areas, the population would effectively have left the area. 

For the 6 hour evacuation delay in Figure 4-20, the total ILCF risk approximately doubles relative 

to the base case without an external filter. The contribution to the ILCF risk from the emergency 

phase increases significantly, by about a factor of 25; however, the emergency phase contribution 

was originally just 3%. Given that the contribution from the intermediate and long-term phase 

remains constant, the overall increase is fairly minor. For the case with an external filter, the 

increase in ILCF risk from the emergency phase is much smaller. The overall CRF increases to 

4.8 relative to the base case as shown in Figure 4-20. 

Figure 4-20 CRF for conditional ILCF risk (0-10 mi) for evacuation sensitivity calculations 

for BWR Mark I MELCOR Case 49
4-75

For the extremely unlikely and hypothetical situation in which the EPZ population does not 

evacuate and instead shelters in place, Figure 4-20 shows the total ILCF risk increases for each 

case (with and without the external filter) because of the increase to the emergency phase.

Without an external filter, the emergency phase contribution to the ILCF risk is about 50 times 

larger than for the base case; however, the total ILCF risk is only about 2.5 times larger than the 

base case. Overall, the CRF increases to 6.0.

The following are conclusions from the additional evacuation sensitivity calculations for BWR Mark 

I MELCOR case 49:

• Changes to the evacuation model show an increase in the potential benefit of an

external filter; however, the effect is relatively small (up to about a factor of two for a

CRF=6.0 compared to CRF=3.1).

• Assuming protective actions are taken in the intermediate and long-term phases, the

ILCF risk is maintained at a level well below the QHO, when multiplied by the accident

frequency.

• For all evacuation sensitivities for BWR Mark I MELCOR case 49, there is essentially

zero individual early fatality risk.

4.5 Conditional Offsite Consequences of Different CPRR Alternatives

The NRC has evaluated the conditional offsite consequences associated with the three major 

accident response strategies: (1) no post-core damage external water addition, (2) successful 

post-core damage external water addition, and (3) external water addition with an external filter.

For the Mark I analyses, water addition cases (SAWA and SAWM) are grouped together for 

consequence analysis purposes because there is little difference between them. For the Mark II 

analysis, the water management strategy does not apply because of the containment geometry.

All cases representing an external filter assume a uniform decontamination factor of 10.

For each alternative, a subset of MELCOR cases was selected which equally represent the 

alternative. The consequences of the MACCS source term bin corresponding to each of the 

applicable MELCOR cases are averaged to yield an overall set of offsite consequence results.

This process is shown in Table 4-38, “Average Mark I Conditional Offsite Consequences for the 

Different MELCOR Cases Associated with the CPRR Alternatives,” and Table 4-39, “Average 

Mark II Conditional Offsite Consequences for the Different MELCOR Cases Associated with the 

CPRR Alternatives,” for the Mark I and Mark II analysis, respectively. The average offsite 

consequences for each of the CPRR alternatives are then compared in Table 4-40, “Conditional 

Mark I and Mark II Offsite Consequences and Consequence Reduction Factor (CRF) for each 

CPRR Alternative.” Table 4-40 also shows the consequence reduction factor (CRF) associated 

with each alternative relative to the case with no external water addition. 
4-76

Table 4-38 Average Mark I conditional offsite consequences for the different MELCOR 

cases associated with the CPRR alternatives

MELCOR 

Case

MACCS 

Bin

Individual Latent 

Cancer Fatality 

Risk

(0-10 mi)

Population Dose 

(rem)

 (0-50 mi)

Offsite Cost 

($ 2013)

(0-50 mi)

Land 

Contamination

(sq. miles)

 (0-50 mi)

Population Subject 

to Long-Term 

Protective Actions

(0-50 mi)

1 12 2.91E-04 1,720,000 13,000,000,000 549 64,500

2 15 2.59E-04 1,740,000 15,700,000,000 573 111,000

4 10 4.06E-04 1,360,000 9,900,000,000 479 51,400

5 10 4.06E-04 1,360,000 9,900,000,000 479 51,400

6 12 2.91E-04 1,720,000 13,000,000,000 549 64,500

3.30E-04 1,600,000 12,000,000,000 530 69,000

8 11 1.35E-04 1,110,000 5,960,000,000 286 40,500

9 7 1.21E-04 524,000 2,740,000,000 190 15,000

10 7 1.21E-04 524,000 2,740,000,000 190 15,000

11 7 1.21E-04 524,000 2,740,000,000 190 15,000

12 11 1.35E-04 1,110,000 5,960,000,000 286 40,500

13 7 1.21E-04 524,000 2,740,000,000 190 15,000

14 7 1.21E-04 524,000 2,740,000,000 190 15,000

15 7 1.21E-04 524,000 2,740,000,000 190 15,000

16 7 1.21E-04 524,000 2,740,000,000 190 15,000

21 11 1.35E-04 1,110,000 5,960,000,000 286 40,500

22 12 2.91E-04 1,720,000 13,000,000,000 549 64,500

23 11 1.35E-04 1,110,000 5,960,000,000 286 40,500

25 7 1.21E-04 524,000 2,740,000,000 190 15,000

26 7 1.21E-04 524,000 2,740,000,000 190 15,000

28 7 1.21E-04 524,000 2,740,000,000 190 15,000

29 6 7.95E-05 253,000 1,150,000,000 116 3,440

30 7 1.21E-04 524,000 2,740,000,000 190 15,000

1.30E-04 720,000 4,000,000,000 230 23,000

8DF10 6 7.95E-05 253,000 1,150,000,000 116 3,440

9DF10 5 2.03E-05 71,200 220,000,000 41 118

10DF10 5 2.03E-05 71,200 220,000,000 41 118

11DF10 5 2.03E-05 71,200 220,000,000 41 118

12DF10 6 7.95E-05 253,000 1,150,000,000 116 3,440

13DF10 5 2.03E-05 71,200 220,000,000 41 118

14DF10 5 2.03E-05 71,200 220,000,000 41 118

15DF10 5 2.03E-05 71,200 220,000,000 41 118

16DF10 5 2.03E-05 71,200 220,000,000 41 118

21DF10 6 7.95E-05 253,000 1,150,000,000 116 3,440

22DF10 6 7.95E-05 253,000 1,150,000,000 116 3,440

23DF10 6 7.95E-05 253,000 1,150,000,000 116 3,440

25DF10 5 2.03E-05 71,200 220,000,000 41 118

26DF10 5 2.03E-05 71,200 220,000,000 41 118

28DF10 5 2.03E-05 71,200 220,000,000 41 118

29DF10 4 1.72E-05 48,400 141,000,000 23 7

30DF10 5 2.03E-05 71,200 220,000,000 41 118

3.80E-05 120,000 490,000,000 62 1,100

Successful 

External 

Water 

Addition & 

External 

Filter 

(DF=10)

Average:

Average:

No 

External 

Water 

Addition

Successful 

External 

Water 

Addition

Average:
4-77

Table 4-39 Average Mark II conditional offsite consequences for the different MELCOR 

cases associated with the CPRR alternatives

Table 4-40 Conditional Mark I and Mark II offsite consequences and consequence 

reduction factor (CRF) for each CPRR alternative

MELCOR 

Case

MACCS 

Bin

Individual Latent 

Cancer Fatality 

Risk

(0-10 mi)

Population 

Dose (rem)

 (0-50 mi)

Offsite Cost 

($ 2013)

(0-50 mi)

Land 

Contamination

(sq. miles)

 (0-50 mi)

Population Subject 

to Long-Term 

Protective Actions

(0-50 mi)

1 8 4.70E-04 6,110,000 85,500,000,000 854 721,000

5 6 2.29E-04 2,160,000 24,000,000,000 303 62,400

6 7 3.08E-04 4,140,000 80,800,000,000 698 619,000

3.40E-04 4,100,000 63,000,000,000 620 470,000

10 5 1.35E-04 689,000 4,250,000,000 130 15,400

11 4 7.90E-05 202,000 844,000,000 44 1,030

24 6 2.29E-04 2,160,000 24,000,000,000 303 62,400

1.50E-04 1,000,000 9,700,000,000 160 26,000

10DF10 4 7.90E-05 202,000 844,000,000 44 1,030

11DF10 3 6.58E-06 20,700 393,000,000 2 0

24DF10 4 7.90E-05 202,000 844,000,000 44 1,030

5.50E-05 140,000 690,000,000 30 690

No 

External 

Water 

Addition

Average:

Successful 

External 

Water 

Addition

Average:

Successful 

External 

Water 

Addition & 

External 

Filter 

(DF=10) Average:
4-78

Figure 4-21 and Figure 4-22 show graphical depictions of the offsite consequences of the CPRR 

alternatives as a percentage of the status quo.

Figure 4-21 Conditional Mark I offsite consequences for each CPRR alternative as a 

percentage of the case with no external water addition


)LJXUH &RQGLWLRQDO0DUN,,RIIVLWHFRQVHTXHQFHVIRUHDFK&355DOWHUQDWLYHDVD

SHUFHQWDJHRIWKHFDVHZLWKQRH[WHUQDOZDWHUDGGLWLRQ

7KHIROORZLQJJHQHUDOFRQFOXVLRQVFDQEHGUDZQIURPWKHWDEOHVDQGILJXUHVDERYHUHJDUGLQJWKH

FRQGLWLRQDORIIVLWHFRQVHTXHQFHVSHUHYHQWIRUWKH&355DOWHUQDWLYHV

x )RUWKH0DUN,DQG0DUN,,DQDO\VLVUHODWLYHWRFDVHVZLWKRXWH[WHUQDOZDWHUDGGLWLRQ

VXFFHVVIXOZDWHULQMHFWLRQ6$:$6$:0UHVXOWVLQDQRWDEOHUHGXFWLRQLQRIIVLWH

FRQVHTXHQFHVIRUDOOILYHPHDVXUHV,/&)ULVNSRSXODWLRQGRVHRIIVLWHFRVWODQG

FRQWDPLQDWLRQDQGSRSXODWLRQVXEMHFWWRORQJWHUPSKDVHSURWHFWLYHDFWLRQV

x )RUWKH0DUN,DQDO\VLVUHODWLYHWRFDVHVZLWKRXWH[WHUQDOZDWHUDGGLWLRQWKH

FRPELQDWLRQRIVXFFHVVIXOZDWHULQMHFWLRQZLWKDQH[WHUQDOILOWHUUHVXOWVLQDJUHDWHU

UHGXFWLRQRIRIIVLWHFRQVHTXHQFHV

x )RUERWKWKH0DUN,DQG0DUN,,DQDO\VLVWKHRIIVLWHFRQVHTXHQFHPHWULFWKDWVKRZVWKH

ODUJHVWUHGXFWLRQIDFWRULVWKHSRSXODWLRQVXEMHFWWRORQJWHUPSURWHFWLYHDFWLRQV

&RPSDUHGWRFDVHVZLWKRXWH[WHUQDOZDWHUDGGLWLRQVXFFHVVIXOZDWHUDGGLWLRQFRPELQHG

ZLWKDQH[WHUQDOILOWHUSURYLGHVDFRQVHTXHQFHUHGXFWLRQIDFWRURIDSSUR[LPDWHO\IRU

0DUN,DQGDSSUR[LPDWHO\IRU0DUN,,

x 7KH&355DOWHUQDWLYHVJHQHUDOO\VKRZJUHDWHUUHGXFWLRQLQRIIVLWHFRQVHTXHQFHVUHODWLYH

WRFDVHVZLWKRXWH[WHUQDOZDWHUDGGLWLRQIRUWKH0DUN,,DQDO\VLVFRPSDUHGWRWKH0DUN,

DQDO\VLV7KLVLVOLNHO\GXHWRWKHKLJKHUSRSXODWLRQDURXQGWKH0DUN,,VLWH/LPHULFN
4-80

• The conditional ILCF risk for the cases without external water addition, 3x10-4 per event

for both Mark I and Mark II analyses, is sufficiently low that when multiplied by the

accident frequency, there is substantial margin to the NRC Safety Goal (QHOs) for the

ILCF risk.

4.6 Offsite Consequence Summary and Conclusions

In support of the CPRR rulemaking, offsite consequence calculations were performed using the 

MACCS code. The modeling approach was to develop site-specific MACCS models for a 

reference Mark I site and a reference Mark II site based on the most current information available 

and the latest modeling best practices. Site-specific features of the model include weather, 

population, land use, economic values, emergency response characteristics and timing, and the 

long-term phase habitability criterion. Source terms, generated by MELCOR, were each 

considered in four variations to evaluate the potential range of external filter performance: with no 

external filter (DF=1) and with an external filter (DF=10, DF=100, and DF=1000). After applying 

the DFs, the source terms were binned according to cesium and iodine release magnitudes.

Individual source term cases were then selected to represent the 18 Mark I source term bins and 

9 Mark II source term bins.

The potential effectiveness of an external filter on reducing the environmental release is heavily 

influenced by the fraction of the source term that flows through the wetwell vent or drywell vent, 

where the external filter is attached. For accident cases in which the entire release flows through a 

vent pathway, the external filter can reduce the environmental release, and there is significant 

incremental benefit to increasing the filter DF. In contrast, for accident cases in which most of the 

release does not flow through the vent, perhaps due to drywell liner melt-through or a main steam 

line creep rupture, there is little benefit that can be achieved from an external filter. There are also 

many cases in between these, in which much, but not all of the release flows through a vent 

pathway. In these cases, an external filter with DF=10 can reduce the environmental release, 

However, there is little or no incremental benefit of having an external filter that can achieve a 

higher DF (100 or 1000). Note that these analyses do not consider the reliability of the external 

filter; each case assumes that it always performs at its specified DF for the release going through 

the filter. 

Results calculated and presented in this section include individual early fatality risk, individual LCF 

risk, total population dose, total offsite cost, land exceeding the long-term phase habitability 

criterion, and population subject to long-term phase protective actions. All results are mean values 

over approximately 1,000 weather trials and are presented as conditional (per event) on the 

accident scenario occurring (no frequency consideration).

For all Mark I and Mark II source terms, there is essentially zero individual early fatality risk, 

because even the largest source terms do not cause acute doses that exceed the dose threshold 

for acute radiation syndrome fatalities. All of the source terms begin their release to the 

environment long enough after the accident’s initiation to reasonably allow time for the EPZ 

population to evacuate. Thus the contribution to individual LCF risk and total population dose from 

the early phase is very low, especially for the 10-mile area. All other calculated consequences 

beside early fatality risk generally increase with source term magnitude and with source term 

duration. Larger source terms create more offsite contamination, leading to higher consequence 

results. Source terms that are released over a longer period of time allow more time for the wind 

to shift direction, thereby increasing the probability that contamination is spread over a larger area.

The larger consequences of a longer, more gradual release compared with a shorter, more 

punctuated release are more pronounced at shorter distances and less apparent at longer 
4-81

distances. This trend results from the fact that the exposures from narrower, more concentrated 

plume patterns provide significant exposures at longer distances, whereas less concentrated 

plumes that are spread over multiple compass sectors become depleted at shorter distances.

Sensitivity studies were conducted to assess the impact of many of the site-specific modeling 

features on the calculated results. These modeling features include site population and economic 

values, an evacuation delay, a larger nonevacuating population cohort, the intermediate phase 

duration, and the long-term habitability criterion. The sensitivity calculations lead to the following 

general insights:

• An evacuation delay of 1 hour applied uniformly to all cohorts has no effect on any of the

results.

• Baseline calculations use 0.5% or 2% for the nonevacuating cohort fraction. Sensitivities

using a larger nonevacuating cohort (5%) show a small (≤ 20%) increase to the

individual LCF risk for the 10-mile area.

• Baseline calculations use the Pennsylvania long-term phase habitability criterion of 500

mrem-per-year. Sensitivities using the less strict 2 rem-per-year habitability criterion

based on the EPA PAG Manual [18] generally show:

o higher individual LCF risk (~50-250% higher within 10 miles for the medium and

large source terms)

o higher population dose (up to ~50% higher within 50 miles for the medium and

large source terms)

o lower offsite cost (up to ~75% lower within 50 miles for the medium and large

source terms)

o lower land contamination (up to ~50% lower within 50 miles for the medium and

large source terms)

o lower population subject to long-term protective actions (up to ~95% lower within

50 miles for the medium and large source terms)

• Baseline calculations used 3 months as the duration of the intermediate phase.

Compared to baseline calculations, sensitivities using no intermediate phase generally

show:

o higher individual LCF risk for the small source term (up to ~15% higher within 10

miles) and lower individual LCF risk for the medium and large source terms

(decrease of ~4-17% within 10 miles)

o higher population dose (up to ~10% higher within 50 miles)

o lower offsite cost (up to ~45% lower within 50 miles for the medium and large

source terms)

o higher land contamination (up to ~80% higher within 50 miles for the medium and

large source terms)

o higher population subject to long-term protective actions (up to ~550% higher

within 50 miles for the medium and large source terms)

• Compared to baseline calculations, sensitivities using a 1-year intermediate phase

generally show:

o lower individual LCF risk for the small source term (~10% lower within 10 miles)

and higher individual LCF risk for the medium and large source terms (increase of

up to ~18% within 10 miles)
4-82

o lower population dose (up to ~7% lower within 50 miles)

o higher offsite cost (up to ~60% higher within 50 miles for the large source term)

o less land contamination (up to ~40% lower within 50 miles for the medium and

large source terms)

o lower population subject to long-term protective actions (up to ~80% lower within

50 miles for the medium and large source terms)

• Population sensitivity calculations were performed by using site file data for low,

medium, and high population Mark I and Mark II sites. Site files contain population, land

use, and economic values. Population dose, offsite cost, and population subject to longterm protective actions generally increase with population. Since the Mark I and Mark II

reference sites were the high population sites (Peach Bottom and Limerick), these three

result types were generally lower for the low and medium sites compared to those

reported in Section 4 of this report. Individual LCF risk is a population-weighted metric so

it does not scale as clearly with population; rather, it depends on population distributions

compared with wind rose probabilities. The area of land exceeding the long-term phase

habitability criterion also does not show any clear relationship with population and

economic data; rather, it depends on land fraction (land versus water) for the area

around the site.

• MACCS sensitivity calculations, which generally showed the greatest increase in

individual latent cancer fatality risk were those with medium and large releases in which

the EPA PAG habitability criterion (2 rem in the first year), were used instead of the

Pennsylvania habitability criterion (500 mrem in the first year), thus allowing a first year

dose up to four times higher. The increase in individual latent cancer fatality risk was

commonly a factor of 2-3 higher than for the base cases.

Overall, sensitivity calculations do not change any of the consequence analysis insights related to 

the QHO metrics. Individual early fatality risk remains essentially zero for all sensitivity 

calculations conducted, and ILCF risk remains well below the QHO, even those assuming a larger 

habitability criterion (e.g., 2 rem per year instead of 500 mrem per year).

4.7 References for Chapter 4

1. Electric Power Research Institute, EPRI Report 1025287, “Seismic Evaluation Guidance:

Screening, Prioritization and Implementation Details (SPID) for the Resolution of Fukushima

Near-Term Task Force Recommendation 2.1: Seismic,” November 2012, ADAMS Accession

No. ML12333A170.

2. U.S. Nuclear Regulatory Commission, NUREG/CR-4691, Vol. 2, “MELCOR Accident

Consequence Code System (MACCS): Model Description,” Washington, DC, February 1990.

3. U.S. Nuclear Regulatory Commission, NUREG/CR-6613, “Code Manual for MACCS2:

Volume 1, User’s Guide,” Washington, DC, May 1998.

4. U.S. Nuclear Regulatory Commission, Draft NUREG/CR, “WinMACCS, a MACCS2 Interface

for Calculating Health and Economic Consequences from Accidental Release of Radioactive

Materials into the Atmosphere: User’s Guide and Reference Manual for WinMACCS Version

3,” Washington, DC, December 2009.
4-83

5. U.S. Nuclear Regulatory Commission, NUREG/CR-6525, Rev. 1, SAND2003-1648P,

“SECPOP2000: Sector Population, Land Fraction, and Economic Estimation Program,”

Washington, DC, August 2003.

6. McFadden, Katherine et al, “MELMACCS Models Document,” Sigma Software L.L.C., Draft,

January 2011.

7. U.S. Nuclear Regulatory Commission, NUREG/BR-0058, Rev. 4, “Regulatory Analysis

Guidelines of the U.S. Nuclear Regulatory Commission,” Washington, DC, September 2004.

8. U.S. Nuclear Regulatory Commission, NUREG/BR-0184, “Regulatory Analysis Technical

Evaluation Handbook,” Washington, DC, January 1997.

9. U.S. Nuclear Regulatory Commission, NUREG-1935, “State-of-the-Art Reactor Consequence

Analyses (SOARCA) Report,” Washington, DC, November 2012.

10. U.S. Nuclear Regulatory Commission, NUREG/CR-7110, Vol. 1, Rev. 1, “State-of-the-Art

Reactor Consequence Analyses Project Volume 1: Peach Bottom Integrated Analysis,”

Washington, DC, May 2013.

11. U.S. Nuclear Regulatory Commission, NUREG/CR-7110, Vol. 2, Rev. 1, “State-of-the-Art

Reactor Consequence Analyses Project Volume 2: Surry Integrated Analysis,” Washington,

DC, August 2013.

12. U.S. Nuclear Regulatory Commission, NUREG/BR-0359, Rev. 1, “Modeling Potential Reactor

Accident Consequences,” Washington, DC, December 2012.

13. U.S. Nuclear Regulatory Commission, NUREG/CR-7009, “MACCS Best Practices as Applied

in the SOARCA Project,” Washington, DC, September 2014.

14. U.S. Nuclear Regulatory Commission, Draft NUREG/CR-7155, “State-of-the-Art Reactor

Consequence Analyses Project: Uncertainty Analysis of the Unmitigated Long-Term Station

Blackout of the Peach Bottom Atomic Power Station,” Washington, DC, June 2013.

15. U.S. Nuclear Regulatory Commission, NUREG-2161, “Consequence Study of a BeyondDesign-Basis Earthquake Affecting the Spent Fuel Pool for a U.S. Mark I Boiling Water

Reactor,” Washington, DC, September 2014.

16. U.S. Nuclear Regulatory Commission, SECY-12-0157, Enclosure 5B, “Consideration of

Additional Requirements for Containment Venting Systems for Boiling Water Reactors with

Mark I and Mark II Containments: MACCS Consequence Analysis,” Washington, DC,

December 2012.

17. U.S. Nuclear Regulatory Commission, SECY-12-0123, “Update on Staff Plans to Apply the

Full-Scope Site Level 3 PRA Project Results to the NRC’s Regulatory Framework,”

September 2012.

18. U.S. Environmental Protection Agency, EPA 402-R-92-001, “Manual of Protective Action

Guides and Protective Actions for Nuclear Incidents,” Washington, DC, May 1992.

19. U.S. Environmental Protection Agency, EPA 402-R-99-001, “Federal Guidance Report No.

13, Cancer Risk Coefficients for Environmental Exposure to Radionuclides,” Washington, DC,

September 1999.

20. U.S. Nuclear Regulatory Commission, NUREG-1350, Vol. 25, “Information Digest, 2013-

2014,” Washington, DC, August 2013.

21. OECD/NEA/CSNI Status Report on Filtered Containment Venting, NEA/CSNI/R(2014)7, July

2014, https://www.oecd-nea.org/nsd/docs/2014/csni-r2014-7.pdf
4-84

22. KLD TR-636, Rev. 0, “Development of Evacuation Time Estimates: Peach Bottom, Units, 1,

2, and 3,” April 2014, http://pbadupws.nrc.gov/docs/ML1414/ML14141A046.html.

23. U.S. Department of Labor, Bureau of Labor Statistics, http://www.bls.gov/cpi

24. U.S. Environmental Protection Agency, EPA-454/R-99-005, “Meteorological Monitoring

Guidance for Regulatory Modeling Applications,” Research Triangle Park, NC, February

2000.

25. U.S. Nuclear Regulatory Commission, NUREG-0917, “Nuclear Regulatory Commission Staff

Computer Programs for Use with Meteorological Data,” Washington, DC, July 1982.

26. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.23, Rev. 1, “Meteorological

Monitoring Programs for Nuclear Power Plants,” Washington, DC, March 2007.

27. U.S. Environmental Protection Agency, Support Center for Regulatory Atmospheric Modeling,

“SCRAM Surface Meteorological Archived Data: 1984-1992,”

http://www.epa.gov/ttn/scram/surfacemetdata.htm

28. U.S. Environmental Protection Agency, Office of Air Programs, AP-101, “Mixing Heights,

Wind Speeds, and Potential for Urban Air Pollution Throughout the Contiguous United

States,” Research Triangle Park, NC, January 1972.

29. Pasquill, F., “The Estimation of the Dispersion of Windborn Material,” The Meteorological

Magazine, 90:33-49, 1961.

30. Tadmor, J., and Y. Gur., “Analytical Expressions for the Vertical and Lateral Dispersion

Coefficients in Atmospheric Diffusion,” Atmospheric Environment, Vol. 3, pp. 688–689, 1969.

31. Eimutis E.C. and M.G. Konicek, “Derivations of Continuous Functions for the Lateral and

Vertical Atmospheric Dispersion Coefficients,” Atmospheric Environment, 6:859-63, 1972.

32. Napier, B.A., J.P. Rishel, and N.E. Bixler, PNNL-20990, “Final Review of Safety Assessment

Issues at Savannah River Site, August 2011,” Pacific Northwest National Laboratory,

December 2011.

33. U.S. Department of Agriculture, National Agricultural Statistics Service, CropScape 2013,

http://nassgeodata.gmu.edu/CropScape

34. U.S. Environmental Protection Agency, Multi-Resolution Land Characteristics Consortium,

National Land Cover Class Definitions, http://www.epa.gov/mrlc/definitions.html

35. Whelan, G., et al., DOE/RL/87-09, “The Remedial Action Priority System (RAPS):

Mathematical Formulations, U.S. Department of Energy, Washington, DC, August 1987.

36. U.S. Nuclear Regulatory Commission, NUREG/CR-7161, “Synthesis of Distributions

Representing Important Non-Site-Specific Parameters in Offsite Consequence Analyses,”

Washington, DC, April 2013.

37. U.S. Nuclear Regulatory Commission and Commission of European Communities,

NUREG/CR-6545, “Probabilistic Accident Consequence Uncertainty Analysis: Early Health

Effects Uncertainty Analysis,” Washington, DC, December 1997.

38. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.145, “Atmospheric Dispersion

Models for Potential Accident Consequence Assessments at Nuclear Power Plants,”

Washington, DC, November 1982.

39. U.S. Nuclear Regulatory Commission, NUREG/CR-2260, “Technical Basis for Regulatory

Guide 1.145, ‘Atmospheric Dispersion Models for Potential Accident Consequence

Assessments at Nuclear Power Plants,’” Washington, DC, October 1981.
4-85

40. U.S. Nuclear Regulatory Commission, NUREG-0654/FEMA-REP-1, Rev. 1, “Criteria for

Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in

Support of Nuclear Power Plants,” Washington, DC, November 1980.

41. U.S. Nuclear Regulatory Commission, NUREG/CR-7002, SAND2010-0016P, "Criteria for

Development of Evacuation Time Estimate Studies," Washington, DC, November 2011.

42. U.S. Nuclear Regulatory Commission, NUREG/CR-6953, Vol. 2, “Review of NUREG-0654,

Supplement 3, ‘Criteria for Protective Action Recommendations for Severe Accidents’: Focus

Groups and Telephone Survey”, Washington, DC, October 2008.

43. Wolshon, Brian, J. Jones, and F. Walton, “The Evacuation Tail and Its Effect on Evacuation

Decision Making," Journal of Emergency Management. Vol. 8, No. 1, January/February 2010.

44. U.S. Nuclear Regulatory Commission, NUREG/CR-6864, "Identification and Analysis of

Factors Affecting Emergency Evacuations," Washington, DC, January 2005.

45. KLD TR-617, Rev. 0, “Development of Evacuation Time Estimates: Limerick Units 1 and 2,”

January 2014, http://pbadupws.nrc.gov/docs/ML1404/ML14042A181.pdf.

46. U.S. Nuclear Regulatory Commission, NUREG-1150, “Severe Accident Risks: An

Assessment for Five U.S. Nuclear Power Plants,” Washington, DC, December 1990.

47. U.S. Nuclear Regulatory Commission, NUREG/CR-4551, Vol. 2, Rev. 1, Part 7, “Evaluation

of Severe Accident Risks: Quantification of Major Input Parameters: MACCS Input,”

Washington, DC, December 1990.

48. U.S. Environmental Protection Agency, “PAG Manual: Protective Action Guides and Planning

Guidance for Radiological Incidents: Draft for Interim Use and Public Comment,” Washington,

DC, March 2013.

49. U.S. Nuclear Regulatory Commission, NUREG/CR-6981, “Assessment of Emergency

Response Planning and Implementation for Large Scale Evacuations,” Washington, DC,

October 2008.

50. Mileti, Dennis S., John H. Sorensen and Paul W. O’Brien, “Toward an Explanation of Mass

Care Shelter Use in Evacuations,” International Journal of Mass Emergencies and Disasters.

10 (1): 25-42, 1992.

51. NHK News, “Earthquake Report – JAIF,” No. 95: 18:00, May 28, 2011.

http://www.jaif.or.jp/nglish/news_images/pdf/ENGNEWS01_1306574680P.pdf

52. Japan Ministry of the Environment, “Progress on Off-site Cleanup Efforts in Japan,” April

2014.

53. U.S. Department of Labor, “U.S. Bureau of Labor Statistics Occupational Outlook Handbook,”

January 2014. http://www.bls.gov/ooh

54. S&P Case-Shiller National Home Price Index, http://www.us.spindices.com/index-family/realestate/sp-case-shiller-us-national-home-price-index

55. National Council of Real Estate Investment Fiduciaries, U.S. National Property Index,

https://www.ncreif.org/property-index-returns.aspx

56. Davis, M.A. and J. Heathcote, “The Price and Quantity of Residential Land in the United

States,” Journal of Monetary Economics, 54 (2595-2620), June 2007.

57. U.S. Nuclear Regulatory Commission, NUREG/CR-4214, “Health Effects Model for Nuclear

Power Plant Accident Consequence Analysis,” Washington, DC, July 1985.
4-86

58. Eckerman, K.F., Letter Report, “Radiation Dose and Health Risk Estimation: Technincal Basis

for the State-of-the-Art Reactor Consequence Analysis Project,” ML12159A259, Oak Ridge

National Laboratory, Oak Ridge, TN, January 2012.

59. National Academy of Sciences, “Health Effects of Exposure to Low Levels of Ionizing

Radiation: BEIR V,” National Research Council, National Academy Press, Washington, DC,

1990.

60. International Commission on Radiological Protection, ICRP Publication 60, “1990

Recommendations of the International Commission on Radiological Protection,” 1991.

61. 51 FR 30028, “Safety Goals for the Operation of Nuclear Power Plants,” August 1986.
5-1

5 RISK INTEGRATION

5.1 Results

Table 5-1 provides the main results of the risk integration, the risks estimates of each regulatory 

analysis sub-alternative. These results are graphically depicted in Figure 5-1 through Figure 5-5. 

The similarity among these figures implies that risk insights developed by comparing regulatory 

analysis sub-alternatives do not strongly depend on which risk metric is used to make the 

comparison.

The estimates of individual early fatality risk for all sub-alternatives are essentially zero since the 

conditional individual early fatality risks developed by the offsite consequence analysis are 

calculated as zero for all cases analyzed.

The risk estimates (prompt or early fatality risk and latent cancer fatality risk) show a reduction 

in risk when post-accident water injection is used. The risks of sub-alternatives 1 and 2A are 

larger than any of the other sub-alternatives since sub-alternatives 1 and 2A cannot prevent 

containment structural failure caused by liner melt-through. As discussed in Section 2, it 

became apparent that, as the project was nearing completion and SECY-15-0085 was being 

developed, all licensees intended to comply with Phase 2 of Order EA-13-109 by implementing 

the post-accident water injection strategy. 

The risk estimates show relatively minor decreases when post-accident water injection is made 

to the RPV (sub-alternatives 3A, 4Ai(1), 4Aii(1) and 4Aiii(1)) as compared to injection made to 

the DW (sub-alternatives 3B, 4Ai(2), 4Aii(2) and 4Aiii(2)). Sub alternatives involving postaccident injection to the RPV provide the capability to arrest a severe accident before vessel 

breach occurs.
5-2

Table 5-1 Risk estimates by regulatory analysis sub-alternative 

Index

Regulatory Analysis

Sub-Alternative

Fraction of CoreDamage 

Frequency

Individual 

Early 

Fatality

Risk (/y)

Individual Latent Cancer

Fatality Risk (/y)

Population Dose 

(person-rem/y)

Offsite Cost

($ 2013/y)

Land Exceeding LongTerm Habitability 

Criterion (square 

miles/y)

Population Subject to 

Long-Term Protective 

Actions

(persons/y)

Vented

Uncontrolled 

Release

0-1.3 mi

and beyond

0-10 mi

0-50 mi

0-100 mi

0-50 mi

0-100 mi

0-50 mi

0-100 mi

0-50 mi

0-100 mi

0-50 mi

0-100 mi

1 1 0% 100% 0.0E+00 3.0E-09 8.6E-10 4.2E-10 1.3E+01 2.3E+01 9.9E+04 1.3E+05 4.4E-03 7.6E-03 5.1E-01 5.8E-01

2 2A 0% 100% 0.0E+00 3.0E-09 8.6E-10 4.2E-10 1.3E+01 2.3E+01 9.9E+04 1.3E+05 4.4E-03 7.6E-03 5.1E-01 5.8E-01

3 3A 58% 42% 0.0E+00 1.8E-09 5.5E-10 2.7E-10 8.6E+00 1.5E+01 6.5E+04 8.5E+04 2.9E-03 5.0E-03 3.3E-01 3.9E-01

4 3B 42% 58% 0.0E+00 2.1E-09 6.7E-10 3.4E-10 1.1E+01 1.9E+01 7.4E+04 1.0E+05 3.4E-03 6.4E-03 4.1E-01 4.9E-01

5 4Ai(1) 58% 42% 0.0E+00 1.8E-09 5.5E-10 2.7E-10 8.6E+00 1.5E+01 6.5E+04 8.5E+04 2.9E-03 5.0E-03 3.3E-01 3.9E-01

6 4Ai(2) 42% 58% 0.0E+00 2.1E-09 6.1E-10 3.1E-10 9.5E+00 1.7E+01 6.8E+04 9.0E+04 3.2E-03 5.8E-03 3.6E-01 4.1E-01

7 4Aii(1) 58% 42% 0.0E+00 1.8E-09 5.5E-10 2.7E-10 8.6E+00 1.5E+01 6.5E+04 8.5E+04 2.9E-03 5.0E-03 3.3E-01 3.9E-01

8 4Aii(2) 42% 58% 0.0E+00 2.4E-09 7.7E-10 3.9E-10 1.2E+01 2.2E+01 8.9E+04 1.2E+05 3.9E-03 7.3E-03 4.8E-01 5.8E-01

9 4Aiii(1) 58% 42% 0.0E+00 1.8E-09 5.5E-10 2.7E-10 8.6E+00 1.5E+01 6.5E+04 8.5E+04 2.9E-03 5.0E-03 3.3E-01 3.9E-01

10 4Aiii(2) 42% 58% 0.0E+00 2.0E-09 5.6E-10 2.7E-10 8.7E+00 1.5E+01 6.2E+04 7.9E+04 3.0E-03 5.1E-03 3.1E-01 3.4E-01

11 4Bi(1) 58% 42% 0.0E+00 1.3E-09 3.1E-10 1.5E-10 4.5E+00 7.8E+00 2.9E+04 3.7E+04 1.6E-03 2.5E-03 1.5E-01 1.6E-01

12 4Bi(2) 42% 58% 0.0E+00 1.4E-09 3.3E-10 1.5E-10 4.8E+00 8.2E+00 3.1E+04 3.8E+04 1.8E-03 2.7E-03 1.6E-01 1.6E-01

13 4Bii 42% 58% 0.0E+00 1.4E-09 3.2E-10 1.5E-10 4.6E+00 7.9E+00 3.0E+04 3.7E+04 1.7E-03 2.6E-03 1.5E-01 1.5E-01

14 4Biii 42% 58% 0.0E+00 1.4E-09 3.2E-10 1.5E-10 4.7E+00 8.1E+00 3.1E+04 3.7E+04 1.7E-03 2.6E-03 1.5E-01 1.6E-01

15 4Biv 40% 60% 0.0E+00 1.3E-09 3.1E-10 1.5E-10 4.6E+00 7.8E+00 3.0E+04 3.6E+04 1.7E-03 2.6E-03 1.5E-01 1.5E-01

16 4Ci(1) 58% 42% 0.0E+00 1.3E-09 3.1E-10 1.5E-10 4.5E+00 7.8E+00 2.9E+04 3.7E+04 1.6E-03 2.5E-03 1.5E-01 1.6E-01

17 4Ci(2) 42% 58% 0.0E+00 1.3E-09 3.1E-10 1.4E-10 4.5E+00 7.6E+00 3.0E+04 3.7E+04 1.6E-03 2.4E-03 1.5E-01 1.6E-01

18 4Cii 42% 58% 0.0E+00 1.3E-09 3.0E-10 1.4E-10 4.4E+00 7.4E+00 2.9E+04 3.6E+04 1.5E-03 2.3E-03 1.5E-01 1.5E-0119 4Ciii 42% 58% 0.0E+00 1.3E-09 3.1E-10 1.4E-10 4.4E+00 7.6E+00 3.0E+04 3.7E+04 1.6E-03 2.4E-03 1.5E-01 1.6E-01

20 4Civ 40% 60% 0.0E+00 1.3E-09 3.0E-10 1.4E-10 4.3E+00 7.4E+00 2.9E+04 3.6E+04 1.5E-03 2.3E-03 1.5E-01 1.5E-01


)LJXUH &RPSDULVRQRIUHJXODWRU\DQDO\VLVVXEDOWHUQDWLYHVXVLQJLQGLYLGXDOODWHQW

FDQFHUIDWDOLW\ULVNPLOHV

)LJXUH &RPSDULVRQRIUHJXODWRU\DQDO\VLVVXEDOWHUQDWLYHVXVLQJSRSXODWLRQGRVHULVN

PLOHV


)LJXUH &RPSDULVRQRIUHJXODWRU\DQDO\VLVVXEDOWHUQDWLYHVXVLQJRIIVLWHFRVW

ULVN±PLOHV

)LJXUH &RPSDULVRQRIUHJXODWRU\DQDO\VLVVXEDOWHUQDWLYHVXVLQJDUHDRIODQG

H[FHHGLQJORQJWHUPKDELWDELOLW\FULWHULRQPLOHV


)LJXUH &RPSDULVRQRIUHJXODWRU\DQDO\VLVVXEDOWHUQDWLYHVXVLQJSRSXODWLRQVXEMHFW

WRORQJWHUPKDELWDELOLW\FULWHULRQPLOHV

7KHULVNHVWLPDWHVVKRZQRGLVFHUQDEOHUHGXFWLRQLQULVNZKHQILOWUDWLRQVWUDWHJLHVFRQWDLQPHQW

YHQWLQJF\FOLQJDQGVHYHUHDFFLGHQWZDWHUPDQDJHPHQWDUHXVHG7KLVREVHUYDWLRQLVEDVHGRQ

FRPSDULVRQRIVXEDOWHUQDWLYHV$L$LLDQG$LLLZLWKVXEDOWHUQDWLYH$DQGRQ

FRPSDULVRQRIVXEDOWHUQDWLYHV$L$LLDQG$LLZLWKVXEDOWHUQDWLYH%

7KHULVNHVWLPDWHVVKRZDQRWLFHDEOHUHGXFWLRQLQULVNZKHQHQJLQHHUHGILOWHUVDUHXVHG

6SHFLILFDOO\WKHULVNHVWLPDWHVIRUVXEDOWHUQDWLYHV%L%L%LL%LLL%LY&L&LL&LLL

DQG&LYDUHQRWLFHDEO\VPDOOHUDVFRPSDUHGWRWKHULVNHVWLPDWHVRIWKHRWKHUVXEDOWHUQDWLYHV

+RZHYHUWKHUHLVQRDSSDUHQWEHQHILWWRXVLQJODUJHHQJLQHHUHGILOWHUVZKLFKKDYHDQDVVXPHG

GHFRQWDPLQDWLRQIDFWRURIDVRSSRVHGWRXVLQJVPDOOHQJLQHHUHGILOWHUVZKLFKKDYHDQ

DVVXPHGGHFRQWDPLQDWLRQIDFWRURI6SHFLILFDOO\WKHULVNHVWLPDWHVVKRZQRGLVFHUQDEOH

UHGXFWLRQLQULVNIRUVXEDOWHUQDWLYHV&L&L&LL&LLLDQG&LYDVFRPSDUHGWR

VXEDOWHUQDWLYHV%L%L%LL%LLLDQG%LY

6HQVLWLYLW\DQG8QFHUWDLQW\$QDO\VLV

,QRUGHUWRGHYHORSDGGLWLRQDOULVNLQVLJKWVDERXWWKH&355VWUDWHJLHVWZRVXSSOHPHQWDODQDO\VHV

ZHUHSHUIRUPHG)LUVWWKHVHQVLWLYLW\RIFRUHGDPDJHIUHTXHQF\WRWKHVFRSLQJKXPDQHUURU

SUREDELOLWLHVZDVH[SORUHG6HFRQGDQDSSUR[LPDWHSDUDPHWULFXQFHUWDLQW\DQDO\VLVRIWKH

LQGLYLGXDOODWHQWFDQFHUIDWDOLW\ULVNHVWLPDWHVZDVFRQGXFWHG(DFKRIWKHVHVXSSOHPHQWDO

DQDO\VHVLVGLVFXVVHGLQWKHIROORZLQJVHFWLRQV
5-6

5.2.1 Sensitivity of Core Damage Frequency to Human Error Probabilities

To gain further perspective on the importance of operator actions and the adequacy of the 

scoping HEPs used in the accident sequence analysis, an en masse sensitivity analysis was 

performed. The en masse sensitivity analysis was conducted by simultaneously varying the HEPs 

of all operators contained in the CDETs. The results of this analysis, depicted as “heat maps,” are 

shown in Figure 5-6, Figure 5-7 and Figure 5-8. Note that the lowest core-damage frequency 

appears in the upper left corner of the heat map. The core-damage frequency increases as the incontrol room HEPs increase (towards the bottom of the heat map) and as the ex-control HEPs 

increase (towards the righthand side of the heat map). The highest core-damage frequency 

appears in the lower right corner of the heat map.

Figure 5-6 En masse sensitivity of core-damage frequency to human error probability for 

Wetwell-First venting strategy

Figure 5-7 En Masse sensitivity of core-damage frequency to human error probability for 

Drywell-First (passive actuation) venting strategy

0 0.001 0.002 0.003 0.006 0.01 0.02 0.03 0.06 0.1 0.2 0.3 0.6 1

0 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.2E-06 5.3E-06 5.6E-06 6.7E-06 8.5E-06 1.5E-05 1.9E-05

0.001 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.2E-06 5.3E-06 5.6E-06 6.7E-06 8.5E-06 1.5E-05 1.9E-05

0.002 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.2E-06 5.3E-06 5.6E-06 6.7E-06 8.5E-06 1.5E-05 1.9E-05

0.003 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.2E-06 5.3E-06 5.6E-06 6.7E-06 8.5E-06 1.5E-05 1.9E-05

0.006 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.2E-06 5.3E-06 5.6E-06 6.8E-06 8.5E-06 1.5E-05 1.9E-05

0.01 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.2E-06 5.3E-06 5.6E-06 6.8E-06 8.5E-06 1.5E-05 1.9E-05

0.02 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.2E-06 5.3E-06 5.6E-06 6.8E-06 8.6E-06 1.5E-05 1.9E-05

0.03 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.2E-06 5.3E-06 5.6E-06 6.9E-06 8.6E-06 1.5E-05 1.9E-05

0.06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.2E-06 5.2E-06 5.2E-06 5.4E-06 5.7E-06 7.0E-06 8.7E-06 1.5E-05 1.9E-05

0.1 5.2E-06 5.2E-06 5.2E-06 5.2E-06 5.2E-06 5.2E-06 5.3E-06 5.3E-06 5.5E-06 5.9E-06 7.2E-06 8.9E-06 1.5E-05 1.9E-05

0.2 5.5E-06 5.5E-06 5.5E-06 5.5E-06 5.5E-06 5.5E-06 5.6E-06 5.7E-06 5.9E-06 6.3E-06 7.6E-06 9.3E-06 1.5E-05 1.9E-05

0.3 6.0E-06 6.0E-06 6.0E-06 6.0E-06 6.0E-06 6.0E-06 6.1E-06 6.2E-06 6.4E-06 6.8E-06 8.1E-06 9.8E-06 1.6E-05 1.9E-05

0.6 8.7E-06 8.7E-06 8.7E-06 8.7E-06 8.7E-06 8.7E-06 8.7E-06 8.8E-06 8.9E-06 9.2E-06 1.0E-05 1.1E-05 1.6E-05 1.9E-05

1 1.5E-05 1.5E-05 1.5E-05 1.5E-05 1.5E-05 1.5E-05 1.5E-05 1.5E-05 1.5E-05 1.4E-05 1.4E-05 1.5E-05 1.7E-05 1.9E-05

Ex-Control_Room HEP

In-Control-Room HEP

0 0.001 0.002 0.003 0.006 0.01 0.02 0.03 0.06 0.1 0.2 0.3 0.6 1

0 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.3E-06 5.5E-06 6.5E-06 8.0E-06 1.4E-05 1.9E-05

0.001 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.3E-06 5.5E-06 6.5E-06 8.0E-06 1.4E-05 1.9E-05

0.002 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.3E-06 5.5E-06 6.5E-06 8.0E-06 1.4E-05 1.9E-05

0.003 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.2E-06 5.3E-06 5.5E-06 6.5E-06 8.0E-06 1.4E-05 1.9E-05

0.006 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.2E-06 5.3E-06 5.5E-06 6.6E-06 8.0E-06 1.4E-05 1.9E-05

0.01 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.2E-06 5.3E-06 5.5E-06 6.6E-06 8.1E-06 1.4E-05 1.9E-05

0.02 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.2E-06 5.3E-06 5.6E-06 6.6E-06 8.1E-06 1.4E-05 1.9E-05

0.03 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.2E-06 5.3E-06 5.6E-06 6.6E-06 8.1E-06 1.4E-05 1.9E-05

0.06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.2E-06 5.2E-06 5.4E-06 5.7E-06 6.8E-06 8.3E-06 1.4E-05 1.9E-05

0.1 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.2E-06 5.2E-06 5.4E-06 5.7E-06 6.9E-06 8.4E-06 1.4E-05 1.9E-05

0.2 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.2E-06 5.2E-06 5.3E-06 5.5E-06 5.9E-06 7.1E-06 8.7E-06 1.4E-05 1.9E-05

0.3 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.2E-06 5.2E-06 5.3E-06 5.3E-06 5.6E-06 6.0E-06 7.3E-06 8.8E-06 1.4E-05 1.9E-05

0.6 5.1E-06 5.2E-06 5.2E-06 5.2E-06 5.2E-06 5.2E-06 5.3E-06 5.4E-06 5.7E-06 6.2E-06 7.6E-06 9.1E-06 1.4E-05 1.9E-05

1 5.2E-06 5.2E-06 5.2E-06 5.2E-06 5.3E-06 5.3E-06 5.4E-06 5.5E-06 5.9E-06 6.4E-06 7.8E-06 9.3E-06 1.5E-05 1.9E-05

Ex-Control_Room HEP

In-Control-Room HEP
5-7

Figure 5-8 En Masse sensitivity of core-damage frequency to human error probability for 

Drywell-First (manual actuation) venting strategy

The results of the en masse sensitivity analysis show that CDF ranges from a low of 5.1x10-6/ry 

(assuming no operator errors) to a high of 1.9x10-5/ry (giving no credit to the plant operators). Two 

conclusions may be drawn from these results:

• About 60% of the total CDF is due to accident sequences that only involve equipment

failure.

• Assuming the HEP scoping values are appropriate, the CDF approximately doubles

when operators completely fail to take beneficial actions (1.9E-5 / 8.9E-6 = 2.1).

5.2.2 Parametric Uncertainty Analysis

Monte Carlo methods were used to conduct an approximate parametric uncertainty analysis of the 

individual latent cancer fatality risk estimates for each regulatory analysis sub-alternative. 

Parametric uncertainties were considered for all inputs, as explained in Table 5-2: 

Table 5-2 Uncertainty analysis inputs

0 0.001 0.002 0.003 0.006 0.01 0.02 0.03 0.06 0.1 0.2 0.3 0.6 1

0 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.2E-06 5.3E-06 5.6E-06 6.7E-06 8.5E-06 1.5E-05 1.9E-05

0.001 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.2E-06 5.3E-06 5.6E-06 6.7E-06 8.5E-06 1.5E-05 1.9E-05

0.002 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.2E-06 5.3E-06 5.6E-06 6.7E-06 8.5E-06 1.5E-05 1.9E-05

0.003 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.2E-06 5.3E-06 5.6E-06 6.7E-06 8.5E-06 1.5E-05 1.9E-05

0.006 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.2E-06 5.3E-06 5.6E-06 6.8E-06 8.5E-06 1.5E-05 1.9E-05

0.01 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.2E-06 5.3E-06 5.6E-06 6.8E-06 8.5E-06 1.5E-05 1.9E-05

0.02 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.2E-06 5.3E-06 5.6E-06 6.8E-06 8.6E-06 1.5E-05 1.9E-05

0.03 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.2E-06 5.3E-06 5.6E-06 6.9E-06 8.6E-06 1.5E-05 1.9E-05

0.06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.1E-06 5.2E-06 5.2E-06 5.2E-06 5.4E-06 5.7E-06 7.0E-06 8.7E-06 1.5E-05 1.9E-05

0.1 5.2E-06 5.2E-06 5.2E-06 5.2E-06 5.2E-06 5.2E-06 5.3E-06 5.3E-06 5.5E-06 5.9E-06 7.2E-06 8.9E-06 1.5E-05 1.9E-05

0.2 5.5E-06 5.5E-06 5.5E-06 5.5E-06 5.5E-06 5.5E-06 5.6E-06 5.7E-06 5.9E-06 6.3E-06 7.6E-06 9.3E-06 1.5E-05 1.9E-05

0.3 6.0E-06 6.0E-06 6.0E-06 6.0E-06 6.0E-06 6.0E-06 6.1E-06 6.2E-06 6.4E-06 6.8E-06 8.1E-06 9.8E-06 1.6E-05 1.9E-05

0.6 8.7E-06 8.7E-06 8.7E-06 8.7E-06 8.7E-06 8.7E-06 8.7E-06 8.8E-06 8.9E-06 9.2E-06 1.0E-05 1.1E-05 1.6E-05 1.9E-05

1 1.5E-05 1.5E-05 1.5E-05 1.5E-05 1.5E-05 1.5E-05 1.5E-05 1.5E-05 1.5E-05 1.4E-05 1.4E-05 1.5E-05 1.7E-05 1.9E-05

Ex-Control_Room HEP

In-Control-Room HEP

Events Distribution Remarks 

Frequency of 

ELAPs due to 

internal events 

Log-normal 

Mean = point estimate 

Error factor =15 

An error factor of 15 maximizes the ratio of the 95th 

percentile to the mean value. This approach does not 

explicitly consider the uncertainty in the offsite power 

recovery curves or the uncertainty in the EPS reliability 

parameters (failure rate and failure-on-demand 

probability) 

Seismic hazard 

curves 

Log-normal 

Normal parameters were developed for each point on 

the seismic hazard curve using the fractile information 

provided by licensees in their responses to the 10 CFR 

50.54(f) information request concerning NNTF 

Recommendation 2.1 

Seismic fragilities 

Double log-normal, 

using the developed 

values of C50, βR, and 

βU

Traditional approach to modeling uncertainty in seismic 

fragility. 

Hardware-related 

failures

Log-normal 

Mean = point estimate 

Error factor = 15 

An error factor of 15 maximizes the ratio of the 95th

percentile to the mean value. 


+XPDQIDLOXUH

HYHQWV

&RQVWUDLQHG

QRQLQIRUPDWLYHSULRU

$FRQVWUDLQHGQRQLQIRUPDWLYHSULRUGLVWULEXWLRQLVDEHWD

GLVWULEXWLRQZLWKPHDQ SRLQWHVWLPDWHDQGĮ 

&RQGLWLRQDO

FRQVHTXHQFHV

/RJQRUPDO

0HDQ SRLQWHVWLPDWH

(UURUIDFWRU 

,QIRUPHGE\SUHOLPLQDU\UHVXOWVRIWKH62$5&$

XQFHUWDLQW\DQDO\VLVSURMHFW>@

,QHVVHQFHWKHSDUDPHWULFXQFHUWDLQW\DQDO\VLVFRQVLVWVRIUHFDOFXODWLQJWKHORJLFPRGHOXVLQJD

UDQGRPVDPSOHRISDUDPHWHUYDOXHV(DFK&'(7ZDVUHTXDQWLILHGWLPHVIRUHDFKSODQW

ZLWKLQWKHVFRSHRIWKHDFFLGHQWVHTXHQFHDQDO\VLVDQGWKHUHVXOWLQJVHPLSODQWVSHFLILFUDQGRP

VDPSOHVRI3'6IUHTXHQFLHVZHUHDYHUDJHGWRGHYHORSUDQGRPVDPSOHVRIWKHIOHHW

DYHUDJH3'6IUHTXHQFLHV(DFK$3(7ZDVUHFDOFXODWHGWLPHVIRUHDFKLQSXW3'6WR

GHWHUPLQHUDQGRPVDPSOHVRIWKH5&IUHTXHQFLHV7KHUDQGRPVDPSOHVRI5&IUHTXHQFLHV

ZHUHFRPELQHGZLWKUDQGRPVDPSOHVRIWKHFRQGLWLRQDOFRQVHTXHQFHVWRGHYHORS

UDQGRPVDPSOHVRIWKHLQGLYLGXDOODWHQWFDQFHUIDWDOLW\ULVNIRUHDFKUHJXODWRU\DQDO\VLV

VXEDOWHUQDWLYH7KHUHVXOWVRIWKHSDUDPHWULFXQFHUWDLQW\DQDO\VLVDUHLOOXVWUDWHGLQ)LJXUH,W

VKRXOGEHQRWHGWKDWLQFRQWUDVWWRWKHXQFHUWDLQW\DQDO\VLVUHVXOWVSUHVHQWHGLQWKHGUDIW

UHJXODWRU\EDVLVDWWDFKHGWR6(&<WKHUHJXODWRU\DQDO\VLVVXEDOWHUQDWLYHVLQ)LJXUH

PDWFKWKHRUGHULQJXVHGLQ7DEOH

)LJXUH 5HVXOWVRIWKHSDUDPHWULFXQFHUWDLQW\DQDO\VLV

Events Distribution Remarks 
5-9

Uncertainty in the individual latent cancer fatality risk estimates ranges about two orders of 

magnitude. The major source of uncertainty is due to uncertainty in the seismic hazard curves, a 

result that is consistent with previous seismic PRAs.

As shown in Figure 5-9, the changes in risk among the various CPRR sub-alternatives are much 

smaller than the estimated parametric uncertainty ranges. This result does not necessarily imply 

that differences among the CPRR sub-alternatives are insignificant. The parametric uncertainty 

risk estimates of the various DPRR sub-alternatives are stochastically dependent (correlated) 

because they share a common database; as the correlation increases, the uncertainty in the 

changes in risk decreases. It should be noted that the impact evaluation contained in the draft 

regulatory basis attached to SECY-15-0085 uses the mean difference between sub-alternatives.

As shown in Figure 5-9, the mean risk estimate for each CPRR sub-alternative is approximately 

1,000 times lower than the QHO, and the 95th percentiles risk estimate for each CPRR 

sub-alternative is more than 100 times lower than the QHO. As a result, it may concluded that the 

risk due to ELAP events is very small, regardless of which specific CPRR regulatory analysis 

sub-alternative is implemented.

5.3 Conclusions

The risk integration, which was performed to support the draft regulatory basis attached to 

SECY-15-0085, assessed the risk impacts of twenty potential CPRR strategies for severe 

accidents initiated by ELAP events due to internal events and seismic events occurring at 

operating BWRs with Mark I containment designs and reactor core isolation cooling (RCIC) 

systems. Based on the results of the risk integration, the following conclusions were developed:

1. The risk due to ELAP events is very small, regardless of which specific CPRR regulatory

analysis sub alternative is implemented.

2. FLEX strategies reduce the probability of core damage due to ELAP events by more

than 50%.

3. The contribution of ELAPs initiated by seismic events is generally greater than the

contribution of ELAPs initiated by internal events.

4. The most important contributors to CDF from ELAPs are seismic failures of the station

batteries or dc switchgear, seismic failures of the emergency diesel generators (EDGs)

and their supporting systems, random failures of the portable FLEX pump (failure to start

and failure to run), and random failures of the RCIC pump (failure to start).

5. The capability to vent the containment during a severe accident limits the probability of

containment structural failure due to overpressurization to a few percent.

6. There is a reduction in risk when containment liner melt-through is prevented through the

use of post-accident water injection. Injection to the RPV somewhat lowers the risk as

compared to injection directly to the DW.

7. There is no discernable reduction in risk relative to sub-alternatives 3A and 3B (the

baseline for the impact evaluation provided in the draft regulatory evauation attached to

SECY-15-0085) when different filtration strategies (containment vent cycling and severe

accident water management) are considered.
5-10

8. There is a reduction in risk (for many of the accident sequences evaluated) when

engineered filters are used. However, there is no apparent benefit to using large type

(DF = 1000) engineered filters as opposed to using small type (DF = 10) engineered

filters.

9. About 60% of the total CDF is due to accident sequences that only involve equipment

failures. About 40% of the total CDF is due to accident sequences that involve

combinations of equipment failures and operator actions.

5.4 References for Chapter 5

1. U.S. Nuclear Regulatory Commission, NUREG/CR-7155 (draft), “State-of-the-Art Reactor

Consequence Analyses Project, Uncertainty Analysis of the Unmitigated Long-Term

Station Blackout of the Peach Bottom Atomic Power Station,” August 2013, ADAMS

Accession No. ML13189A145.
6-1

6 TECHNICAL EVALUATION SUMMARY AND KEY INSIGHTS

This report documents the results of research conducted to support the agency initiative to 

address the containment venting issue for the boiling water reactor (BWR) Mark I and Mark II 

containments. The research focused on developing the technical basis for a potential rulemaking 

action on containment protection and release reduction (CPRR), and covered three areas of 

analyses: (1) accident sequence analysis (event tree development) to identify accident sequences 

deemed to be the most significant risk contributors; (2) accident analysis of these sequences and 

assessment of radiological source terms; and (3) analysis of consequences with particular 

emphasis on health effects, both short term and long term. 

The accident sequence analysis evaluated a large number of sequences representing various 

operator actions such as RPV depressurization, wetwell and drywell venting, and water addition 

strategies. Also, seismically induced equipment failures were analyzed in detail. Models to 

estimate the frequency of ELAP events resulting from internal events and earthquakes were 

based on seismic hazard estimates developed by the industry. The human reliability aspect was 

considered in the accident sequence formulation, and despite an initial attempt to develop a 

comprehensive human reliability assessment (HRA), only a bounding approach to incorporating 

HRA into the accident sequences was implemented at the end. Sensitivity evaluations were done 

to gain insight into how human error probability affects the accident sequence frequencies. 

In the accident sequence analysis for the CPRR rulemaking (SECY-15-0085), the core-damage 

frequency (CDF) due to ELAPs is calculated to be 8.9x10-6 per reactor year, which is two times 

lower than the value of 1.6x10-5 that was estimated for SECY-12-0157. The CDF was calculated 

by averaging together the CDF for each BWR plant that was included in the scope of the accident 

sequence analysis (see Table 2-1). Also, the conditional core-damage probability (CCDP) given

the occurrence of an ELAP was calculated to be about 47% (i.e., the mitigation strategies required 

by Order EA-12-049 reduce the CDF by about 53%).

There was no fundamental shift in the scope and technical approach with regard to MELCOR

analysis performed in support of the CPRR rulemaking when compared to what was done in 

SECY-12-0157. The scope of MELCOR analysis falls broadly into two categories: (1) reactor 

systems and containment thermal-hydraulics under severe accident conditions, and (2) 

assessment of source terms (i.e., timing and magnitude of fission product releases to the 

environment). The technical approach in both cases (SECY-12-0157 and SECY-15-0085) takes 

into account best estimate modeling of accident progression, and incorporates both preventative

and mitigative accident management measures including venting, water addition and/or water

management, and the use of engineered filters. The selection of accident sequences covered by 

the MELCOR calculation matrix was informed by the set of accident sequences delineated in the 

accident sequence analysis. That said, the current analysis incorporates operator actions (e.g., 

opening and closing the wetwell vent early in an ELAP, anticipatory early venting rather than 

keeping the vent closed until core damage is imminent, RPV depressurization to 200-400 psig in 

order to minimize SRV cycling and heatup of the suppression pool, severe accident water 

management, etc.) that were not considered in SECY-12-0157. Some of these are strategies

selected by the industry to comply with Order EA-12-049, “Mitigation Strategies for BeyondDesign-Basis External Events.” 
6-2

The outcome of MELCOR analysis in the thermal-hydraulics category includes RPV and 

containment temperature and pressure signatures, and hydrogen distribution in the containment, 

reactor building, and vent line, all indicative of the state of containment vulnerability under severe 

accident conditions. These quantities provide needed information to assess containment integrity 

and also provide technical insights for developing staff guidance for Order EA-13-109 (severe 

accident capable hardened vent). The MELCOR analysis also provides source terms (estimates 

of fission products release to the environment), which are used by MACCS to calculate offsite 

consequences.

The MACCS analysis documented in this report, likewise, is similar to that in SECY-12-0157. With 

regard to the technical approach, the analysis in both cases uses source term estimates from 

MELCOR to calculate atmospheric transport and dispersion, protective actions, exposures, and 

resulting offsite consequences. The offsite consequence results are presented in terms of 

individual early fatality risk, individual latent cancer fatality (ILCF) risk, population dose, offsite 

cost, contaminated land area, and population subject to long-term protective actions. 

The quantitative results from the current analysis indicate no early fatality risk for all cases 

analyzed and the ILCF risk a factor of 10 lower than the recommended Safety Goal QHO 

acceptance level. The conditional ILCF risk (per event) is dominated by long-term phase 

exposures to contaminated areas. Because of the habitability criterion, this metric is relatively 

insensitive to the source term magnitude, whereas the societal consequence metrics are often 

more sensitive. The Commission direction in the SRM-SECY-12-0110, “Consideration of 

Economic Consequences within the U.S. Nuclear Regulatory Commission’s Regulatory 

Framework,” was considered by the staff in weighing land contamination and other factors.

6.1 Key Insights from Accident Sequence Analysis and Risk Integration

The accident sequence analysis results show a low value for core damage frequency from an 

ELAP event, and provide insights into which initiating events, mitigation systems, and operator 

actions contribute the most to overall core damage frequency for the BWR plants with Mark I and 

Mark II containments. Key insights from the accident sequence analysis include: 

• The major contribution to seismically induced ELAP is from earthquakes whose ground

motion exceeds the plant design basis (the safe shutdown earthquake). Specifically,

earthquakes with peak ground accelerations in the range of 0.3 to 0.75g are the major

contributors.

• Significant contributors to CDF include seismic failures of the batteries, DC switchgear,

and the EDGs and their supporting equipment. Failure of the portable FLEX pump and

failure to start of the RCIC pump are also significant contributors.

• CDF is not particularly sensitive to the human error probabilities for in-control-room and

ex-control-room operator actions.

• The estimated mean individual latent cancer fatality risk (0-10 miles) is more than two

orders of magnitude below the NRC Safety Goal QHO. The risk is low because the coredamage frequency is low and the conditional latent cancer fatality risk is low. The range

of parametric uncertainty in the risk estimates is more than one order of magnitude, and

is largely driven by uncertainty in the seismic hazard curves.
6-3

• The estimated individual early fatality risk is essentially zero in all cases analyzed and

for all alternatives considered, consistent with the findings previously in SOARCA and

SFP studies. This risk remains unchanged for a wide range of sensitivity analysis.

6.2 Key Insights from Accident Progression Analysis

The MELCOR analysis investigated detailed accident progression, source terms, and the containment 

response following an ELAP subject to appropriate initial and boundary conditions for the 

representative Mark I and Mark II containment designs. The analysis included sensitivities to venting, 

RPV depressurization, water injection (to RPV or drywell), and water management strategy. Additional 

sensitivities were performed for Mark II containments to examine the impact of the different cavity 

designs among the fleet. Following are some key insights from the MELCOR analysis:

Venting of Mark I and Mark II containments effectively prevents containment overpressure failure.

Pre-core damage anticipatory venting reduces the containment base pressure at the time of core 

damage and results in a delay when post core damage venting is required. Post-core damage 

containment venting is efficient in purging hydrogen and other non-condensables from the 

containment, and reduces the likelihood of combustible quantities of non-condensable 

accumulation in the reactor building. 

• Creep rupture of the main steam line seems unlikely if the reactor pressure is maintained

low. The failure of the main steam line results in failure of the containment (opening of

the upper drywell head) and bypass of the suppression pool. The failure also results in

migration of hydrogen to the refueling bay of the Mark I containment, and consequent

hydrogen combustion and release of fission products directly to the environment.

• Venting alone, however, is not adequate, as it does not prevent other modes of

containment failure such as liner melt-through and over-temperature failure of the upper

drywell head, bypass of the suppression pool and direct release of radioactivity to the

environment. A combination of venting and water injection is required to prevent such

failures, and the current work provides a sound technical basis to that effect thus

supporting adequate protection argument.

• Addition of water either into the RPV or the drywell has the following benefits: (1) cooling

of the core debris and containment atmosphere; (2) preventing over-temperature failure

of the upper drywell head; (3) preventing and/or delaying liner melt through in Mark I

containments; (4) maintaining a steam inerted atmosphere which can preclude an

energetic hydrogen combustion; and (5) mitigating radiological releases as it effectively

provides means for fission products scrubbing.

• Environmental releases from Mark II containments are in general comparable to or lower

than those from Mark I containments. Sensitivity analysis performed to investigate

variations in lower cavity configurations of Mark II containments indicate the

environmental releases for all configurations are within the range of releases predicted

for Mark I containments.

• Assuming the condensate storage tank survives a beyong design basis accident, RCIC

suction initially taken from the CST provides a better alternative to suction from

suppression pool (SP) as this action will likely extend the duration of RCIC duration.
6-4

6.3 Key Insights from Offsite Consequence Analysis

The MACCS evaluation results include individual early fatality risk, ILCF risk, total population 

dose, total offsite cost, land contamination, and population subject to long-term phase protective 

actions for all Mark I and II source terms calculated by MELCOR. The analysis included several 

sensitivities such as evacuation delay, non-evacuating population, and habitability criteria.

Following are key insights from the MACCS analysis:

• There is essentially zero individual early fatality risk for all Mark I and Mark II source

term bin cases analyzed because the releases are not large enough to exceed the

threshold for the acute dose to the red bone marrow, which is typically the most sensitive

tissue for early fatalities.

• For all Mark I and II source terms analyzed, the conditional ILCF risk is sufficiently low

that when multiplied by the accident frequency, there is at least a two order of magnitude

margin to the QHO.

• The calculated releases to the environment are delayed long enough after the accident

initiation to allow time for the emergency planning zone (EPZ) population to evacuate.

Therefore the ILCF risk is dominated by long-term phase exposures to slightly

contaminated areas (under the 500 mrem per year habitability criterion threshold).

• In general, a larger release of radioactive materials to the environment displaces more

people for more time, and therefore incurs a larger societal cost. However, for a larger

release, the cancer fatality risk to the public shows a nonlinear effect because protective

actions are in place primarily to reduce exposures (habitability criterion) at the tradeoff of

other societal costs such as land contamination and economic losses.

• The potential effectiveness of an external filter on reducing the environmental release is

heavily influenced by release pathways. For accident cases in which the entire release is

through the venting system, there can be significant reductions to the environment.

However, there is considerably less benefit from an external filter when any of the

release bypasses the venting system.

• MACCS sensitivity calculations which generally showed the greatest increase in

individual latent cancer fatality risk were those with medium and large releases in which

the EPA PAG habitability criterion (2 rem in the first year) was used instead of the

Pennsylvania habitability criterion (500 mrem in the first year), thus allowing a first year

dose up to four times higher. The individual latent cancer fatality risk was commonly 2-3

times higher than for the base cases.

• All sensitivity calculations conducted, including those with changes to population,

evacuation characteristics, intermediate phase duration, and long-term habitability

criterion, do not change any of the consequence analysis insights related to the QHO

metrics. Individual early fatality risk remains essentially zero for all sensitivity

calculations conducted, and ILCF risk remains well below the QHO.
6-5

6.4 Concluding Remarks

Coincident with the analyses by the NRC staff and industry related to the CPRR rulemaking and 

related Orders, licensees were developing revisions to the severe accident managmenet

guidelines (SAMGs) to address lessons learned from the Fukushima accident. The analyses

performed to address issues related to containment venting and severe accident water addition 

provided valuable insights and supported actual revisions to the SAMGs for plants with Mark I and 

Mark II containments. This incorporation of insights from the modelling of beyond-design-basis

events and severe accidents into plant guidance documents provides a useful example of the 

potential benefits of the efforts by the NRC and industry to develop improved analytical

capabilities. Another example is the use of the results and technical insights in formulating the

regulatory basis for the mitigation of beyond design basis events rulemaking. This rulemaking,

though not imposing any regulatory footprint on SAMG, nevertheless directs the NRC staff to

provide periodic oversight to industry’s SAMG implementation through NRC’s updated Reactor

Oversight Program (ROP). A third and related example is that of a recent initiative in many

Organization for Economic Cooperation and Development (OECD) countries and a collective 

effort in the OECD program of work to develop technical insights on how the type of beyond 

design basis analysis work documented in this report can inform SAMG in a positive way and in 

so doing, enhance the safety of nuclear power plants.

The NRC staff are exploring possible ways to ensure that future insights from severe accident

research and analyses are shared with and considered by licensees even when some of those 

insights are unlikely to initiate regulatory actions. The NRC staff, through its continued 

engagement in severe accident research, plans to remain cognizant about both the industry and 

the regulatory activities in other countries.

A-1 

APPENDIX A 

CORE DAMAGE EVENT TREES

Figure A-1 CDET for WWF venting strategy, Sheet 1

Extended 

Loss of AC 

Power

Phase 1 DC Phase 1 

RCIC

RPV 

Depress

RPV 

Pressure for 

ST RCIC

Phase 2 DC Early WW 

Vent

Early DW 

Vent

ELAP DC1 RCIC1 P0 P1 DC2 WW DW

to Sheet 2, Seqs. 1 to 22

to Sheet 3, Seqs. 23 to 44

WW

DW 45 CD M-MP-IS-OK-XX

to Sheet 4, Seqs. 46 to 67

DC2 to Sheet 5, Seqs. 68 to 89

WWnodc

DWnodc 90 CD M-MP-IS-LT-XX

P1 to Sheet 6, Seqs. 91 to 112

to Sheet 7, Seqs. 113 to 134

to Sheet 8, Seqs. 135 to 156

WW

DW 157 CD M-HP-IS-OK-XX

to Sheet 9, Seqs. 158 to 179

P0DP DC2 to Sheet 10, Seqs. 180 to 201

WWnodc

DWnodc 202 CD M-HP-IS-LT-XX

P1NODP to Sheet 11, Seqs. 203 to 224

225 CD E-LP-IS-XX-XX

RCIC1

P0MSCL 226 CD E-HP-IS-XX-XX

to Sheet 12, Seqs. 227 to 248

to Sheet 13, Seqs. 249 to 270

WWnodc

DWnodc 271 CD M-MP-IS-ST-XX

P1nodc to Sheet 14, Seqs. 272 to 282

to Sheet 15, Seqs. 283 to 304

to Sheet 16, Seqs. 305 to 326

WWnodc

P0DPnodc DWnodc 327 CD M-HP-IS-ST-XX

DC1

P1NODPnodc to Sheet 17, Seqs. 328 to 338

339 CD E-LP-IS-ST-XX

RCIC1nodc

P0MSCLnodc 340 CD E-HP-IS-ST-XX

Plant Damage 

State No. Status
A-2 

Figure A-2 CDET for WWF venting strategy, Sheet 2, Sequences 1 to 22

Phase 2 

RCIC

RPV 

Pressure for 

LT RCIC

SP Makeup RPV 

Injection

FLEX

Pump

RPV 

Pressure for 

RPV 

Injection

Reclose 

Vents

RCIC2 P2 SPMU RPVINJ PUMP PINJ CVENT

1 OK

2 CD L-MP-IS-OK-F

RPVHW

CVENT 3 CD L-MP-WW-OK-F

4 OK

5 CD L-MP-IS-OK-OK

PINJ

CVENT 6 CD L-MP-WW-OK-OK

7 CD L-MP-IS-OK-F

SPHFE RPVHW

CVENT 8 CD L-MP-WW-OK-F

9 CD L-MP-IS-OK-H

RPVHFE

CVENT 10 CD L-MP-WW-OK-H

11 OK

Entry Conditions

DC1 up 12 CD M-LP-IS-OK-F

RCIC1 up P2 RPVHW

P0 up CVENT 13 CD M-LP-WW-OK-F

P1 up

DC2 up 14 CD M-LP-IS-OK-H

WW up RPVHFE

CVENT 15 CD M-LP-WW-OK-H

16 OK

17 CD M-MP-IS-OK-OK

PINJ

CVENT 18 CD M-MP-WW-OK-OK

19 CD M-MP-IS-OK-F

RCIC2 RPVHW

CVENT 20 CD M-MP-WW-OK-F

21 CD M-MP-IS-OK-H

RPVHFE

CVENT 22 CD M-MP-WW-OK-H

No. Status Plant Damage 

State

Transfer from

Sheet 1
A-3

Figure A-3 CDET for WWF venting strategy, Sheet 3, Sequences 23 to 44

Phase 2 

RCIC

RPV 

Pressure for 

LT RCIC

SP Makeup RPV 

Injection

FLEX

Pump

RPV 

Pressure for 

RPV 

Injection

Reclose 

Vents

RCIC2 P2 SPMU RPVINJ PUMP PINJ CVENT

23 OK

24 CD L-MP-IS-OK-F

RPVHW

CVENT 25 CD L-MP-DW-OK-F

26 OK

27 CD L-MP-IS-OK-OK

PINJ

CVENT 28 CD L-MP-DW-OK-OK

29 CD L-MP-IS-OK-F

SPHFE RPVHW

CVENT 30 CD L-MP-DW-OK-F

31 CD L-MP-IS-OK-H

RPVHFE

CVENT 32 CD L-MP-DW-OK-H

Entry Conditions 33 OK

DC1 up

RCIC1 up 34 CD M-LP-IS-OK-F

P0 up P2 RPVHW

P1 up CVENT 35 CD M-LP-DW-OK-F

DC2 up

WW down 36 CD M-LP-IS-OK-H

DW up RPVHFE

CVENT 37 CD M-LP-DW-OK-H

38 OK

39 CD M-MP-IS-OK-OK

PINJ

CVENT 40 CD M-MP-DW-OK-OK

41 CD M-MP-IS-OK-F

RCIC2 RPVHW

CVENT 42 CD M-MP-DW-OK-F

43 CD M-MP-IS-OK-H

RPVHFE

CVENT 44 CD M-MP-DW-OK-H

Transfer from

Sheet 1 No. Status Plant Damage State
A-4 

Figure A-4 CDET for WWF venting strategy, Sheet 4, Sequences 46 to 67

Phase 2 

RCIC

RPV 

Pressure for 

LT RCIC

SP Makeup RPV 

Injection

FLEX

Pump

RPV 

Pressure for 

RPV 

Injection

Reclose 

Vents

RCIC2 P2 SPMU RPVINJ PUMP PINJ CVENT

46 OK

47 CD L-MP-IS-LT-F

RPVHW

CVENTnodc 48 CD L-MP-WW-LT-F

49 OK

50 CD L-MP-IS-LT-OK

PINJnodc

CVENTnodc 51 CD L-MP-WW-LT-OK

52 CD L-MP-IS-LT-F

SPHFE RPVHW

CVENTnodc 53 CD L-MP-WW-LT-F

54 CD L-MP-IS-LT-H

RPVHFE

CVENTnodc 55 CD L-MP-WW-LT-H

56 OK

Entry Conditions

DC1 up 57 CD M-LP-IS-LT-F

RCIC1 up P2nodc RPVHW

P0 up CVENTnodc 58 CD M-LP-WW-LT-F

P1 up

DC2 down 59 CD M-LP-IS-LT-H

WW up RPVHFE

CVENTnodc 60 CD M-LP-WW-LT-H

61 OK

62 CD M-MP-IS-LT-OK

PINJnodc

CVENTnodc 63 CD M-MP-WW-LT-OK

64 CD M-MP-IS-LT-F

RCIC2nodc RPVHW

CVENTnodc 65 CD M-MP-WW-LT-F

66 CD M-MP-IS-LT-H

RPVHFE

CVENTnodc 67 CD M-MP-WW-LT-H

Transfer from

Sheet 1 No. Status Plant Damage State
A-5

Figure A-5 CDET for WWF venting strategy, Sheet 5, Sequences 68 to 89

Phase 2 

RCIC

RPV 

Pressure for 

LT RCIC

SP Makeup RPV 

Injection

FLEX

Pump

RPV 

Pressure for 

RPV 

Injection

Reclose 

Vents

RCIC2 P2 SPMU RPVINJ PUMP PINJ CVENT

68 OK

69 CD L-MP-IS-LT-F

RPVHW

CVENTnodc 70 CD L-MP-DW-LT-F

71 OK

72 CD L-MP-IS-LT-OK

PINJnodc

CVENTnodc 73 CD L-MP-DW-LT-OK

74 CD L-MP-IS-LT-F

SPHFE RPVHW

CVENTnodc 75 CD L-MP-DW-LT-F

76 CD L-MP-IS-LT-H

RPVHFE

CVENTnodc 77 CD L-MP-DW-LT-H

Entry Conditions 78 OK

DC1 up

RCIC1 up 79 CD M-LP-IS-LT-F

P0 up P2nodc RPVHW

P1 up CVENTnodc 80 CD M-LP-DW-LT-F

DC2 down

WW down 81 CD M-LP-IS-LT-H

DW up RPVHFE

CVENTnodc 82 CD M-LP-DW-LT-H

83 OK

84 CD M-MP-IS-LT-OK

PINJnodc

CVENTnodc 85 CD M-MP-DW-LT-OK

86 CD M-MP-IS-LT-F

RCIC2nodc RPVHW

CVENTnodc 87 CD M-MP-DW-LT-F

88 CD M-MP-IS-LT-H

RPVHFE

CVENTnodc 89 CD M-MP-DW-LT-H

Transfer from

Sheet 1 No. Status Plant Damage State
A-6 

Figure A-6 CDET for WWF venting strategy, Sheet 6, Sequences 91 to 112

Phase 2 DC Early WW 

Vent

Early DW 

Vent

Phase 2 

RCIC

RPV 

Pressure for 

LT RCIC

SP Makeup RPV 

Injection

FLEX

Pump

RPV 

Pressure for 

RPV 

Injection

Reclose 

Vents

DC2 WW DW RCIC2 P2 SPMU RPVINJ PUMP PINJ CVENT

91 OK

92 CD M-LP-IS-OK-F

RPVHW

CVENT 93 CD M-LP-WW-OK-F

94 CD M-LP-IS-OK-H

RPVHFE

CVENT 95 CD M-LP-WW-OK-H

96 OK

97 CD M-LP-IS-OK-F

RPVHW

CVENT 98 CD M-LP-DW-OK-F

Entry Conditions 99 CD M-LP-IS-OK-H

DC1 up WW RPVHFE

RCIC1 up CVENT 100 CD M-LP-DW-OK-H

P0 up

P1 down DW 101 CD M-LP-IS-OK-XX

102 OK

103 CD M-LP-IS-LT-F

RPVHW

CVENTnodc 104 CD M-LP-WW-LT-F

105 CD M-LP-IS-LT-H

RPVHFE

CVENTnodc 106 CD M-LP-WW-LT-H

DC2

107 OK

108 CD M-LP-IS-LT-F

RPVHW

CVENTnodc 109 CD M-LP-DW-LT-F

110 CD M-LP-IS-LT-H

WWnodc RPVHFE

CVENTnodc 111 CD M-LP-DW-LT-H

DWnodc 112 CD M-LP-IS-LT-XX

Transfer from

Sheet 1 No. Status Plant Damage State
A-7 

Figure A-7 CDET for WWF venting strategy, Sheet 7, Sequences 113 to 134

Phase 2 

RCIC

RPV 

Pressure for 

LT RCIC

SP Makeup RPV 

Injection

FLEX

Pump

RPV 

Pressure for 

RPV 

Injection

Reclose 

Vents

RCIC2 P2 SPMU RPVINJ PUMP PINJ CVENT

113 OK

114 CD L-HP-IS-OK-F

RPVHW

CVENT 115 CD L-HP-WW-OK-F

116 OK

117 CD L-HP-IS-OK-OK

PINJ

CVENT 118 CD L-HP-WW-OK-OK

119 CD L-HP-IS-OK-F

SPHFE RPVHW

CVENT 120 CD L-HP-WW-OK-F

121 CD L-HP-IS-OK-H

RPVHFE

CVENT 122 CD L-HP-WW-OK-H

123 OK

Entry Conditions

DC1 up 124 CD M-LP-IS-OK-F

RCIC1 up P2NODP RPVHW

P0 down CVENT 125 CD M-LP-WW-OK-F

P1 up

DC2 up 126 CD M-LP-IS-OK-H

WW up RPVHFE

CVENT 127 CD M-LP-WW-OK-H

128 OK

129 CD M-HP-IS-OK-OK

PINJ

CVENT 130 CD M-HP-WW-OK-OK

131 CD M-HP-IS-OK-F

RCIC2 RPVHW

CVENT 132 CD M-HP-WW-OK-F

133 CD M-HP-IS-OK-H

RPVHFE

CVENT 134 CD M-HP-WW-OK-H

Transfer from

Sheet 1 No. Status Plant Damage State
A-8

Figure A-8 CDET for WWF venting strategy, Sheet 8, Sequences 135 to 156

Phase 2 

RCIC

RPV 

Pressure for 

LT RCIC

SP Makeup RPV 

Injection

FLEX

Pump

RPV 

Pressure for 

RPV 

Injection

Reclose 

Vents

RCIC2 P2 SPMU RPVINJ PUMP PINJ CVENT

135 OK

136 CD L-HP-IS-OK-F

RPVHW

CVENT 137 CD L-HP-DW-OK-F

138 OK

139 CD L-HP-IS-OK-OK

PINJ

CVENT 140 CD L-HP-DW-OK-OK

141 CD L-HP-IS-OK-F

SPHFE RPVHW

CVENT 142 CD L-HP-DW-OK-F

143 CD L-HP-IS-OK-H

RPVHFE

CVENT 144 CD L-HP-DW-OK-H

Entry Conditions 145 OK

DC1 up

RCIC1 up 146 CD M-LP-IS-OK-F

P0 down P2NODP RPVHW

P1 up CVENT 147 CD M-LP-DW-OK-F

DC2 up

WW down 148 CD M-LP-IS-OK-H

DW up RPVHFE

CVENT 149 CD M-LP-DW-OK-H

150 OK

151 CD M-HP-IS-OK-OK

PINJ

CVENT 152 CD M-HP-DW-OK-OK

153 CD M-HP-IS-OK-F

RCIC2 RPVHW

CVENT 154 CD M-HP-DW-OK-F

155 CD M-HP-IS-OK-H

RPVHFE

CVENT 156 CD M-HP-DW-OK-H

Transfer from

Sheet 1 No. Status Plant Damage State
A-9

Figure A-9 CDET for WWF venting strategy, Sheet 9, Sequences 158 to 179

Phase 2 

RCIC

RPV 

Pressure for 

LT RCIC

SP Makeup RPV 

Injection

FLEX

Pump

RPV 

Pressure for 

RPV 

Injection

Reclose 

Vents

RCIC2 P2 SPMU RPVINJ PUMP PINJ CVENT

158 OK

159 CD L-HP-IS-LT-F

RPVHW

CVENTnodc 160 CD L-HP-WW-LT-F

161 OK

162 CD L-HP-IS-LT-OK

PINJnodc

CVENTnodc 163 CD L-HP-WW-LT-OK

164 CD L-HP-IS-LT-F

SPHFE RPVHW

CVENTnodc 165 CD L-HP-WW-LT-F

166 CD L-HP-IS-LT-H

RPVHFE

CVENTnodc 167 CD L-HP-WW-LT-H

168 OK

Entry Conditions

DC1 up 169 CD M-LP-IS-LT-F

RCIC1 up P2NODPnodc RPVHW

P0 down CVENTnodc 170 CD M-LP-WW-LT-F

P1 up

DC2 down 171 CD M-LP-IS-LT-H

WW up RPVHFE

CVENTnodc 172 CD M-LP-WW-LT-H

173 OK

174 CD M-HP-IS-LT-OK

PINJnodc

CVENTnodc 175 CD M-HP-WW-LT-OK

176 CD M-HP-IS-LT-F

RCIC2nodc RPVHW

CVENTnodc 177 CD M-HP-WW-LT-F

178 CD M-HP-IS-LT-H

RPVHFE

CVENTnodc 179 CD M-HP-WW-LT-H

Transfer from

Sheet 1 No. Status Plant Damage State
A-10

Figure A-10 CDET for WWF venting strategy, Sheet 10, Sequences 180 to 201

Phase 2 

RCIC

RPV 

Pressure for 

LT RCIC

SP Makeup RPV 

Injection

FLEX

Pump

RPV 

Pressure for 

RPV 

Injection

Reclose 

Vents

RCIC2 P2 SPMU RPVINJ PUMP PINJ CVENT

180 OK

181 CD L-HP-IS-LT-F

RPVHW

CVENTnodc 182 CD L-HP-DW-LT-F

183 OK

184 CD L-HP-IS-LT-OK

PINJnodc

CVENTnodc 185 CD L-HP-DW-LT-OK

186 CD L-HP-IS-LT-F

SPHFE RPVHW

CVENTnodc 187 CD L-HP-DW-LT-F

188 CD L-HP-IS-LT-H

RPVHFE

CVENTnodc 189 CD L-HP-DW-LT-H

Entry Conditions 190 OK

DC1 up

RCIC1 up 191 CD M-LP-IS-LT-F

P0 down P2NODPnodc RPVHW

P1 up CVENTnodc 192 CD M-LP-DW-LT-F

DC2 down

WW down 193 CD M-LP-IS-LT-H

DW up RPVHFE

CVENTnodc 194 CD M-LP-DW-LT-H

195 OK

196 CD M-HP-IS-LT-OK

PINJnodc

CVENTnodc 197 CD M-HP-DW-LT-OK

198 CD M-HP-IS-LT-F

RCIC2nodc RPVHW

CVENTnodc 199 CD M-HP-DW-LT-F

200 CD M-HP-IS-LT-H

RPVHFE

CVENTnodc 201 CD M-HP-DW-LT-H

Transfer from

Sheet 1 No. Status Plant Damage State
A-11

Figure A-11 CDET for WWF venting strategy, Sheet 11, Sequences 203 to 224

Phase 2 DC Early WW 

Vent

Early DW 

Vent

Phase 2 

RCIC

RPV 

Pressure for 

LT RCIC

SP Makeup RPV 

Injection

FLEX

Pump

RPV 

Pressure for 

RPV 

Injection

Reclose 

Vents

DC2 WW DW RCIC2 P2 SPMU RPVINJ PUMP PINJ CVENT

203 OK

204 CD M-LP-IS-OK-F

RPVHW

CVENT 205 CD M-LP-WW-OK-F

206 CD M-LP-IS-OK-H

RPVHFE

CVENT 207 CD M-LP-WW-OK-H

208 OK

209 CD M-LP-IS-OK-F

RPVHW

CVENT 210 CD M-LP-DW-OK-F

Entry Conditions 211 CD M-LP-IS-OK-H

DC1 up WW RPVHFE

RCIC1 up CVENT 212 CD M-LP-DW-OK-H

P0 down

P1 down DW 213 CD M-LP-IS-OK-XX

214 OK

215 CD M-LP-IS-LT-F

RPVHW

CVENTnodc 216 CD M-LP-WW-LT-F

217 CD M-LP-IS-LT-H

RPVHFE

CVENTnodc 218 CD M-LP-WW-LT-H

DC2

219 OK

220 CD M-LP-IS-LT-F

RPVHW

CVENTnodc 221 CD M-LP-DW-LT-F

222 CD M-LP-IS-LT-H

WWnodc RPVHFE

CVENTnodc 223 CD M-LP-DW-LT-H

DWnodc 224 CD M-LP-IS-LT-XX

Transfer from

Sheet 1 No. Status Plant Damage State
A-12

Figure A-12 CDET for WWF venting strategy, Sheet 12, Sequences 227 to 248

Phase 2 

RCIC

RPV 

Pressure for 

LT RCIC

SP Makeup RPV 

Injection

FLEX

Pump

RPV 

Pressure for 

RPV 

Injection

Reclose 

Vents

RCIC2 P2 SPMU RPVINJ PUMP PINJ CVENT

227 OK

228 CD L-MP-IS-ST-F

RPVHW

CVENTnodc 229 CD L-MP-WW-ST-F

230 OK

231 CD L-MP-IS-ST-OK

PINJnodc

CVENTnodc 232 CD L-MP-WW-ST-OK

233 CD L-MP-IS-ST-F

SPHFE RPVHW

CVENTnodc 234 CD L-MP-WW-ST-F

235 CD L-MP-IS-ST-H

RPVHFE

CVENTnodc 236 CD L-MP-WW-ST-H

237 OK

Entry Conditions 238 CD M-LP-IS-ST-F

DC1 down P2nodc RPVHW

RCIC1 up CVENTnodc 239 CD M-LP-WW-ST-F

P0 up

P1 up 240 CD M-LP-IS-ST-H

WW up RPVHFE

CVENTnodc 241 CD M-LP-WW-ST-H

242 OK

243 CD M-MP-IS-ST-OK

PINJnodc

CVENTnodc 244 CD M-MP-WW-ST-OK

245 CD M-MP-IS-ST-F

RCIC2nodc RPVHW

CVENTnodc 246 CD M-MP-WW-ST-F

247 CD M-MP-IS-ST-H

RPVHFE

CVENTnodc 248 CD M-MP-WW-ST-H

Transfer from

Sheet 1 No. Status Plant Damage State
A-13

Figure A-13 CDET for WWF venting strategy, Sheet 13, Sequences 249 to 270

Phase 2 

RCIC

RPV 

Pressure for 

LT RCIC

SP Makeup RPV 

Injection

FLEX

Pump

RPV 

Pressure for 

RPV 

Injection

Reclose 

Vents

RCIC2 P2 SPMU RPVINJ PUMP PINJ CVENT

249 OK

250 CD L-MP-IS-ST-F

RPVHW

CVENTnodc 251 CD L-MP-DW-ST-F

252 OK

253 CD L-MP-IS-ST-OK

PINJnodc

CVENTnodc 254 CD L-MP-DW-ST-OK

255 CD L-MP-IS-ST-F

SPHFE RPVHW

CVENTnodc 256 CD L-MP-DW-ST-F

257 CD L-MP-IS-ST-H

RPVHFE

CVENTnodc 258 CD L-MP-DW-ST-H

259 OK

Entry Conditions

DC1 down 260 CD M-LP-IS-ST-F

RCIC1 up P2nodc RPVHW

P0 up CVENTnodc 261 CD M-LP-DW-ST-F

P1 up

WW down 262 CD M-LP-IS-ST-H

DW up RPVHFE

CVENTnodc 263 CD M-LP-DW-ST-H

264 OK

265 CD M-MP-IS-ST-OK

PINJnodc

CVENTnodc 266 CD M-MP-DW-ST-OK

267 CD M-MP-IS-ST-F

RCIC2nodc RPVHW

CVENTnodc 268 CD M-MP-DW-ST-F

269 CD M-MP-IS-ST-H

RPVHFE

CVENTnodc 270 CD M-MP-DW-ST-H

Transfer from

Sheet 1 No. Status Plant Damage State
A-14

Figure A-14 CDET for WWF venting strategy, Sheet 14, Sequences 272 to 282

Phase 2 DC Early WW 

Vent

Early DW 

Vent

Phase 2 

RCIC

RPV 

Pressure for 

LT RCIC

SP Makeup RPV 

Injection

FLEX

Pump

RPV 

Pressure for 

RPV 

Injection

Reclose 

Vents

DC2 WW DW RCIC2 P2 SPMU RPVINJ PUMP PINJ CVENT

272 OK

273 CD M-LP-IS-ST-F

RPVHW

CVENTnodc 274 CD M-LP-WW-ST-F

Entry Conditions

DC1 down 275 CD M-LP-IS-ST-H

RCIC1 up RPVHFE

P0 up CVENTnodc 276 CD M-LP-WW-ST-H

P1 down

277 OK

278 CD M-LP-IS-ST-F

RPVHW

CVENTnodc 279 CD M-LP-DW-ST-F

280 CD M-LP-IS-ST-H

WWnodc RPVHFE

CVENTnodc 281 CD M-LP-DW-ST-H

DWnodc 282 CD M-LP-IS-ST-XX

Transfer from

Sheet 1 No. Status Plant Damage State
A-15

Figure A-15 CDET for WWF venting strategy, Sheet 15, Sequences 283 to 304

Phase 2 

RCIC

RPV 

Pressure for 

LT RCIC

SP Makeup RPV 

Injection

FLEX

Pump

RPV 

Pressure for 

RPV 

Injection

Reclose 

Vents

RCIC2 P2 SPMU RPVINJ PUMP PINJ CVENT

283 OK

284 CD L-HP-IS-ST-F

RPVHW

CVENTnodc 285 CD L-HP-WW-ST-F

286 OK

287 CD L-HP-IS-ST-OK

PINJnodc

CVENTnodc 288 CD L-HP-WW-ST-OK

289 CD L-HP-IS-ST-F

SPHFE RPVHW

CVENTnodc 290 CD L-HP-WW-ST-F

291 CD L-HP-IS-ST-H

RPVHFE

CVENTnodc 292 CD L-HP-WW-ST-H

293 OK

Entry Conditions 294 CD M-LP-IS-ST-F

DC1 down P2NODPnodc RPVHW

RCIC1 up CVENTnodc 295 CD M-LP-WW-ST-F

P0 down

P1 up 296 CD M-LP-IS-ST-H

WW up RPVHFE

CVENTnodc 297 CD M-LP-WW-ST-H

298 OK

299 CD M-HP-IS-ST-OK

PINJnodc

CVENTnodc 300 CD M-HP-WW-ST-OK

301 CD M-HP-IS-ST-F

RCIC2nodc RPVHW

CVENTnodc 302 CD M-HP-WW-ST-F

303 CD M-HP-IS-ST-H

RPVHFE

CVENTnodc 304 CD M-HP-WW-ST-H

Transfer from

Sheet 1 No. Status Plant Damage State
A-16

Figure A-16 CDET for WWF venting strategy, Sheet 16, Sequences 305 to 326

Phase 2 

RCIC

RPV 

Pressure for 

LT RCIC

SP Makeup RPV 

Injection

FLEX

Pump

RPV 

Pressure for 

RPV 

Injection

Reclose 

Vents

RCIC2 P2 SPMU RPVINJ PUMP PINJ CVENT

305 OK

306 CD L-HP-IS-ST-F

RPVHW

CVENTnodc 307 CD L-HP-DW-ST-F

308 OK

309 CD L-HP-IS-ST-OK

PINJnodc

CVENTnodc 310 CD L-HP-DW-ST-OK

311 CD L-HP-IS-ST-F

SPHFE RPVHW

CVENTnodc 312 CD L-HP-DW-ST-F

313 CD L-HP-IS-ST-H

RPVHFE

CVENTnodc 314 CD L-HP-DW-ST-H

315 OK

Entry Conditions

DC1 down 316 CD M-LP-IS-ST-F

RCIC1 up P2NODPnodc RPVHW

P0 down CVENTnodc 317 CD M-LP-DW-ST-F

P1 up

WW down 318 CD M-LP-IS-ST-H

DW up RPVHFE

CVENTnodc 319 CD M-LP-DW-ST-H

320 OK

321 CD M-HP-IS-ST-OK

PINJnodc

CVENTnodc 322 CD M-HP-DW-ST-OK

323 CD M-HP-IS-ST-F

RCIC2nodc RPVHW

CVENTnodc 324 CD M-HP-DW-ST-F

325 CD M-HP-IS-ST-H

RPVHFE

CVENTnodc 326 CD M-HP-DW-ST-H

Transfer from

Sheet 1 No. Status Plant Damage State
A-17

Figure A-17 CDET for WWF venting strategy, Sheet 17, Sequences 328 to 338

Phase 2 DC Early WW 

Vent

Early DW 

Vent

Phase 2 

RCIC

RPV 

Pressure for 

LT RCIC

SP Makeup RPV 

Injection

FLEX

Pump

RPV 

Pressure for 

RPV 

Injection

Reclose 

Vents

DC2 WW DW RCIC2 P2 SPMU RPVINJ PUMP PINJ CVENT

328 OK

329 CD M-LP-IS-ST-F

RPVHW

CVENTnodc 330 CD M-LP-WW-ST-F

Entry Conditions

DC1 down 331 CD M-LP-IS-ST-H

RCIC1 up RPVHFE

P0 down CVENTnodc 332 CD M-LP-WW-ST-H

P1 down

333 OK

334 CD M-LP-IS-ST-F

RPVHW

CVENTnodc 335 CD M-LP-DW-ST-F

336 CD M-LP-IS-ST-H

WWnodc RPVHFE

CVENTnodc 337 CD M-LP-DW-ST-H

DWnodc 338 CD M-LP-IS-ST-XX

Transfer from

Sheet 1 No. Status Plant Damage State
A-18

Figure A-18 CDET for DWF venting strategies, Sheet 1

Extended 

Loss of AC 

Power

Phase 1 DC Phase 1 

RCIC

RPV 

Depress

RPV 

Pressure for 

ST RCIC

Phase 2 DC Early DW 

Vent

Early WW 

Vent

ELAP DC1 RCIC1 P0 P1 DC2 DW WW

to Sheet 2, Seqs. 1 to 22

to Sheet 3, Seqs. 23 to 44

DW

WW 45 CD M-MP-IS-OK-XX

to Sheet 4, Seqs. 46 to 67

DC2 to Sheet 5, Seqs. 68 to 89

DWnodc

WWnodc 90 CD M-MP-IS-LT-XX

P1 to Sheet 6, Seqs. 91 to 112

to Sheet 7, Seqs. 113 to 134

to Sheet 8, Seqs. 135 to 156

DW

WW 157 CD M-HP-IS-OK-XX

to Sheet 9, Seqs. 158 to 179

P0DP DC2 to Sheet 10, Seqs. 180 to 201

DWnodc

WWnodc 202 CD M-HP-IS-LT-XX

P1NODP to Sheet 11, Seqs. 203 to 224

225 CD E-LP-IS-XX-XX

RCIC1

P0MSCL 226 CD E-HP-IS-XX-XX

to Sheet 12, Seqs. 227 to 248

to Sheet 13, Seqs. 249 to 270

DWnodc

WWnodc 271 CD M-MP-IS-ST-XX

P1nodc to Sheet 14, Seqs. 272 to 282

to Sheet 15, Seqs. 283 to 304

to Sheet 16, Seqs. 305 to 326

DWnodc

P0DPnodc WWnodc 327 CD M-HP-IS-ST-XX

DC1

P1NODPnodc to Sheet 17, Seqs. 328 to 338

339 CD E-LP-IS-ST-XX

RCIC1nodc

P0MSCLnodc 340 CD E-HP-IS-ST-XX

Plant Damage 

State No. Status
A-19

Figure A-19 CDET for DWF venting strategies, Sheet 2, Sequences 1 to 22

Phase 2 

RCIC

RPV 

Pressure for 

LT RCIC

SP Makeup RPV 

Injection

FLEX

Pump

RPV 

Pressure for 

RPV 

Injection

Reclose 

Vents

RCIC2 P2 SPMU RPVINJ PUMP PINJ CVENT

1 OK

2 CD L-MP-IS-OK-F

RPVHW

CVENT 3 CD L-MP-DW-OK-F

4 OK

5 CD L-MP-IS-OK-OK

PINJ

CVENT 6 CD L-MP-DW-OK-OK

7 CD L-MP-IS-OK-F

SPHFE RPVHW

CVENT 8 CD L-MP-DW-OK-F

9 CD L-MP-IS-OK-H

RPVHFE

CVENT 10 CD L-MP-DW-OK-H

11 OK

Entry Conditions

DC1 up 12 CD M-LP-IS-OK-F

RCIC1 up P2 RPVHW

P0 up CVENT 13 CD M-LP-DW-OK-F

P1 up

DC2 up 14 CD M-LP-IS-OK-H

DW up RPVHFE

CVENT 15 CD M-LP-DW-OK-H

16 OK

17 CD M-MP-IS-OK-OK

PINJ

CVENT 18 CD M-MP-DW-OK-OK

19 CD M-MP-IS-OK-F

RCIC2 RPVHW

CVENT 20 CD M-MP-DW-OK-F

21 CD M-MP-IS-OK-H

RPVHFE

CVENT 22 CD M-MP-DW-OK-H

No. Status Plant Damage 

State

Transfer from

Sheet 1
A-20

Figure A-20 CDET for DWF venting strategies, Sheet 3, Sequences 23 to 44

Phase 2 

RCIC

RPV 

Pressure for 

LT RCIC

SP Makeup RPV 

Injection

FLEX

Pump

RPV 

Pressure for 

RPV 

Injection

Reclose 

Vents

RCIC2 P2 SPMU RPVINJ PUMP PINJ CVENT

23 OK

24 CD L-MP-IS-OK-F

RPVHW

CVENT 25 CD L-MP-WW-OK-F

26 OK

27 CD L-MP-IS-OK-OK

PINJ

CVENT 28 CD L-MP-WW-OK-OK

29 CD L-MP-IS-OK-F

SPHFE RPVHW

CVENT 30 CD L-MP-WW-OK-F

31 CD L-MP-IS-OK-H

RPVHFE

CVENT 32 CD L-MP-WW-OK-H

Entry Conditions 33 OK

DC1 up

RCIC1 up 34 CD M-LP-IS-OK-F

P0 up P2 RPVHW

P1 up CVENT 35 CD M-LP-WW-OK-F

DC2 up

DW down 36 CD M-LP-IS-OK-H

WW up RPVHFE

CVENT 37 CD M-LP-WW-OK-H

38 OK

39 CD M-MP-IS-OK-OK

PINJ

CVENT 40 CD M-MP-WW-OK-OK

41 CD M-MP-IS-OK-F

RCIC2 RPVHW

CVENT 42 CD M-MP-WW-OK-F

43 CD M-MP-IS-OK-H

RPVHFE

CVENT 44 CD M-MP-WW-OK-H

Transfer from

Sheet 1 No. Status Plant Damage 

State
A-21

Figure A-21 CDET for DWF venting strategies, Sheet 4, Sequences 46 to 67

Phase 2 

RCIC

RPV 

Pressure for 

LT RCIC

SP Makeup RPV 

Injection

FLEX

Pump

RPV 

Pressure for 

RPV 

Injection

Reclose 

Vents

RCIC2 P2 SPMU RPVINJ PUMP PINJ CVENT

46 OK

47 CD L-MP-IS-LT-F

RPVHW

CVENTnodc 48 CD L-MP-DW-LT-F

49 OK

50 CD L-MP-IS-LT-OK

PINJnodc

CVENTnodc 51 CD L-MP-DW-LT-OK

52 CD L-MP-IS-LT-F

SPHFE RPVHW

CVENTnodc 53 CD L-MP-DW-LT-F

54 CD L-MP-IS-LT-H

RPVHFE

CVENTnodc 55 CD L-MP-DW-LT-H

56 OK

Entry Conditions

DC1 up 57 CD M-LP-IS-LT-F

RCIC1 up P2nodc RPVHW

P0 up CVENTnodc 58 CD M-LP-DW-LT-F

P1 up

DC2 down 59 CD M-LP-IS-LT-H

DW up RPVHFE

CVENTnodc 60 CD M-LP-DW-LT-H

61 OK

62 CD M-MP-IS-LT-OK

PINJnodc

CVENTnodc 63 CD M-MP-DW-LT-OK

64 CD M-MP-IS-LT-F

RCIC2nodc RPVHW

CVENTnodc 65 CD M-MP-DW-LT-F

66 CD M-MP-IS-LT-H

RPVHFE

CVENTnodc 67 CD M-MP-DW-LT-H

Transfer from

Sheet 1 No. Status Plant Damage 

State
A-22

Figure A-22 CDET for DWF venting strategies, Sheet 5, Sequences 68 to 89

Phase 2 

RCIC

RPV 

Pressure for 

LT RCIC

SP Makeup RPV 

Injection

FLEX

Pump

RPV 

Pressure for 

RPV 

Injection

Reclose 

Vents

RCIC2 P2 SPMU RPVINJ PUMP PINJ CVENT

68 OK

69 CD L-MP-IS-LT-F

RPVHW

CVENTnodc 70 CD L-MP-WW-LT-F

71 OK

72 CD L-MP-IS-LT-OK

PINJnodc

CVENTnodc 73 CD L-MP-WW-LT-OK

74 CD L-MP-IS-LT-F

SPHFE RPVHW

CVENTnodc 75 CD L-MP-WW-LT-F

76 CD L-MP-IS-LT-H

RPVHFE

CVENTnodc 77 CD L-MP-WW-LT-H

Entry Conditions 78 OK

DC1 up

RCIC1 up 79 CD M-LP-IS-LT-F

P0 up P2nodc RPVHW

P1 up CVENTnodc 80 CD M-LP-WW-LT-F

DC2 down

DW down 81 CD M-LP-IS-LT-H

WW up RPVHFE

CVENTnodc 82 CD M-LP-WW-LT-H

83 OK

84 CD M-MP-IS-LT-OK

PINJnodc

CVENTnodc 85 CD M-MP-WW-LT-OK

86 CD M-MP-IS-LT-F

RCIC2nodc RPVHW

CVENTnodc 87 CD M-MP-WW-LT-F

88 CD M-MP-IS-LT-H

RPVHFE

CVENTnodc 89 CD M-MP-WW-LT-H

Transfer from

Sheet 1 No. Status Plant Damage 

State
A-23

Figure A-23 CDET for DWF venting strategies, Sheet 6, Sequences 91 to 112

Phase 2 DC Early DW 

Vent

Early WW 

Vent

Phase 2 

RCIC

RPV 

Pressure for 

LT RCIC

SP Makeup RPV 

Injection

FLEX

Pump

RPV 

Pressure for 

RPV 

Injection

Reclose 

Vents

DC2 DW WW RCIC2 P2 SPMU RPVINJ PUMP PINJ CVENT

91 OK

92 CD M-LP-IS-OK-F

RPVHW

CVENT 93 CD M-LP-DW-OK-F

94 CD M-LP-IS-OK-H

RPVHFE

CVENT 95 CD M-LP-DW-OK-H

96 OK

97 CD M-LP-IS-OK-F

RPVHW

CVENT 98 CD M-LP-WW-OK-F

Entry Conditions 99 CD M-LP-IS-OK-H

DC1 up DW RPVHFE

RCIC1 up CVENT 100 CD M-LP-WW-OK-H

P0 up

P1 down WW 101 CD M-LP-IS-OK-XX

102 OK

103 CD M-LP-IS-LT-F

RPVHW

CVENTnodc 104 CD M-LP-DW-LT-F

105 CD M-LP-IS-LT-H

RPVHFE

CVENTnodc 106 CD M-LP-DW-LT-H

DC2

107 OK

108 CD M-LP-IS-LT-F

RPVHW

CVENTnodc 109 CD M-LP-WW-LT-F

110 CD M-LP-IS-LT-H

DWnodc RPVHFE

CVENTnodc 111 CD M-LP-WW-LT-H

WWnodc 112 CD M-LP-IS-LT-XX

Transfer from

Sheet 1 No. Status Plant Damage State
A-24

Figure A-24 CDET for DWF venting strategies, Sheet 7, Sequences 113 to 134

Phase 2 

RCIC

RPV 

Pressure for 

LT RCIC

SP Makeup RPV 

Injection

FLEX

Pump

RPV 

Pressure for 

RPV 

Injection

Reclose 

Vents

RCIC2 P2 SPMU RPVINJ PUMP PINJ CVENT

113 OK

114 CD L-HP-IS-OK-F

RPVHW

CVENT 115 CD L-HP-DW-OK-F

116 OK

117 CD L-HP-IS-OK-OK

PINJ

CVENT 118 CD L-HP-DW-OK-OK

119 CD L-HP-IS-OK-F

SPHFE RPVHW

CVENT 120 CD L-HP-DW-OK-F

121 CD L-HP-IS-OK-H

RPVHFE

CVENT 122 CD L-HP-DW-OK-H

123 OK

Entry Conditions

DC1 up 124 CD M-LP-IS-OK-F

RCIC1 up P2NODP RPVHW

P0 down CVENT 125 CD M-LP-DW-OK-F

P1 up

DC2 up 126 CD M-LP-IS-OK-H

DW up RPVHFE

CVENT 127 CD M-LP-DW-OK-H

128 OK

129 CD M-HP-IS-OK-OK

PINJ

CVENT 130 CD M-HP-DW-OK-OK

131 CD M-HP-IS-OK-F

RCIC2 RPVHW

CVENT 132 CD M-HP-DW-OK-F

133 CD M-HP-IS-OK-H

RPVHFE

CVENT 134 CD M-HP-DW-OK-H

Transfer from

Sheet 1 No. Status Plant Damage 

State
A-25

Figure A-25 CDET for DWF venting strategies, Sheet 8, Sequences 135 to 156

Phase 2 

RCIC

RPV 

Pressure for 

LT RCIC

SP Makeup RPV 

Injection

FLEX

Pump

RPV 

Pressure for 

RPV 

Injection

Reclose 

Vents

RCIC2 P2 SPMU RPVINJ PUMP PINJ CVENT

135 OK

136 CD L-HP-IS-OK-F

RPVHW

CVENT 137 CD L-HP-WW-OK-F

138 OK

139 CD L-HP-IS-OK-OK

PINJ

CVENT 140 CD L-HP-WW-OK-OK

141 CD L-HP-IS-OK-F

SPHFE RPVHW

CVENT 142 CD L-HP-WW-OK-F

143 CD L-HP-IS-OK-H

RPVHFE

CVENT 144 CD L-HP-WW-OK-H

Entry Conditions 145 OK

DC1 up

RCIC1 up 146 CD M-LP-IS-OK-F

P0 down P2NODP RPVHW

P1 up CVENT 147 CD M-LP-WW-OK-F

DC2 up

DW down 148 CD M-LP-IS-OK-H

WW up RPVHFE

CVENT 149 CD M-LP-WW-OK-H

150 OK

151 CD M-HP-IS-OK-OK

PINJ

CVENT 152 CD M-HP-WW-OK-OK

153 CD M-HP-IS-OK-F

RCIC2 RPVHW

CVENT 154 CD M-HP-WW-OK-F

155 CD M-HP-IS-OK-H

RPVHFE

CVENT 156 CD M-HP-WW-OK-H

Transfer from

Sheet 1 No. Status Plant Damage 

State
A-26

Figure A-26 CDET for DWF venting strategies, Sheet 9, Sequences 158 to 179

Phase 2 

RCIC

RPV 

Pressure for 

LT RCIC

SP Makeup RPV 

Injection

FLEX

Pump

RPV 

Pressure for 

RPV 

Injection

Reclose 

Vents

RCIC2 P2 SPMU RPVINJ PUMP PINJ CVENT

158 OK

159 CD L-HP-IS-LT-F

RPVHW

CVENTnodc 160 CD L-HP-DW-LT-F

161 OK

162 CD L-HP-IS-LT-OK

PINJnodc

CVENTnodc 163 CD L-HP-DW-LT-OK

164 CD L-HP-IS-LT-F

SPHFE RPVHW

CVENTnodc 165 CD L-HP-DW-LT-F

166 CD L-HP-IS-LT-H

RPVHFE

CVENTnodc 167 CD L-HP-DW-LT-H

168 OK

Entry Conditions

DC1 up 169 CD M-LP-IS-LT-F

RCIC1 up P2NODPnodc RPVHW

P0 down CVENTnodc 170 CD M-LP-DW-LT-F

P1 up

DC2 down 171 CD M-LP-IS-LT-H

DW up RPVHFE

CVENTnodc 172 CD M-LP-DW-LT-H

173 OK

174 CD M-HP-IS-LT-OK

PINJnodc

CVENTnodc 175 CD M-HP-DW-LT-OK

176 CD M-HP-IS-LT-F

RCIC2nodc RPVHW

CVENTnodc 177 CD M-HP-DW-LT-F

178 CD M-HP-IS-LT-H

RPVHFE

CVENTnodc 179 CD M-HP-DW-LT-H

Transfer from

Sheet 1 No. Status Plant Damage 

State
A-27

Figure A-27 CDET for DWF venting strategies, Sheet 10, Sequences 180 to 201

Phase 2 

RCIC

RPV 

Pressure for 

LT RCIC

SP Makeup RPV 

Injection

FLEX

Pump

RPV 

Pressure for 

RPV 

Injection

Reclose 

Vents

RCIC2 P2 SPMU RPVINJ PUMP PINJ CVENT

180 OK

181 CD L-HP-IS-LT-F

RPVHW

CVENTnodc 182 CD L-HP-WW-LT-F

183 OK

184 CD L-HP-IS-LT-OK

PINJnodc

CVENTnodc 185 CD L-HP-WW-LT-OK

186 CD L-HP-IS-LT-F

SPHFE RPVHW

CVENTnodc 187 CD L-HP-WW-LT-F

188 CD L-HP-IS-LT-H

RPVHFE

CVENTnodc 189 CD L-HP-WW-LT-H

Entry Conditions 190 OK

DC1 up

RCIC1 up 191 CD M-LP-IS-LT-F

P0 down P2NODPnodc RPVHW

P1 up CVENTnodc 192 CD M-LP-WW-LT-F

DC2 down

DW down 193 CD M-LP-IS-LT-H

WW up RPVHFE

CVENTnodc 194 CD M-LP-WW-LT-H

195 OK

196 CD M-HP-IS-LT-OK

PINJnodc

CVENTnodc 197 CD M-HP-WW-LT-OK

198 CD M-HP-IS-LT-F

RCIC2nodc RPVHW

CVENTnodc 199 CD M-HP-WW-LT-F

200 CD M-HP-IS-LT-H

RPVHFE

CVENTnodc 201 CD M-HP-WW-LT-H

Transfer from

Sheet 1 No. Status Plant Damage 

State
A-28

Figure A-28 CDET for DWF venting strategies, Sheet 11, Sequences 203 to 224

Phase 2 DC Early DW 

Vent

Early WW 

Vent

Phase 2 

RCIC

RPV 

Pressure for 

LT RCIC

SP Makeup RPV 

Injection

FLEX

Pump

RPV 

Pressure for 

RPV 

Injection

Reclose 

Vents

DC2 DW WW RCIC2 P2 SPMU RPVINJ PUMP PINJ CVENT

203 OK

204 CD M-LP-IS-OK-F

RPVHW

CVENT 205 CD M-LP-DW-OK-F

206 CD M-LP-IS-OK-H

RPVHFE

CVENT 207 CD M-LP-DW-OK-H

208 OK

209 CD M-LP-IS-OK-F

RPVHW

CVENT 210 CD M-LP-WW-OK-F

Entry Conditions 211 CD M-LP-IS-OK-H

DC1 up DW RPVHFE

RCIC1 up CVENT 212 CD M-LP-WW-OK-H

P0 down

P1 down WW 213 CD M-LP-IS-OK-XX

214 OK

215 CD M-LP-IS-LT-F

RPVHW

CVENTnodc 216 CD M-LP-DW-LT-F

217 CD M-LP-IS-LT-H

RPVHFE

CVENTnodc 218 CD M-LP-DW-LT-H

DC2

219 OK

220 CD M-LP-IS-LT-F

RPVHW

CVENTnodc 221 CD M-LP-WW-LT-F

222 CD M-LP-IS-LT-H

DWnodc RPVHFE

CVENTnodc 223 CD M-LP-WW-LT-H

WWnodc 224 CD M-LP-IS-LT-XX

Transfer from

Sheet 1 No. Status Plant Damage State
A-29

Figure A-29 CDET for DWF venting strategies, Sheet 12, Sequences 227 to 248

Phase 2 

RCIC

RPV 

Pressure for 

LT RCIC

SP Makeup RPV 

Injection

FLEX

Pump

RPV 

Pressure for 

RPV 

Injection

Reclose 

Vents

RCIC2 P2 SPMU RPVINJ PUMP PINJ CVENT

227 OK

228 CD L-MP-IS-ST-F

RPVHW

CVENTnodc 229 CD L-MP-DW-ST-F

230 OK

231 CD L-MP-IS-ST-OK

PINJnodc

CVENTnodc 232 CD L-MP-DW-ST-OK

233 CD L-MP-IS-ST-F

SPHFE RPVHW

CVENTnodc 234 CD L-MP-DW-ST-F

235 CD L-MP-IS-ST-H

RPVHFE

CVENTnodc 236 CD L-MP-DW-ST-H

237 OK

Entry Conditions 238 CD M-LP-IS-ST-F

DC1 down P2nodc RPVHW

RCIC1 up CVENTnodc 239 CD M-LP-DW-ST-F

P0 up

P1 up 240 CD M-LP-IS-ST-H

DW up RPVHFE

CVENTnodc 241 CD M-LP-DW-ST-H

242 OK

243 CD M-MP-IS-ST-OK

PINJnodc

CVENTnodc 244 CD M-MP-DW-ST-OK

245 CD M-MP-IS-ST-F

RCIC2nodc RPVHW

CVENTnodc 246 CD M-MP-DW-ST-F

247 CD M-MP-IS-ST-H

RPVHFE

CVENTnodc 248 CD M-MP-DW-ST-H

Transfer from

Sheet 1 No. Status Plant Damage 

State
A-30

Figure A-30 CDET for DWF venting strategies, Sheet 13, Sequences 249 to 270

Phase 2 

RCIC

RPV 

Pressure for 

LT RCIC

SP Makeup RPV 

Injection

FLEX

Pump

RPV 

Pressure for 

RPV 

Injection

Reclose 

Vents

RCIC2 P2 SPMU RPVINJ PUMP PINJ CVENT

249 OK

250 CD L-MP-IS-ST-F

RPVHW

CVENTnodc 251 CD L-MP-WW-ST-F

252 OK

253 CD L-MP-IS-ST-OK

PINJnodc

CVENTnodc 254 CD L-MP-WW-ST-OK

255 CD L-MP-IS-ST-F

SPHFE RPVHW

CVENTnodc 256 CD L-MP-WW-ST-F

257 CD L-MP-IS-ST-H

RPVHFE

CVENTnodc 258 CD L-MP-WW-ST-H

259 OK

Entry Conditions

DC1 down 260 CD M-LP-IS-ST-F

RCIC1 up P2nodc RPVHW

P0 up CVENTnodc 261 CD M-LP-WW-ST-F

P1 up

DW down 262 CD M-LP-IS-ST-H

WW up RPVHFE

CVENTnodc 263 CD M-LP-WW-ST-H

264 OK

265 CD M-MP-IS-ST-OK

PINJnodc

CVENTnodc 266 CD M-MP-WW-ST-OK

267 CD M-MP-IS-ST-F

RCIC2nodc RPVHW

CVENTnodc 268 CD M-MP-WW-ST-F

269 CD M-MP-IS-ST-H

RPVHFE

CVENTnodc 270 CD M-MP-WW-ST-H

Transfer from

Sheet 1 No. Status Plant Damage 

State
A-31

Figure A-31 CDET for DWF venting strategies, Sheet 14, Sequences 272 to 282

Phase 2 DC Early DW 

Vent

Early WW 

Vent

Phase 2 

RCIC

RPV 

Pressure for 

LT RCIC

SP Makeup RPV 

Injection

FLEX

Pump

RPV 

Pressure for 

RPV 

Injection

Reclose 

Vents

DC2 DW WW RCIC2 P2 SPMU RPVINJ PUMP PINJ CVENT

272 OK

273 CD M-LP-IS-ST-F

RPVHW

CVENTnodc 274 CD M-LP-WW-ST-F

Entry Conditions

DC1 down 275 CD M-LP-IS-ST-H

RCIC1 up RPVHFE

P0 up CVENTnodc 276 CD M-LP-WW-ST-H

P1 down

277 OK

278 CD M-LP-IS-ST-F

RPVHW

CVENTnodc 279 CD M-LP-DW-ST-F

280 CD M-LP-IS-ST-H

DWnodc RPVHFE

CVENTnodc 281 CD M-LP-DW-ST-H

WWnodc 282 CD M-LP-IS-ST-XX

Transfer from

Sheet 1 No. Status Plant Damage 

State
A-32

Figure A-32 CDET for DWF venting strategies, Sheet 15, Sequences 283 to 304

Phase 2 

RCIC

RPV 

Pressure for 

LT RCIC

SP Makeup RPV 

Injection

FLEX

Pump

RPV 

Pressure for 

RPV 

Injection

Reclose 

Vents

RCIC2 P2 SPMU RPVINJ PUMP PINJ CVENT

283 OK

284 CD L-HP-IS-ST-F

RPVHW

CVENTnodc 285 CD L-HP-DW-ST-F

286 OK

287 CD L-HP-IS-ST-OK

PINJnodc

CVENTnodc 288 CD L-HP-DW-ST-OK

289 CD L-HP-IS-ST-F

SPHFE RPVHW

CVENTnodc 290 CD L-HP-DW-ST-F

291 CD L-HP-IS-ST-H

RPVHFE

CVENTnodc 292 CD L-HP-DW-ST-H

293 OK

Entry Conditions 294 CD M-LP-IS-ST-F

DC1 down P2NODPnodc RPVHW

RCIC1 up CVENTnodc 295 CD M-LP-DW-ST-F

P0 down

P1 up 296 CD M-LP-IS-ST-H

DW up RPVHFE

CVENTnodc 297 CD M-LP-DW-ST-H

298 OK

299 CD M-HP-IS-ST-OK

PINJnodc

CVENTnodc 300 CD M-HP-DW-ST-OK

301 CD M-HP-IS-ST-F

RCIC2nodc RPVHW

CVENTnodc 302 CD M-HP-DW-ST-F

303 CD M-HP-IS-ST-H

RPVHFE

CVENTnodc 304 CD M-HP-DW-ST-H

Transfer from

Sheet 1 No. Status Plant Damage 

State
A-33

Figure A-33 CDET for DWF venting strategies, Sheet 16, Sequences 305 to 326

Phase 2 

RCIC

RPV 

Pressure for 

LT RCIC

SP Makeup RPV 

Injection

FLEX

Pump

RPV 

Pressure for 

RPV 

Injection

Reclose 

Vents

RCIC2 P2 SPMU RPVINJ PUMP PINJ CVENT

305 OK

306 CD L-HP-IS-ST-F

RPVHW

CVENTnodc 307 CD L-HP-WW-ST-F

308 OK

309 CD L-HP-IS-ST-OK

PINJnodc

CVENTnodc 310 CD L-HP-WW-ST-OK

311 CD L-HP-IS-ST-F

SPHFE RPVHW

CVENTnodc 312 CD L-HP-WW-ST-F

313 CD L-HP-IS-ST-H

RPVHFE

CVENTnodc 314 CD L-HP-WW-ST-H

315 OK

Entry Conditions

DC1 down 316 CD M-LP-IS-ST-F

RCIC1 up P2NODPnodc RPVHW

P0 down CVENTnodc 317 CD M-LP-WW-ST-F

P1 up

DW down 318 CD M-LP-IS-ST-H

WW up RPVHFE

CVENTnodc 319 CD M-LP-WW-ST-H

320 OK

321 CD M-HP-IS-ST-OK

PINJnodc

CVENTnodc 322 CD M-HP-WW-ST-OK

323 CD M-HP-IS-ST-F

RCIC2nodc RPVHW

CVENTnodc 324 CD M-HP-WW-ST-F

325 CD M-HP-IS-ST-H

RPVHFE

CVENTnodc 326 CD M-HP-WW-ST-H

Transfer from

Sheet 1 No. Status Plant Damage 

State
A-34

Figure A-34 CDET for DWF venting strategies, Sheet 17, Sequences 328 to 338

Phase 2 DC Early DW 

Vent

Early WW 

Vent

Phase 2 

RCIC

RPV 

Pressure for 

LT RCIC

SP Makeup RPV 

Injection

FLEX

Pump

RPV 

Pressure for 

RPV 

Injection

Reclose 

Vents

DC2 DW WW RCIC2 P2 SPMU RPVINJ PUMP PINJ CVENT

328 OK

329 CD M-LP-IS-ST-F

RPVHW

CVENTnodc 330 CD M-LP-DW-ST-F

Entry Conditions

DC1 down 331 CD M-LP-IS-ST-H

RCIC1 up RPVHFE

P0 down CVENTnodc 332 CD M-LP-DW-ST-H

P1 down

333 OK

334 CD M-LP-IS-ST-F

RPVHW

CVENTnodc 335 CD M-LP-WW-ST-F

336 CD M-LP-IS-ST-H

WWnodc RPVHFE

CVENTnodc 337 CD M-LP-WW-ST-H

DWnodc 338 CD M-LP-IS-ST-XX

Transfer from

Sheet 1 No. Status Plant Damage 

State
B-1

APPENDIX B 

CORE DAMAGE EVENT TREE SUPPORTING INFORMATION

Table B-1 Core-damage event tree top logic

Top Event Description Boolean Expression 

Probability 

WWF

DWF 

passive 

DWF 

manual 

CVENT failure to reclose containment 

vents 

= HFE-CVENT 0.1 1 0.1 

CVENTnodc failure to reclose containment 

vents (no DC power) 

= HFE-CVENT-NODC 0.3 1 0.3 

DC1 DC power fails short-term = BAT-ST 1.2E-06

DC2 DC power fails long-term = + GEN-FTS HFE-GEN 

+ GENFTR

3.9E-01

DW failure to open drywell vent = HFE-DW 0.1 0.01 0.1 

DWnodc failure to open drywell vent (no 

DC power) 

= HFE-DW-NODC 0.3 0.01 0.3 

P0DP failure to depressurize to 200-

400 psia 

= HFE-DEPRESS 0.1 

P0DPnodc failure to depressurize to 200-

400 psia (no DC power) 

= HFE-DEPRESS-NODC 0.3 

P0MSCL failure to depressurize at 

minimum steam cooling level 

= HFE-MSCL 0.1 

P0MSCLnodc failure to depressurize at 

minimum steam cooling level (no 

DC power) 

= HFE-MSCL-NODC 0.3 

P1 RPV pressure inadequate for 

RCIC short-term 

= SORV1-DP

+ HFE-INVDP

1.3E-01

P1nodc RPV pressure inadequate for 

RCIC short-term (no DC power) 

= SORV1-DP

+ HFE-INVDP-NODC

3.2E-01

P1NODP RPV pressure inadequate to 

RCIC short-term (no depress) 

= SORV1-NODP 2.9E-01

P1NODPnodc RPV pressure inadequate to 

RCIC short-term (no depress or 

DC power) 

= SORV1-NODP 2.9E-01

P2 RPV pressure inadequate for 

RCIC long-term 

= SORV2-DP

+ HFE-INVDP

1.4E-01

P2nodc RPV pressure inadequate for 

RCIC long-term (no DC power) 

= SORV2-DP

+ HFE-INVDP-NODC

3.3E-01

P2NODP RPV pressure inadequate to 

RCIC long-term (no depress) 

= SORV2-NODP 8.2E-02

P2NODPnodc RPV pressure inadequate to 

RCIC long-term (no depress or 

DC power) 

= SORV2-NODP 8.2E-02

PINJ failure to depressurize RPV for 

FLEX injection 

= HFE-PINJ 0.1 

PINJnodc failure to depressurize RPV for 

FLEX injection (no DC power) 

= HFE-PINJ-NODC 0.3 
B-2

RCIC1 RCIC fails short-term = RCIC-FTS

+ RCIC-1

8.9E-03

RCIC1nodc RCIC fails short-term (no DC 

power)

= RCIC-FTS

+ RCIC-1

+ HFE-RCIC-BLACKSTART

3.1E-01

RCIC2 RCIC fails long-term = RCIC-2 6.5E-04

RCIC2nodc RCIC fails long-term (no DC 

power)

= RCIC-2

+HFE- RCIC-BLACKRUN

3.0E-01

RPVHFE operator fails to align RPV 

injection using FLEX

= HFE-RPV 0.3

RPVHW FLEX pump fails during RPV 

injection

= FLEX-FTS

+ FLEXP-FTR

1.50E-01

SPHFE operator fails to align SP makeup = HFE-SP 0.3

SPHW FLEX pump fails while supplying 

SP makeup

= FLEXP-FTS

+ FLEXP-FTR

1.5E-01

WW failure to open wetwell vent = HFE-WW 0.1

WWnodc failure to open wetwell vent (no 

DC power)

= HFE-WW-NODC 0.3

Table B-2 Core-damage event tree human failure events

HFE-CVENT operator fails to reclose the containment vents 

upon CD 0.1 1 0.1

HFE-CVENT-NODC operator fails to reclose the containment vents 

upon CD (no DC power) 0.3 1 0.3

HFE-DEPRESS operator fails to depressurize to 200-400 psia 0.1

HFE-DEPRESS-NODC operator fails to depressurize to 200-400 psia 

(no DC power) 0.3

HFE-DW operator fails to open the drywell vent 0.1 0.01 0.1

HFE-DW-NODC operator fails to open the drywell vent

(no DC power) 0.3 0.01 0.3

HFE-GEN operator fails to align portable generator 0.3

HFE-INVDP operator inadvertantly depressurizes RPV

below RCIC 0.1

HFE-INVDP-NODC operator inadvertantly depressurizes RPV 

below RCIC (no DC power) 0.3

HFE-MSCL operator fails to depressurize at minimum

steam cooling level 0.1

HFE-MSCL-NODC operator fails to depressurize at minimum

steam cooling level (no DC power) 0.3

HFE-PINJ operator fails to depressurize for FLEX

RPV injection 0.1

HFE-PINJ-NODC operator fails to depressurize for FLEX RPV 

injection (no DC power) 0.3

HFE-RCIC-BLACKRUN operator fails to blackrun RCIC during Phase 

2 0.3

HFE-RCIC-BLACKSTART operator fails to blackstart and blackrun RCIC 

during Phase 1 0.3

Top Event Description Boolean Expression 

Probability 

WWF 

DWF 

passive 

DWF 

manual 

Human Failure Event Description 

Probability 

WWF DWF 

passive 

DWF 

manual 
B-3

HFE-RPV operator fails to align FLEX pump for

RPV injection 0.3

HFE-SP operator fails to align FLEX pump for

SP makeup 0.3

HFE-WW operator fails to open the wetwell vent 0.1

HFE-WW-NODC operator fails to open the wetwell vent

(no DC power) 0.3

Table B-3 Core-damage event tree hardware-related basic events

Human Failure Event Description 

Probability 

WWF DWF 

passive 

DWF 

manual 

Name Description Probability Calculation Details 

BAT-ST station batteries fail 

short-term 2.3E-06

P=1-exp(-λT) 

λ = 5.9E-07 per hour 

 Basis: SPAR 2010 Parameter Estimation Update 

[X] 

T = 2 hours 

Basis: Phase 1 mission time 

FLEXP-FTR FLEX pump fails to run 1.4E-01

P=1-exp(-λ1 – λ2T)

λ1 = 1.3E-03 /h = failure rate < 1 h 

 Basis: SPAR 2010 Parameter Estimation Update 

λ2 = 2.3E-03 /h = failure rate > 1 h 

 Basis: SPAR 2010 Parameter Estimation Update 

T = 67 h 

Basis: Phase 2 mission time – 1 h 

FLEXP-FTS FLEX pump fails to start 5.1E-03 Failure probability per demand 

Basis: SPAR 2010 Parameter Estimation Update 

GEN-FTR portable generator 

fails to run 8.5E-02

P=1-exp(-λT) 

λ = 1.3E-03 per hour 

 Basis: SPAR 2010 Parameter Estimation Update 

T = 68 hours 

Basis: Phase 2 mission time 

GEN-FTS portable generator 

fails to start 4.3E-02 Failure probability per demand Basis: SPAR 2010 Parameter Estimation Update 

RCIC-1 RCIC pump fails to run 

in Phase 1 3.7E-05

P=1-exp(-λT) 

λ = 9.3E-06 per hour 

 Basis: SPAR 2010 Parameter Estimation Update 

T = 2 hours 

Basis: Phase 1 mission time 

RCIC-2 RCIC pump fails to run 

in Phase 2 6.4E-04

P=1-exp(-λT) 

λ = 9.3E-06 per hour 

 Basis: SPAR 2010 Parameter Estimation Update 

T = 68 hours 

Basis: Phase 2 mission time 

RCIC-FTS RCIC fails to start 8.9E-03 Failure probability per demand 

Basis: SPAR 2010 Parameter Estimation Update 

SORV1-DP

stuck-open SRV given 

Phase 1 depressurization 

and cooldown 

3.4E-02

P=1-(1-pD)

n

pD = 8.6E-04 per demand 

 Basis: SPAR 2010 Parameter Estimation Update 

n = 40 demands 

Basis: thermal-hydraulic calculations 

SORV1-NODP 

stuck-open SRV given no 

Phase 1 depressurization 

and cooldown 

2.9E-01

P=1-(1-pD)

n

pD = 8.6E-04 per demand 

 Basis: SPAR 2010 Parameter Estimation Update 

n = 400 demands 

Basis: thermal-hydraulic calculations 
B-4

SORV2-DP

stuck-open SRV in Phase 

2 given Phase 1 

depressurization and 

cooldown 

4.2E-02

P=1-(1-pD)

n

pD = 8.6E-04 per demand 

 Basis: SPAR 2010 Parameter Estimation Update 

n = 50 demands 

Basis: thermal-hydraulic calculations 

SORV2-NODP 

stuck-open SRV in Phase 

2 given no Phase 1 

depressurization and 

cooldown 

8.2E-02

P=1-(1-pD)

n

pD = 8.6E-04 per demand 

 Basis: SPAR 2010 Parameter Estimation Update 

n = 100 demands 

Basis: thermal-hydraulic calculations 

Name Description Probability Calculation Details 
C-1

APPENDIX C 

ACCIDENT PROGRESSION EVENT TREES

Figure C-1 APET for RPV injection and WWF venting strategies (APETs 1 and 2), Sheet 1, 

Sequences 1 to 42

PDS Transfer 

from CD 

Event Tree

restore DC 

power 

before core 

relocation

RPV 

injection 

before core 

relocation - 

human

RPV 

injection 

before core 

relocation - 

pump

depressurize 

before core 

relocation

RPV 

injection 

before LMT - 

human

RPV 

injection 

before LMT - 

pump

SRV fails 

open 

(random)

MSL Creep 

Rupture

Late WW 

Vent

Late DW 

Vent

PDS DCCR RPVINJCRH RPVINJCRP PRPVINJCR RPVINJLMTH RPVINJLMTP PPCD MSLCR WWPCD DWPCD

1 SRV WW IVR

depressurized

2 SRV DW IVR

WWPCD

DWPCD 3 SRV OP IVR

works 4 SRV WW IVR

fails open

5 SRV DW IVR

WWPCD

DWPCD 6 SRV OP IVR

PRPVINJCR

7 HP WW EVR

no

8 HP DW EVR

WWPCD

cycling DWPCD 9 HP OP EVR

aligned

10 MSLCR WW IVR

yes

11 MSLCR DW IVR

WWPCD

DWPCD 12 MSLCR OP IVR

13 SRV WW LMT

fails open

14 SRV DW LMT

WWPCD

DWPCD 15 SRV OP LMT

failed 16 HP WW LMT

no

17 HP DW LMT

WWPCD

cycling DWPCD 18 HP OP LMT

19 MSLCR WW LMT

yes

20 MSLCR DW LMT

WWPCD

DWPCD 21 MSLCR OP LMT

22 SRV WW EVR

works

23 SRV DW EVR

WWPCD

DWPCD 24 SRV OP EVR

aligned 25 SRV WW LMT

fails open

26 SRV DW LMT

WWPCD

DWPCD 27 SRV OP LMT

failed

28 HP WW LMT

no

29 HP DW LMT

WWPCD

cycling DWPCD 30 HP OP LMT

not aligned

31 MSLCR WW LMT

yes

32 MSLCR DW LMT

WWPCD

DWPCD 33 MSLCR OP LMT

34 SRV WW LMT

fails open

35 SRV DW LMT

WWPCD

DWPCD 36 SRV OP LMT

not aligned

37 HP WW LMT

no

38 HP DW LMT

WWPCD

cycling DWPCD 39 HP OP LMT

40 MSLCR WW LMT

yes

41 MSLCR DW LMT

WWPCD

DWPCD 42 MSLCR OP LMT

no post-accident DC to Sheet 2, Seqs. 43 to 84

No.

RPV 

Depress 

Mode

Cont. 

Pressure 

Control

Core 

Debris 

Location
C-2

Figure C-2 APET for RPV injection and WWF venting strategies (APETs 1 and 2), Sheet 2, 

Sequences 43 to 84

RPV 

injection 

before core 

relocation - 

human

RPV 

injection 

before core 

relocation - 

pump

depressurize 

before core 

relocation

RPV 

injection 

before LMT - 

human

RPV 

injection 

before LMT - 

pump

SRV fails 

open 

(random)

MSL Creep 

Rupture

Late WW 

Vent

Late DW 

Vent

RPVINJCRH RPVINJCRP PRPVINJCR RPVINJLMTH RPVINJLMTP PPCD MSLCR WWPCD DWPCD

43 SRV WW IVR

depressurized

44 SRV DW IVR

WWPCDnodc

DWPCDnodc 45 SRV OP IVR

works 46 SRV WW IVR

fails open

47 SRV DW IVR

WWPCDnodc

DWPCDnodc 48 SRV OP IVR

PRPVINJCRnodc

49 HP WW EVR

no

50 HP DW EVR

WWPCDnodc

cycling DWPCDnodc 51 HP OP EVR

aligned

52 MSLCR WW IVR

yes

53 MSLCR DW IVR

WWPCDnodc

DWPCDnodc 54 MSLCR OP IVR

55 SRV WW LMT

fails open

56 SRV DW LMT

WWPCDnodc

DWPCDnodc 57 SRV OP LMT

failed 58 HP WW LMT

no

59 HP DW LMT

WWPCDnodc

cycling DWPCDnodc 60 HP OP LMT

61 MSLCR WW LMT

yes

62 MSLCR DW LMT

WWPCDnodc

DWPCDnodc 63 MSLCR OP LMT

64 SRV WW EVR

works

65 SRV DW EVR

WWPCDnodc

DWPCDnodc 66 SRV OP EVR

aligned 67 SRV WW LMT

fails open

68 SRV DW LMT

WWPCDnodc

DWPCDnodc 69 SRV OP LMT

failed

70 HP WW LMT

no

71 HP DW LMT

WWPCDnodc

cycling DWPCDnodc 72 HP OP LMT

not aligned

73 MSLCR WW LMT

yes

74 MSLCR DW LMT

WWPCDnodc

DWPCDnodc 75 MSLCR OP LMT

76 SRV WW LMT

fails open

77 SRV DW LMT

WWPCDnodc

DWPCDnodc 78 SRV OP LMT

not aligned

79 HP WW LMT

no

80 HP DW LMT

WWPCDnodc

cycling DWPCDnodc 81 HP OP LMT

82 MSLCR WW LMT

yes

83 MSLCR DW LMT

WWPCDnodc

DWPCDnodc 84 MSLCR OP LMT

No.

RPV 

Depress 

Mode

Cont. 

Pressure 

Control

Core 

Debris 

Location

Transfer

from

Sheet 1
C-3

Figure C-3 APET for DW Injection and WWF Venting Strategies (APET 3), Sheet 1, 

Sequences 1 to 36

PDS Transfer 

from CD 

Event Tree

restore DC 

power 

before core 

relocation

DW injection 

- human

DW injection 

- pump

post-CD 

depress

SRV fails 

open 

(random)

MSL Creep 

Rupture

Late WW 

Vent

Late DW 

Vent

PDS DCCR DWINJH DWINJP PDWINJPCD PPCD MSLCR WWPCD DWPCD

1 SRV WW EVR

depressurized

2 SRV DW EVR

WWPCD

DWPCD 3 SRV OP EVR

4 HP WW EVR

works no

5 HP DW EVR

WWPCD

cycling DWPCD 6 HP OP EVR

aligned

7 MSLCR WW EVR

yes

PDWINJPCD 8 MSLCR DW EVR

WWPCD

DWPCD 9 MSLCR OP EVR

10 SRV WW EVR

fails open

11 SRV DW EVR

WWPCD

DWPCD 12 SRV OP EVR

13 SRV WW LMT

depressurized

14 SRV DW LMT

WWPCD

DWPCD 15 SRV OP LMT

16 HP WW LMT

failed no

17 HP DW LMT

WWPCD

cycling DWPCD 18 HP OP LMT

19 MSLCR WW LMT

yes

PDWINJPCD 20 MSLCR DW LMT

WWPCD

DWPCD 21 MSLCR OP LMT

22 SRV WW LMT

fails open

23 SRV DW LMT

WWPCD

DWPCD 24 SRV OP LMT

25 SRV WW LMT

depressurized

26 SRV DW LMT

WWPCD

DWPCD 27 SRV OP LMT

28 HP WW LMT

not aligned no

29 HP DW LMT

WWPCD

cycling DWPCD 30 HP OP LMT

31 MSLCR WW LMT

yes

PDWINJPCD 32 MSLCR DW LMT

WWPCD

DWPCD 33 MSLCR OP LMT

34 SRV WW LMT

fails open

35 SRV DW LMT

WWPCD

DWPCD 36 SRV OP LMT

no post-accident DC to Sheet 2, Seqs. 37 to 72

No.

RPV 

Depress 

Mode

Cont. 

Pressure 

Control

Core 

Debris 

Location
C-4

Figure C-4 APET for DW Injection and WWF Venting Strategies (APET 3), Sheet 2, 

Sequences 37 to 72

DW injection 

- human

DW injection 

- pump

post-CD 

depress

SRV fails 

open 

(random)

MSL Creep 

Rupture

Late WW 

Vent

Late DW 

Vent

DWINJH DWINJP PDWINJPCD PPCD MSLCR WWPCD DWPCD

37 SRV WW EVR

depressurized

38 SRV DW EVR

WWPCDnodc

DWPCDnodc 39 SRV OP EVR

40 HP WW EVR

works no

41 HP DW EVR

WWPCDnodc

cycling DWPCDnodc 42 HP OP EVR

aligned

43 MSLCR WW EVR

yes

PDWINJPCDnodc 44 MSLCR DW EVR

WWPCDnodc

DWPCDnodc 45 MSLCR OP EVR

46 SRV WW EVR

fails open

47 SRV DW EVR

WWPCDnodc

DWPCDnodc 48 SRV OP EVR

49 SRV WW LMT

depressurized

50 SRV DW LMT

WWPCDnodc

DWPCDnodc 51 SRV OP LMT

52 HP WW LMT

failed no

53 HP DW LMT

WWPCDnodc

cycling DWPCDnodc 54 HP OP LMT

55 MSLCR WW LMT

yes

PDWINJPCDnodc 56 MSLCR DW LMT

WWPCDnodc

DWPCDnodc 57 MSLCR OP LMT

58 SRV WW LMT

fails open

59 SRV DW LMT

WWPCDnodc

DWPCDnodc 60 SRV OP LMT

61 SRV WW LMT

depressurized

62 SRV DW LMT

WWPCDnodc

DWPCDnodc 63 SRV OP LMT

64 HP WW LMT

not aligned no

65 HP DW LMT

WWPCDnodc

cycling DWPCDnodc 66 HP OP LMT

67 MSLCR WW LMT

yes

PDWINJPCDnodc 68 MSLCR DW LMT

WWPCDnodc

DWPCDnodc 69 MSLCR OP LMT

70 SRV WW LMT

fails open

71 SRV DW LMT

WWPCDnodc

DWPCDnodc 72 SRV OP LMT

No.

RPV 

Depress 

Mode

Cont. 

Pressure 

Control

Core 

Debris 

Location

Trnasfer

from

Sheet1
C-5

Figure C-5 APET for DW Injection and DWF Venting Strategies (APETs 4 to 6), Sheet 1, 

Sequences 1 to 36

PDS Transfer 

from CD 

Event Tree

restore DC 

power 

before core 

relocation

DW injection 

- human

DW injection 

- pump

post-CD 

depress

SRV fails 

open 

(random)

MSL Creep 

Rupture

Late DW 

Vent

Late WW 

Vent

PDS DCCR DWINJH DWINJP PDWINJPCD PPCD MSLCR DWPCD WWPCD

1 SRV DW EVR

depressurized

2 SRV WW EVR

DWPCD

WWPCD 3 SRV OP EVR

4 HP DW EVR

works no

5 HP WW EVR

DWPCD

cycling WWPCD 6 HP OP EVR

aligned

7 MSLCR DW EVR

yes

PDWINJPCD 8 MSLCR WW EVR

DWPCD

WWPCD 9 MSLCR OP EVR

10 SRV DW EVR

fails open

11 SRV WW EVR

DWPCD

WWPCD 12 SRV OP EVR

13 SRV DW LMT

depressurized

14 SRV WW LMT

DWPCD

WWPCD 15 SRV OP LMT

16 HP DW LMT

failed no

17 HP WW LMT

DWPCD

cycling WWPCD 18 HP OP LMT

19 MSLCR DW LMT

yes

PDWINJPCD 20 MSLCR WW LMT

DWPCD

WWPCD 21 MSLCR OP LMT

22 SRV DW LMT

fails open

23 SRV WW LMT

DWPCD

WWPCD 24 SRV OP LMT

25 SRV DW LMT

depressurized

26 SRV WW LMT

DWPCD

WWPCD 27 SRV OP LMT

28 HP DW LMT

not aligned no

29 HP WW LMT

DWPCD

cycling WWPCD 30 HP OP LMT

31 MSLCR DW LMT

yes

PDWINJPCD 32 MSLCR WW LMT

DWPCD

WWPCD 33 MSLCR OP LMT

34 SRV DW LMT

fails open

35 SRV WW LMT

DWPCD

WWPCD 36 SRV OP LMT

no post-accident DC

No.

RPV 

Depress 

Mode

Cont. 

Pressure 

Control

Core 

Debris 

Location
C-6

Figure C-6 APET for DW Injection and DWF Venting Strategies (APETs 4 to 6), Sheet 2, 

Sequences 37 to 72 accident progression event tree supporting information

DW injection 

- human

DW injection 

- pump

post-CD 

depress

SRV fails 

open 

(random)

MSL Creep 

Rupture

Late DW 

Vent

Late WW 

Vent

DWINJH DWINJP PDWINJPCD PPCD MSLCR DWPCD WWPCD

37 SRV DW EVR

depressurized

38 SRV WW EVR

DWPCDnodc

WWPCDnodc 39 SRV OP EVR

40 HP DW EVR

works no

41 HP WW EVR

DWPCDnodc

cycling WWPCDnodc 42 HP OP EVR

aligned

43 MSLCR DW EVR

yes

PDWINJPCDnodc 44 MSLCR WW EVR

DWPCDnodc

WWPCDnodc 45 MSLCR OP EVR

46 SRV DW EVR

fails open

47 SRV WW EVR

DWPCDnodc

WWPCDnodc 48 SRV OP EVR

49 SRV DW LMT

depressurized

50 SRV WW LMT

DWPCDnodc

WWPCDnodc 51 SRV OP LMT

52 HP DW LMT

failed no

53 HP WW LMT

DWPCDnodc

cycling WWPCDnodc 54 HP OP LMT

55 MSLCR DW LMT

yes

PDWINJPCDnodc 56 MSLCR WW LMT

DWPCDnodc

WWPCDnodc 57 MSLCR OP LMT

58 SRV DW LMT

fails open

59 SRV WW LMT

DWPCDnodc

WWPCDnodc 60 SRV OP LMT

61 SRV DW LMT

depressurized

62 SRV WW LMT

DWPCDnodc

WWPCDnodc 63 SRV OP LMT

64 HP DW LMT

not aligned no

65 HP WW LMT

DWPCDnodc

cycling WWPCDnodc 66 HP OP LMT

67 MSLCR DW LMT

yes

PDWINJPCDnodc 68 MSLCR WW LMT

DWPCDnodc

WWPCDnodc 69 MSLCR OP LMT

70 SRV DW LMT

fails open

71 SRV WW LMT

DWPCDnodc

WWPCDnodc 72 SRV OP LMT

No.

RPV 

Depress 

Mode

Cont. 

Pressure 

Control

Core 

Debris 

Location

Transfer

from

Sheet 1
D-1

Table D-1 provides the down-branch probabilities used to quantify each accident progression 

event tree (APET). The six APETs (APET-1 through APET-6) were developed using a common 

set of APET headers (listed in the first column of Table C-1), although a specific APET may not 

use some of the APET common headers. For example, APET header RPVINJCRP (align the 

portable pump to provide post-accident water injection via the RPV prior to core relocation) only

appears in APET-1 and APET-2. Similarly, APET header DWINJP (align the portable pump to

provide post-accident water injection via the drywell) only appears in APET-3, APET-4, APET-5, 

and APET-6. 

The APET down-branch probabilities depend on which plant damage state (PDS) is input to the

APET. The second column of Table C-1 indicates which PDS attribute controls the down-branch 

probability, and the fourth column indicates the down-branch probability used for specific values of 

the controlling PDS attribute. For example, if the PDS E-LP-IS-ST-XX is input to APET-2, then the 

following down-branch probabilities are used to quantify APET-2: 

Pr{DCRR} = 1

Pr{DWPCD} = 0.1

Pr{DWPCDnodc} = 0.3

Pr{MSLCR} = 0

Pr{PPCD} = 0

Pr{PRPVINJCR} = 0

Pr{PRPVINJCRnodc} = 0

Pr{RPVINJCRH} = 0.3

Pr{RPVINJCRP} = 0.147

Pr{RPVINJLMTH} = 0.3

Pr{RPVINJLMTP} = 0.147

Pr{WWPCD} = 0.1

Pr{WWPCDnodc} = 0.3

APPENDIX D

ACCIDENT PROGRESSION EVENT TREE SUPPORTING INFORMATION
D-2

APET 

Heading 

PDS 

Attribute 

APET DownBranch Identifier 

PDS 

Attribute 

Value 

APET1 APET2 APET3 APET4 APET5 APET6 

DCCR DC 

status 

DCRR OK 0 0 0 0 0 0 

ST 1 1 1 1 1 1 

LT 0.3 0.3 0.3 0.3 0.3 0.3 

XX 0.3 0.3 0.3 0.3 0.3 0.3 

DWINJH pump 

status 

DWINJH OK n/a n/a 0 0 0 0 

F 0 0 0 0 

H 0.3 0.3 0.3 0.3 

XX 0.3 0.3 0.3 0.3 

DWINJP pump 

status 

RDWINJP OK n/a n/a 1 1 1 1 

F 1 1 1 1 

H 0.147 0.147 0.147 0.147 

XX 0.147 0.147 0.147 0.147 

DWPCD vent 

status 

DWPCD WW n/a n/a n/a 0 0 0 

DW 0 0 0 1 1 1 

IS 0.1 0.1 0.1 0.01 0.01 0.1 

DWPCDnodc WW n/a n/a n/a 0 0 0 

DW 0 0 0 1 1 1 

IS 0.3 0.3 0.3 0.01 0.01 0.3 

MSLCR RPV 

pressure 

MSLCR HP 0.5 0.5 0.5 0.5 0.5 0.5 

MP 0 0 0 0 0 0 

LP 0 0 0 0 0 0 

PDWINJPCD RPV 

pressure 

PDWINJPCD HP n/a n/a 0.1 0.1 0.1 0.1 

MP 0.1 0.1 0.1 0.1 

LP 0 0 0 0 

PDWINJPCDnodc HP 0.3 0.3 0.3 0.3 

MP 0.3 0.3 0.3 0.3 

LP 0 0 0 0 

PPCD RPV 

pressure 

PPCD HP 8.21E02

8.21E028.21E028.21E028.21E028.21E02

MP 8.21E028.21E028.21E028.21E028.21E028.21E02

LP 0 0 0 0 0 0 

PRPVINJCR RPV 

pressure 

PRPVINJCR HP 0.1 0.1 n/a n/a n/a n/a 

MP 0.1 0.1 

LP 0 0 

PRPVINJCRnodc HP 0.3 0.3 

MP 0.3 0.3 

LP 0 0 

RPVINJCRH pump 

status 

RPVINJCRH OK 0 0 n/a n/a n/a n/a 

F 0 0 

H 0 0.3 

XX 0 0.3 

RPVINJCRP pump 

status 

RPVINJCRP OK 1 1 n/a n/a n/a n/a 

F 1 1 

H 1 0.147 

XX 1 0.147 

RPVINJLMTH pump 

status 

RPVINJLMTH OK 0 0 n/a n/a n/a n/a 

F 0 0 

H 0 0.3 

XX 0 0.3 

RPVINJLMTP pump 

status 

RPVINJLMTP OK 0 0 n/a n/a n/a n/a 

F 1 1 

H 1 0.147 

XX 1 0.147 

Table D-1 Accident progression event tree down-branch probabilities. 
D-3

WWPCD vent 

status

WWPCD WW 0 0 0 n/a n/a n/a

DW 1 1 1 0 0 0 

IS 0.1 0.1 0.1 0.1 0.1 0.1

WWPCDnodc WW 0 0 0 n/a n/a n/a

DW 1 1 1 0 0 0 

IS 0.3 0.3 0.3 0.3 0.3 0.3

Note: Down-branch probabilities based on the scoping HRA are highlighted.

APET 

Heading 

PDS 

Attribute 

APET DownBranch Identifier 

PDS 

Attribute 

Value 

APET1 APET2 APET3 APET4 APET5 APET6 

E-1

APPENDIX E

INTERNAL EVENT ELAP FREQUENCY SUPPORTING INFORMATION 

Table E-2 Emergency power system reliability parameters

LOOP Category

LOOP Frequency

(per year) 

Offsite Power Recovery Curve 

Lognormal Parameters 

μ σ

Plant Centered 2.07E-03 -0.76 1.287 

Switchyard Centered 1.04E-02 -0.391 1.256 

Grid Related 1.86E-02 0.3 1.064 

Weather Related 4.83E-03 0.793 1.982 

Source: U.S. Nuclear Regulatory Commission, “Reevaluation of Station Blackout Risk at 

Nuclear Power Plants,” NUREG/CR-6890, Vol. 1, ADAMS Accession No. ML060200477, 

November 2005. 

EPS Class 

EPS Probability of Failure to

Start (per demand)

EPS Rate of Failure to 

Run (per hour) 

2 3.1E-04 1.5E-04 

3 1.6E-04 7.2E-05 

4 2.5E-05 7.5E-06 

Source: Derived from information in the 2011 Update to NUREG/CR-5500, Vol. 5 

Table E-1 Loss of offsite power frequencies and recovery curve parameters
E-2

Table E-3 Internal event elap frequenciesby eps class and sbo coping time

EPS Class 2; SBO Coping Time = 4 hours 

EPS

Failure 

Mode Description 

Plant 

Centered 

Switchyard 

Centered 

Grid 

Related 

Weather 

Related Total 

EPS failure 

to start 

frequency (per year) 3.0E-08 2.5E-07 8.8E-07 5.7E-07 1.7E-06

contribution 0.4% 3.6% 12.7% 8.2% 25.1% 

EPS failure 

to run 

frequency (per year) 5.3E-08 4.7E-07 1.6E-06 3.0E-06 5.2E-06

contribution 0.8% 6.9% 23.1% 44.2% 74.9% 

Total frequency (per year) 8.3E-08 7.2E-07 2.5E-06 3.6E-06 6.9E-06

contribution 1.2% 10.5% 35.8% 52.5% 100.0% 

EPS Class 3; SBO Coping Time = 4 hours 

EPS

Failure 

Mode Description 

Plant 

Centered 

Switchyard 

Centered 

Grid 

Related 

Weather 

Related Total 

EPS failure 

to start 

frequency (per year) 1.5E-08 1.3E-07 4.5E-07 2.9E-07 8.8E-07

contribution 0.4% 3.7% 12.9% 8.4% 25.4% 

EPS failure 

to run 

frequency (per year) 2.6E-08 2.4E-07 7.9E-07 1.5E-06 2.6E-06

contribution 0.8% 6.8% 23.0% 44.0% 74.6% 

Total frequency (per year) 4.2E-08 3.6E-07 1.2E-06 1.8E-06 3.4E-06

contribution 1.2% 10.5% 35.9% 52.4% 100.0% 

EPS Class 4; SBO Coping Time = 4 hours 

EPS

Failure 

Mode Description 

Plant 

Centered 

Switchyard 

Centered 

Grid 

Related 

Weather 

Related Total 

EPS failure 

to start 

frequency (per year) 2.4E-09 2.0E-08 7.0E-08 4.5E-08 1.4E-07

contribution 0.6% 4.9% 17.3% 11.2% 34.0% 

EPS failure 

to run 

frequency (per year) 2.7E-09 2.5E-08 8.3E-08 1.6E-07 2.7E-07

contribution 0.7% 6.1% 20.3% 39.0% 66.0% 

Total frequency (per year) 5.2E-09 4.5E-08 1.5E-07 2.0E-07 4.1E-07

contribution 1.3% 11.0% 37.6% 50.1% 100.0% 

EPS Class 2; SBO Coping Time = 8 hours 

EPS

Failure 

Mode Description 

Plant 

Centered 

Switchyard 

Centered 

Grid 

Related 

Weather 

Related Total 

EPS failure 

to start 

frequency (per year) 8.7E-09 7.9E-08 2.7E-07 3.8E-07 7.4E-07

contribution 0.2% 1.9% 6.6% 9.4% 18.1% 

EPS failure 

to run 

frequency (per year) 2.2E-08 2.1E-07 6.7E-07 2.4E-06 3.3E-06

contribution 0.5% 5.2% 16.4% 59.7% 81.9% 

Total frequency (per year) 3.1E-08 2.9E-07 9.4E-07 2.8E-06 4.1E-06

contribution 0.8% 7.1% 23.0% 69.1% 100.0% 

EPS Class 3; SBO Coping Time = 8 hours 

EPS

Failure 

Mode Description 

Plant 

Centered 

Switchyard 

Centered 

Grid 

Related 

Weather 

Related Total 

EPS failure 

to start 

frequency (per year) 4.4E-09 4.0E-08 1.4E-07 1.9E-07 3.8E-07

contribution 0.2% 2.0% 6.7% 9.5% 18.4% 

EPS failure frequency (per year) 1.1E-08 1.1E-07 3.3E-07 1.2E-06 1.7E-06
E-3

to run contribution 0.5% 5.2% 16.3% 59.5% 81.6%

Total frequency (per year) 1.5E-08 1.5E-07 4.7E-07 1.4E-06 2.0E-06

contribution 0.8% 7.1% 23.0% 69.1% 100.0%

EPS

Failure 

Mode Description

Plant 

Centered

Switchyard 

Centered

Grid 

Related

Weather 

Related Total

EPS failure

to start

frequency (per year) 7.0E-10 6.3E-09 2.2E-08 3.1E-08 5.9E-08

contribution 0.3% 2.7% 9.3% 13.2% 25.4%

EPS failure

to run

frequency (per year) 1.2E-09 1.1E-08 3.5E-08 1.3E-07 1.7E-07

contribution 0.5% 4.7% 14.9% 54.4% 74.6%

Total frequency (per year) 1.9E-09 1.7E-08 5.6E-08 1.6E-07 2.3E-07

contribution 0.8% 7.4% 24.2% 67.6% 100.0%

EPS Class 4; SBO Coping Time = 8 hours 

F-1

APPENDIX F

SEISMIC ELAP FREQUENCY SUPPORTING INFORMATION

Table F-1 Seismic ELAP top logic

Top Event Description Boolean

Expression Basic Event Description 

SELAP ELAP due to seismic 

event 

= SLOOP 

* ( SP

+ DC)

seismic LOOP 

onsite power fails after seismic event 

DC power fails after seismic event 

SELAP1 

(applies to CDET 

Seqs. 1 to 224) 

ELAP due to seismic 

event and DC succeeds 

and RCIC-ST succeeds 

= SLOOP 

* SP

* /D

* /RCI -ST

seismic LOOP 

onsite power fails after seismic event 

complement of top event DC 

complement of top event RCIC-ST

SELAP2 

(Applies to 

CDET 

Seqs. 225, 226) 

ELAP due to seismic 

event and DC succeeds 

and RCIC-ST fails 

= SLOOP 

* SP

* /D

* R IC-ST

seismic LOOP 

onsite power fails after seismic event 

complement of top event DC 

RCIC fails after seismic event 

SELAP3 

(Applies to 

CDET 

Seqs. 227 to 

338)

ELAP due to seismic 

event and DC fails 

and RCIC-ST succeeds 

= SLOOP 

* D

* /RCI -ST

seismic LOOP 

DC power fails after seismic event 

complement of top event RCIC-ST

SELAP4 

(Applies to 

CDET 

Seqs. 339, 340) 

ELAP due to seismic 

event and DC fails 

and RCIC-ST fails 

= SLOOP 

* D

* R IC-ST

seismic LOOP 

DC power fails after seismic event 

RCIC fails after seismic event 

OSP Onsite power fails after

seismic event 

= ACSWGR-S 

+ EDG-S-CTRL

+ EDG-S-DT

+ EDG-S-EG

+ EDG-S-FOST

+ EDG-S-SAR

+ EPS-FTR

+ EPS-FTS

seismic failure of AC switchgear 

seismic failure of EDG controls 

seismic failure of EDG day tank 

seismic failure of EDG engine or 

generator 

seismic failure of EDG fuel oil storage 

tank 

seismic failure of EDG starting air receiver 

EPS fails to run (random) 

EPS fails to run (random) 

DC DC power fails after 

seismic event 

= BATTERY-S 

+ DCSWGR-S

seismic failure of batteries 

seismic failure of DC switchgear 

RCIC-ST RCIC fails after seismic 

event 

= RCIC-FTR1 

+ RCIC-FTS

+ RCIC-S

RCIC fails to run in Phase 1 

RCIC fails to start 

RCIC seismic failure 
F-2

Table F-2 Summary of seismic fragility information

Identifier Description 

Median Seismic 

Capacitya, C50

(g) 

Logarithmic Standard 

Deviations 

Randomnessb

βR

Uncertaintya

βU

ACSWGR-S seismic failure of AC switchgear 1.17 0.24 0.58 

BATTERY-S seismic failure of batteries 1.11 0.24 0.63 

DCSWGR-S seismic failure of DC switchgear 1.39 0.24 0.40 

EDG-S-CTRL seismic failure of EDG controls 1.00 0.24 0.45 

EDG-S-DT seismic failure of EDG day tank 1.33 0.24 0.98 

EDG-S-EG seismic failure of EDG engine or 

generator 1.16 0.24 0.54 

EDG-S-FOST seismic failure of EDG fuel oil storage 

tank 1.03 0.24 0.52 

EDG-S-SAR seismic failure of EDG starting air 

receiver 0.78 0.24 0.61 

LOOP-S seismic LOOP 0.29 0.24 0.45 

RCIC-S seismic failure of RCIC 1.33 0.24 0.59 

a

Inputs and results of the opinion pool used to estimate the median seismic capacity (C50) and logarithmic 

standard deviation due to uncertainty (βU) are provided in the seismic fragility worksheets. 

bThe logarithmic standard deviation due to randomness (βR) is generic value recommended in the SPID 
F-3

Table F-3 AC switchgear seismic fragility worksheet

Plant 

IPEEE Information Median Capacity 

Percentiles 

Opinion Pool 

C50 βU 5th 95th importance weight 

Beaver Valley 2 0.29 1.24 3.22 2 6.9% 

Catawba 1.32 0.55 0.53 3.26 2 6.9% 

Columbia 1.1 0.28 0.69 1.74 2 6.9% 

D.C. Cook 0.66 0.32 0.39 1.12 1 3.4% 

Hope Creek 1.5 0.32 0.89 2.54 1 3.4% 

Indian Point 2 1.5 0.32 0.89 2.54 1 3.4% 

Indian Point 3 0.67 0.321 0.40 1.14 2 6.9% 

Kewaunee 3.47 0.36 1.92 6.27 2 6.9% 

McGuire 0.95 0.34 0.54 1.66 2 6.9% 

Oconee 1.4 0.25 0.93 2.11 1 3.4% 

Oyster Creek 1 0.32 0.59 1.69 1 3.4% 

Palisades 1.08 0.24 0.73 1.60 1 3.4% 

Pilgrim 2 0.35 1.12 3.56 2 6.9% 

Point Beach 1.03 0.36 0.57 1.86 2 6.9% 

Salem 1.5 0.32 0.89 2.54 1 3.4% 

Seabrook 0.57 0.4 0.30 1.10 2 6.9% 

Surry 1.17 0.25 0.78 1.77 2 6.9% 

Three Mile Island 1 0.97 0.28 0.61 1.54 2 6.9% 

opinion pool 1.17 n/a 0.49 3.31 n/a n/a 

log-normal fit 1.17 0.58 0.45 3.04 n/a n/a 

Shaded cells: 

C50

βU

screening value as reported in the IPEEE 

generic value recommended in the SPID 
F-4

Table F-4 Battery seismic fragility worksheet

Plant 

IPEEE Information Median Capacity 

Percentiles 

Opinion Pool 

C50 βU 5th 95th importance weight 

Beaver Valley 0.52 0.35 0.29 0.92 2 7.1% 

Catawba 1.94 0.48 0.88 4.27 2 7.1% 

Columbia 1.5 0.32 0.89 2.54 1 3.6% 

D.C. Cook 0.71 0.21 0.50 1.00 2 7.1% 

Hope Creek 1.5 0.32 0.89 2.54 1 3.6% 

Indian Point 2 1.5 0.32 0.89 2.54 1 3.6% 

Indian Point 3 0.88 0.321 0.52 1.49 2 7.1% 

Kewaunee 2.74 0.36 1.52 4.95 2 7.1% 

McGuire 0.95 0.34 0.54 1.66 2 7.1% 

Oconee 1.39 0.53 0.58 3.32 2 7.1% 

Oyster Creek 1 0.32 0.59 1.69 1 3.6% 

Palisades 0.22 0.42 0.11 0.44 1 3.6% 

Pilgrim 1.2 0.32 0.71 2.03 1 3.6% 

Point Beach 1.2 0.36 0.66 2.17 2 7.1% 

Salem 1.06 0.35 0.60 1.89 2 7.1% 

Seabrook 2 0.32 1.18 3.39 1 3.6% 

Surry 1.02 0.42 0.51 2.04 2 7.1% 

Three Mile Island 1 1 0.32 0.59 1.69 1 3.6% 

opinion pool 1.11 n/a 0.39 3.05 n/a n/a 

log-normal fit 1.11 0.63 0.40 3.10 n/a n/a 

Shaded cells: 

C50

βU

screening value as reported in the IPEEE 

generic value recommended in the SPID 
F-5

Table F-5 DC switchgear seismic fragility worksheet

Plant 

IPEEE Information Median Capacity 

Percentiles 

Opinion Pool 

C50 βU 5th 95th importance weight 

Beaver Valley 1.65 0.35 0.93 2.93 2 9.5% 

Catawba 1.48 0.36 0.82 2.68 2 9.5% 

Columbia 1.5 0.32 0.89 2.54 1 4.8% 

Hope Creek 1.5 0.32 0.89 2.54 1 4.8% 

Indian Point 2 1.5 0.32 0.89 2.54 1 4.8% 

Indian Point 3 1.2 0.32 0.71 2.03 1 4.8% 

Kewaunee 1.1 0.36 0.61 1.99 2 9.5% 

Oconee 2 0.32 1.18 3.39 1 4.8% 

Oyster Creek 1 0.32 0.59 1.69 1 4.8% 

Palisades 1.08 0.24 0.73 1.60 1 4.8% 

Pilgrim 2.11 0.35 1.19 3.75 2 9.5% 

Point Beach 1.2 0.36 0.66 2.17 2 9.5% 

Salem 1.5 0.32 0.89 2.54 1 4.8% 

Seabrook 1.37 0.28 0.86 2.17 2 9.5% 

Three Mile Island 1 1 0.32 0.59 1.69 1 4.8% 

opinion pool 1.39 n/a 0.74 2.73 n/a n/a 

log-normal fit 1.39 0.40 0.72 2.67 n/a n/a 

Shaded cells: 

C50

βU

screening value as reported in the IPEEE 

generic value recommended in the SPID 
F-6

Table F-6 EDG controls seismic fragility worksheet

Plant 

IPEEE Information Median Capacity 

Percentiles 

Opinion Pool 

C50 βU 5th 95th importance weight 

Catawba 0.82 0.31 0.49 1.37 2 13.3% 

Columbia 0.95 0.28 0.60 1.51 2 13.3% 

Hope Creek 1.5 0.32 0.89 2.54 2 13.3% 

Indian Point 2 0.65 0.44 0.32 1.34 1 6.7% 

Indian Point 3 0.88 0.321 0.52 1.49 1 6.7% 

Kewaunee 1.57 0.32 0.93 2.66 1 6.7% 

McGuire 0.99 0.22 0.69 1.42 2 13.3% 

Palisades 0.66 0.32 0.39 1.12 2 13.3% 

Salem 1.5 0.32 0.89 2.54 2 13.3% 

opinion pool 1.00 n/a 0.49 2.12 n/a n/a 

log-normal fit 1.00 0.45 0.48 2.09 n/a n/a 

Shaded cells: 

C50

βU

screening value as reported in the IPEEE 

generic value recommended in the SPID 
F-7

Table F-7 EDG day tank seismic fragility worksheet

Plant 

IPEEE Information Median Capacity 

Percentiles 

Opinion Pool 

C50 βU 5th 95th importance weight 

D. C. Cook 0.27 0.32 0.16 0.46 2 11.8% 

Indian Point 3 0.75 0.213 0.53 1.06 2 11.8% 

Kewaunee 1.86 0.32 1.10 3.15 2 11.8% 

McGuire 1.44 0.28 0.91 2.28 1 5.9% 

Oconee 1.95 0.54 0.80 4.74 1 5.9% 

Palisades 0.22 0.32 0.13 0.37 1 5.9% 

Pilgrim 4.81 0.35 2.70 8.55 2 11.8% 

Salem 1.5 0.32 0.89 2.54 2 11.8% 

Seabrook 2 0.32 1.18 3.39 2 11.8% 

Surry 1.08 0.45 0.52 2.26 1 5.9% 

TMI-1 1.01 0.1 0.86 1.19 1 5.9% 

opinion pool 1.33 n/a 0.21 5.23 n/a n/a 

log-normal fit 1.33 0.98 0.27 6.66 n/a n/a 

Shaded cells: 

C50

βU

screening value as reported in the IPEEE 

generic value recommended in the SPID 
F-8

Table F-8 EDG engine and generator seismic fragility worksheet

Plant 

IPEEE Information Median Capacity 

Percentiles 

Opinion Pool 

C50 βU 5th 95th importance weight 

Beaver Valley 1.24 0.4 0.64 2.39 2 9.5% 

Catawba 2 0.32 1.18 3.39 1 4.8% 

D. C Cook 0.91 0.31 0.55 1.52 2 9.5% 

Indian Point 3 1.16 0.321 0.68 1.97 2 9.5% 

Kewaunee 3.67 0.32 2.17 6.21 1 4.8% 

Oconee 1.42 0.29 0.88 2.29 2 9.5% 

Oyster Creek 0.56 0.31 0.34 0.93 2 9.5% 

Pilgrim 1.39 0.57 0.54 3.55 2 9.5% 

Point Beach 1.2 0.32 0.71 2.03 1 4.8% 

Seabrook 1.51 0.35 0.85 2.69 2 9.5% 

Surry 1.07 0.3 0.65 1.75 2 9.5% 

TMI-1 1.02 0.14 0.81 1.28 2 9.5% 

opinion pool 1.16 n/a 0.52 3.06 n/a n/a 

log-normal fit 1.16 0.54 0.48 2.82 n/a n/a 

Shaded cells: 

C50

βU

screening value as reported in the IPEEE 

generic value recommended in the SPID 
F-9

Table F-9 EDG fuel oil storage tank seismic fragility worksheet

Plant 

IPEEE Information Median Capacity 

Percentiles 

Opinion Pool 

C50 βU 5th 95th importance weight 

Beaver Valley 1.31 0.4 0.68 2.53 2 18.2% 

Indian Point 3 0.75 0.213 0.53 1.06 2 18.2% 

Oconee 1.75 0.43 0.86 3.55 2 18.2% 

Palisades 0.65 0.32 0.38 1.10 1 9.1% 

Point Beach 0.8 0.36 0.44 1.45 2 18.2% 

Salem 1.5 0.5 0.66 3.41 2 18.2% 

opinion pool 1.03 n/a 0.51 2.79 n/a n/a 

log-normal fit 1.03 0.52 0.44 2.41 n/a n/a 

Shaded cells: 

C50

βU

screening value as reported in the IPEEE 

generic value recommended in the SPID 
F-10

Table F-10 EDG starting air receiver seismic fragility worksheet

Plant 

IPEEE Information Median Capacity 

Percentiles 

Opinion Pool 

C50 βU 5th 95th importance weight 

Hope Creek 1.5 0.32 0.89 2.54 1 11.1% 

Indian Point 3 0.75 0.213 0.53 1.06 2 22.2% 

McGuire 0.68 0.4 0.35 1.31 2 22.2% 

Salem 1.5 0.32 0.89 2.54 1 11.1% 

Seabrook 2 0.32 1.18 3.39 1 11.1% 

TMI-1 0.4 0.25 0.27 0.60 2 22.2% 

opinion pool 0.79 n/a 0.32 2.37 n/a n/a 

log-normal fit 0.79 0.61 0.29 2.13 n/a n/a 

Shaded cells: 

C50

βU

screening value as reported in the IPEEE 

generic value recommended in the SPID 
F-11

Table F-11 Loss of offsite power seismic fragility worksheet

Plant 

IPEEE Information Median Capacity 

Percentiles 

Opinion Pool 

C50 βU 5th 95th importance weight 

Beaver Valley 0.25 0.25 0.17 0.38 2 6.3% 

Catawba 0.3 0.5 0.13 0.68 2 6.3% 

Columbia 0.31 0.43 0.15 0.63 2 6.3% 

D.C. Cook 0.2 0.25 0.13 0.30 2 6.3% 

Hope Creek 0.31 0.43 0.15 0.63 2 6.3% 

Indian Point 2 0.3 0.5 0.13 0.68 2 6.3% 

Indian Point 3 0.3 0.3 0.18 0.49 2 6.3% 

Kewaunee 0.35 0.5 0.15 0.80 2 6.3% 

McGuire 0.3 0.5 0.13 0.68 2 6.3% 

Oconee 0.3 0.5 0.13 0.68 2 6.3% 

Oyster Creek 0.3 0.5 0.13 0.68 2 6.3% 

Pilgrim 0.35 0.55 0.14 0.86 2 6.3% 

Salem 0.31 0.43 0.15 0.63 2 6.3% 

Seabrook 0.3 0.5 0.13 0.68 2 6.3% 

Surry 0.3 0.4 0.16 0.58 2 6.3% 

Three Mile Island 1 0.3 0.25 0.20 0.45 2 6.3% 

opinion pool 0.29 n/a 0.14 0.64 n/a n/a 

log-normal fit 0.29 0.45 0.14 0.62 n/a n/a 

Shaded cells: 

C50

βU

screening value as reported in the IPEEE 

generic value recommended in the SPID 
F-12

Table F-12 RCIC seismic fragility worksheet

Plant 

IPEEE Information Median Capacity 

Percentiles 

Opinion Pool 

C50 βU 5th 95th importance weight 

Catawba 2.00 0.32 1.18 3.39 1 4.8% 

Columbia 1.05 0.28 0.66 1.66 2 9.5% 

D.C. Cook 2.28 0.27 1.46 3.55 2 9.5% 

Hope Creek 1.50 0.32 0.89 2.54 1 4.8% 

Indian Point 2 1.50 0.32 0.89 2.54 1 4.8% 

Indian Point 3 1.20 0.32 0.71 2.03 1 4.8% 

Kewaunee 1.57 0.36 0.87 2.84 2 9.5% 

McGuire 0.48 0.37 0.26 0.88 2 9.5% 

Oconee 2.00 0.32 1.18 3.39 1 4.8% 

Palisades 1.08 0.24 0.73 1.60 2 9.5% 

Pilgrim 2.50 0.35 1.41 4.45 2 9.5% 

Point Beach 1.20 0.32 0.71 2.03 1 4.8% 

Salem 1.50 0.32 0.89 2.54 1 4.8% 

Surry 0.68 0.3 0.42 1.11 2 9.5% 

opinion pool 1.33 n/a 0.45 3.11 n/a n/a 

log-normal fit 1.33 0.59 0.51 3.50 n/a n/a 

Shaded cells: 

C50

βU

screening value as reported in the IPEEE 

generic value recommended in the SPID 
F-13

Table F-13 Browns Ferry seismic ELAP frequencies 

Frequency 

(per year) 

Bin 1 

0.03 

to

0.05 

g 

Bin 2 

0.05 

to

0.075 

g 

Bin 3 

0.075 

to

0.1 g 

Bin 4 

0.1 

to

0.15 

g 

Bin 5 

0.15 

to

0.3 g 

Bin 6 

0.3 

to

0.5 g 

Bin 7 

0.5 

to 

0.75 

g 

Bin 8 

0.75 

to

1 g 

Bin 9 

1 to 

1.5 g 

Bin 

10

1.5 

to

3 g Total % 

Seismic Bin 

1.5E036.9E043.1E042.5E041.7E043.1E058.0E062.4E061.5E066.5E07 n/a n/a 

Total ELAP 1.2E10

2.4E091.5E081.6E073.3E068.6E066.0E062.3E061.5E066.5E072.2E05 100.0% 

ELAP and

DC success 

and

RCIC-ST 

success 

(Seqs. 1 to 

224)

1.2E102.3E091.5E081.5E073.0E066.9E063.8E069.1E072.2E074.6E091.5E05 66.8% 

ELAP and

DC success 

and

RCIC-ST 

fails 

(Seqs. 225, 

226)

1.1E122.1E111.4E101.4E093.3E082.5E075.2E073.2E071.8E071.5E081.3E06 5.9% 

ELAP and

DC fails 

and

RCIC-ST 

success 

(Seqs. 227 

to 338) 

1.9E146.8E122.1E106.1E093.1E071.3E061.4E068.1E075.9E071.4E074.6E06 20.6% 

ELAP and

DC fails 

and

RCIC-ST 

fails 

(Seqs. 339, 

340)

1.7E166.1E141.9E125.6E113.4E094.8E082.0E072.8E074.9E074.8E071.5E06 6.7% 
F-14

Table F-14 Brunswick seismic ELAP frequencies 

Frequency 

(per year) 

Bin 1 

0.03 

to

0.05 

g 

Bin 2 

0.05 

to

0.075 

g 

Bin 3 

0.075 

to

0.1 g 

Bin 4 

0.1 

to

0.15 

g 

Bin 5 

0.15 

to

0.3 g 

Bin 6 

0.3 

to

0.5 g 

Bin 7 

0.5 

to 

0.75 

g 

Bin 8 

0.75 

to

1 g 

Bin 9 

1 to 

1.5 g 

Bin 

10

1.5 

to

3 g Total % 

Seismic Bin 

7.6E043.8E041.7E041.4E048.5E051.5E053.5E069.4E075.2E072.0E07 n/a n/a 

Total ELAP 6.7E10

1.0E084.0E082.2E072.1E064.2E062.6E069.0E075.1E072.0E071.1E05 100.0% 

ELAP and

DC success 

and

RCIC-ST 

success 

(Seqs. 1 to 

224)

6.6E101.0E083.9E082.1E072.0E063.4E061.7E063.5E077.5E081.4E097.7E06 71.5% 

ELAP and

DC success 

and

RCIC-ST 

fails 

(Seqs. 225, 

226)

6.0E129.1E113.5E101.9E092.2E081.2E072.3E071.2E076.2E084.6E095.6E07 5.2% 

ELAP and

DC fails 

and

RCIC-ST 

success 

(Seqs. 227 

to 338) 

9.6E153.8E121.2E103.4E091.6E076.3E076.2E073.1E072.1E074.4E082.0E06 18.4% 

ELAP and

DC fails 

and

RCIC-ST 

fails 

(Seqs. 339, 

340)

8.7E173.4E141.1E123.1E111.8E092.3E088.5E081.1E071.7E071.5E075.3E07 4.9% 
F-15

Table F-15 Cooper seismic ELAP frequencies 

Frequency 

(per year) 

Bin 1 

0.03 

to

0.05 

g 

Bin 2 

0.05 

to

0.075 

g 

Bin 3 

0.075 

to

0.1 g 

Bin 4 

0.1 

to

0.15 

g 

Bin 5 

0.15 

to

0.3 g 

Bin 6 

0.3 

to

0.5 g 

Bin 7 

0.5 

to 

0.75 

g 

Bin 8 

0.75 

to

1 g 

Bin 9 

1 to 

1.5 g 

Bin 

10

1.5 

to

3 g Total % 

Seismic Bin 

4.8E041.6E045.7E054.3E052.9E056.5E061.8E065.1E072.7E079.6E08 n/a n/a 

Total ELAP 4.2E10

4.1E091.3E086.7E087.3E071.8E061.3E064.8E072.7E079.6E084.9E06 100.0% 

ELAP and

DC success 

and

RCIC-ST 

success 

(Seqs. 1 to 

224)

4.2E104.1E091.3E086.5E086.7E071.5E068.6E071.9E074.0E086.8E103.3E06 68.9% 

ELAP and

DC success 

and

RCIC-ST 

fails 

(Seqs. 225, 

226)

3.8E123.7E111.2E105.9E107.4E095.5E081.2E076.6E083.3E082.3E092.8E07 5.8% 

ELAP and

DC fails 

and

RCIC-ST 

success 

(Seqs. 227 

to 338) 

6.0E151.5E123.9E111.0E095.5E082.8E073.2E071.7E071.1E072.1E089.6E07 19.7% 

ELAP and

DC fails 

and

RCIC-ST 

fails 

(Seqs. 339, 

340)

5.5E171.4E143.5E139.6E126.1E101.0E084.4E085.8E089.1E087.2E082.8E07 5.7% 
F-16

Table F-16 Duane Arnold seismic ELAP frequencies 

Frequency 

(per year) 

Bin 1 

0.03 

to

0.05 

g 

Bin 2 

0.05 

to

0.075 

g 

Bin 3 

0.075 

to

0.1 g 

Bin 4 

0.1 

to

0.15 

g 

Bin 5 

0.15 

to

0.3 g 

Bin 6 

0.3 

to

0.5 g 

Bin 7 

0.5 

to 

0.75 

g 

Bin 8 

0.75 

to

1 g 

Bin 9 

1 to 

1.5 g 

Bin 

10

1.5 

to

3 g Total % 

Seismic Bin 

1.9E046.0E052.1E051.6E051.1E052.8E069.1E073.1E072.0E078.6E08 n/a n/a 

Total ELAP 1.7E10

1.6E094.8E092.5E082.8E077.9E076.9E072.9E071.9E078.6E082.4E06 100.0% 

ELAP and

DC success 

and

RCIC-ST 

success 

(Seqs. 1 to 

224)

1.7E101.6E094.8E092.4E082.5E076.4E074.4E071.2E072.8E086.1E101.5E06 64.1% 

ELAP and

DC success 

and

RCIC-ST 

fails 

(Seqs. 225, 

226)

1.5E121.4E114.3E112.2E102.8E092.3E086.0E084.0E082.4E082.0E091.5E07 6.4% 

ELAP and

DC fails 

and

RCIC-ST 

success 

(Seqs. 227 

to 338) 

2.4E155.9E131.4E113.9E102.1E081.2E071.6E071.0E077.8E081.9E085.0E07 21.4% 

ELAP and

DC fails 

and

RCIC-ST 

fails 

(Seqs. 339, 

340)

2.2E175.3E151.3E133.5E122.3E104.3E092.2E083.5E086.5E086.4E081.9E07 8.1% 
F-17

Table F-17 Fermi seismic ELAP frequencies 

Frequency 

(per year) 

Bin 1 

0.03 

to

0.05 

g 

Bin 2 

0.05 

to

0.075 

g 

Bin 3 

0.075 

to

0.1 g 

Bin 4 

0.1 

to

0.15 

g 

Bin 5 

0.15 

to

0.3 g 

Bin 6 

0.3 

to

0.5 g 

Bin 7 

0.5 

to 

0.75 

g 

Bin 8 

0.75 

to

1 g 

Bin 9 

1 to 

1.5 g 

Bin 

10

1.5 

to

3 g Total % 

Seismic Bin 

5.1E041.9E047.6E056.2E054.5E051.1E053.6E061.2E067.0E072.9E07 n/a n/a 

Total ELAP 4.1E11

6.5E103.8E094.0E089.0E073.1E062.7E061.1E067.0E072.9E078.8E06 100.0% 

ELAP and

DC success 

and

RCIC-ST 

success 

(Seqs. 1 to 

224)

4.0E116.4E103.7E093.8E088.0E072.5E061.7E064.3E071.0E072.1E095.6E06 63.6% 

ELAP and

DC success 

and

RCIC-ST 

fails 

(Seqs. 225, 

226)

3.6E135.8E123.4E113.4E108.8E099.1E082.3E071.5E078.5E087.0E095.7E07 6.5% 

ELAP and

DC fails 

and

RCIC-ST 

success 

(Seqs. 227 

to 338) 

6.5E151.9E125.2E111.5E098.5E084.8E076.4E073.8E072.8E076.6E081.9E06 22.0% 

ELAP and

DC fails 

and

RCIC-ST 

fails 

(Seqs. 339, 

340)

5.8E171.7E144.7E131.4E119.3E101.8E088.7E081.3E072.3E072.2E076.9E07 7.8% 
F-18

Table F-18 FitzPatrick seismic ELAP frequencies 

Frequency 

(per year) 

Bin 1 

0.03 

to

0.05 

g 

Bin 2 

0.05 

to

0.075 

g 

Bin 3 

0.075 

to

0.1 g 

Bin 4 

0.1 

to

0.15 

g 

Bin 5 

0.15 

to

0.3 g 

Bin 6 

0.3 

to

0.5 g 

Bin 7 

0.5 

to

0.75 

g 

Bin 8 

0.75 

to

1 g 

Bin 9 

1 to 

1.5 g 

Bin 

10

1.5 

to

3 g Total % 

Seismic Bin 4.9E04

1.6E045.8E054.1E052.3E054.6E061.3E064.1E072.5E071.1E07 n/a n/a 

Total ELAP 3.9E11

5.6E102.9E092.6E084.7E071.3E069.7E073.9E072.5E071.1E073.5E06 100.0% 

ELAP and

DC success 

and

RCIC-ST 

success 

(Seqs. 1 to 

224)

3.9E115.5E102.8E092.5E084.2E071.0E066.2E071.5E073.6E087.5E102.3E06 65.6% 

ELAP and

DC success 

and

RCIC-ST 

fails 

(Seqs. 225, 

226)

3.5E135.0E122.6E112.2E104.6E093.7E088.4E085.3E083.0E082.5E092.1E07 6.1% 

ELAP and

DC fails 

and

RCIC-ST 

success 

(Seqs. 227 

to 338)

6.2E151.6E123.9E119.8E104.4E082.0E072.3E071.4E071.0E072.4E087.3E07 21.2% 

ELAP and

DC fails 

and

RCIC-ST 

fails 

(Seqs. 339, 

340)

5.6E171.4E143.5E139.0E124.8E107.1E093.2E084.7E088.3E088.0E082.5E07 7.2% 
F-19

Table F-19 Hatch seismic ELAP frequencies 

Frequency 

(per year) 

Bin 1 

0.03 

to

0.05 

g 

Bin 2 

0.05 

to

0.075 

g 

Bin 3 

0.075 

to

0.1 g 

Bin 4 

0.1 

to

0.15 

g 

Bin 5 

0.15 

to

0.3 g 

Bin 6 

0.3 

to

0.5 g 

Bin 7 

0.5 

to

0.75 

g 

Bin 8 

0.75 

to

1 g 

Bin 9 

1 to 

1.5 g 

Bin 

10

1.5 

to

3 g Total % 

Seismic Bin 7.0E04

3.2E041.4E041.1E046.8E051.3E053.3E061.0E066.4E073.0E07 n/a n/a 

Total ELAP 3.2E10

4.5E091.8E081.2E071.5E063.6E062.5E069.6E076.4E073.0E079.6E06 100.0% 

ELAP and

DC success 

and

RCIC-ST 

success 

(Seqs. 1 to 

224)

3.1E104.5E091.8E081.1E071.4E062.9E061.6E063.8E079.3E082.1E096.5E06 67.3% 

ELAP and

DC success 

and

RCIC-ST 

fails 

(Seqs. 225, 

226)

2.8E124.1E111.6E101.0E091.5E081.0E072.1E071.3E077.7E087.1E095.5E07 5.8% 

ELAP and

DC fails 

and

RCIC-ST 

success 

(Seqs. 227 

to 338) 

8.8E153.1E129.3E112.6E091.3E075.4E075.9E073.4E072.5E076.7E081.9E06 20.1% 

ELAP and

DC fails 

and

RCIC-ST 

fails 

(Seqs. 339, 

340)

8.0E172.8E148.4E132.4E111.4E092.0E088.1E081.2E072.1E072.2E076.5E07 6.8% 
F-20

Table F-20 Hope Creek seismic ELAP frequencies 

Frequency 

(per year) 

Bin 1 

0.03 

to

0.05 

g 

Bin 2 

0.05 

to

0.075 

g 

Bin 3 

0.075 

to

0.1 g 

Bin 4 

0.1 

to

0.15 

g 

Bin 5 

0.15 

to

0.3 g 

Bin 6 

0.3 

to

0.5 g 

Bin 7 

0.5 

to 

0.75 

g 

Bin 8 

0.75 

to

1 g 

Bin 9 

1 to 

1.5 g 

Bin 

10

1.5 

to

3 g Total % 

Seismic Bin 

5.5E042.2E049.0E057.3E055.2E051.1E052.8E067.4E073.7E071.1E07 n/a n/a 

Total ELAP 2.5E10

3.1E091.2E087.8E081.2E063.1E062.1E067.1E073.7E071.1E077.6E06 100.0% 

ELAP and

DC success 

and

RCIC-ST 

success 

(Seqs. 1 to 

224)

2.5E103.1E091.2E087.5E081.0E062.5E061.3E062.8E075.3E088.0E105.3E06 69.8% 

ELAP and

DC success 

and

RCIC-ST 

fails 

(Seqs. 225, 

226)

2.2E122.8E111.1E106.9E101.2E089.1E081.8E079.7E084.4E082.7E094.3E07 5.6% 

ELAP and

DC fails 

and

RCIC-ST 

success 

(Seqs. 227 

to 338) 

6.9E152.1E126.1E111.8E099.7E084.7E075.1E072.5E071.5E072.5E081.5E06 19.6% 

ELAP and

DC fails 

and

RCIC-ST 

fails 

(Seqs. 339, 

340)

6.2E171.9E145.5E131.6E111.1E091.7E086.9E088.6E081.2E078.4E083.8E07 5.0% 
F-21

Table F-21 Monticello seismic ELAP frequencies 

Frequency 

(per year) 

Bin 1 

0.03 

to

0.05 

g 

Bin 2 

0.05 

to

0.075 

g 

Bin 3 

0.075 

to

0.1 g 

Bin 4 

0.1 

to

0.15 

g 

Bin 5 

0.15 

to

0.3 g 

Bin 6 

0.3 

to

0.5 g 

Bin 7 

0.5 

to 

0.75 

g 

Bin 8 

0.75 

to

1 g 

Bin 9 

1 to 

1.5 g 

Bin 

10

1.5 

to

3 g Total % 

Seismic Bin 2.8E04

1.1E044.7E054.0E053.0E058.0E062.6E068.8E075.6E072.3E07 n/a n/a 

Total ELAP 2.5E10

3.0E091.1E086.2E087.6E072.3E062.0E068.4E075.6E072.3E076.7E06 100.0% 

ELAP and

DC success 

and

RCIC-ST 

success 

(Seqs. 1 to 

224)

2.5E103.0E091.1E086.1E087.0E071.8E061.3E063.3E078.1E081.6E094.3E06 64.0% 

ELAP and

DC success 

and

RCIC-ST 

fails 

(Seqs. 225, 

226)

2.2E122.7E119.7E115.5E107.7E096.7E081.7E071.2E076.8E085.5E094.4E07 6.5% 

ELAP and

DC fails 

and

RCIC-ST 

success 

(Seqs. 227 

to 338) 

3.6E151.1E123.2E119.7E105.7E083.4E074.7E072.9E072.2E075.2E081.4E06 21.5% 

ELAP and

DC fails 

and

RCIC-ST 

fails 

(Seqs. 339, 

340)

3.2E171.0E142.9E138.9E126.3E101.2E086.5E081.0E071.9E071.7E075.4E07 8.0% 
F-22

Table F-22 Peach Bottom seismic ELAP frequencies 

Frequency 

(per year) 

Bin 1 

0.03 

to

0.05 

g 

Bin 2 

0.05 

to

0.075 

g 

Bin 3 

0.075 

to

0.1 g 

Bin 4 

0.1 

to

0.15 

g 

Bin 5 

0.15 

to

0.3 g 

Bin 6 

0.3 

to

0.5 g 

Bin 7 

0.5 

to 

0.75 

g 

Bin 8 

0.75 

to

1 g 

Bin 9 

1 to 

1.5 g 

Bin

10

1.5 

to

3 g Total % 

Seismic Bin 4.8E04

2.4E041.2E041.2E041.1E043.9E051.6E056.2E064.6E062.6E06 n/a n/a 

Total ELAP 2.2E10

3.5E091.7E081.3E072.6E061.1E051.2E055.9E064.6E062.6E063.8E05 100.0% 

ELAP and

DC success 

and

RCIC-ST 

success 

(Seqs. 1 to 

224)

2.2E103.4E091.6E081.2E072.3E068.8E067.4E062.3E066.7E071.8E082.2E05 56.8% 

ELAP and

DC success 

and

RCIC-ST 

fails 

(Seqs. 225, 

226)

1.9E123.1E111.5E101.1E092.6E083.2E071.0E068.0E075.5E076.1E082.8E06 7.3% 

ELAP and

DC fails 

and

RCIC-ST 

success 

(Seqs. 227 

to 338) 

6.1E152.4E128.4E112.9E092.2E071.7E062.8E062.1E061.8E065.7E079.1E06 23.9% 

ELAP and

DC fails 

and

RCIC-ST 

fails 

(Seqs. 339, 

340)

5.5E172.2E147.6E132.7E112.4E096.1E083.8E077.1E071.5E061.9E064.6E06 12.0% 
F-23

Table F-23 Pilgrim seismic ELAP frequencies 

Frequency 

(per year) 

Bin 1 

0.03 

to

0.05 

g 

Bin 2 

0.05 

to

0.075 

g 

Bin 3 

0.075 

to

0.1 g 

Bin 4 

0.1 

to

0.15 

g 

Bin 5 

0.15 

to

0.3 g 

Bin 6 

0.3 

to

0.5 g 

Bin 7 

0.5 

to 

0.75 

g 

Bin 8 

0.75 

to

1 g 

Bin 9 

1 to 

1.5 g 

Bin

10

1.5 

to

3 g Total % 

Seismic Bin 1.1E03

5.3E042.6E042.5E042.3E047.3E052.9E051.1E057.5E063.5E06 n/a n/a 

Total ELAP 9.5E10

1.4E085.9E083.9E075.7E062.1E052.1E051.0E057.5E063.5E067.0E05 100.0% 

ELAP and

DC success 

and

RCIC-ST 

success 

(Seqs. 1 to 

224)

9.4E101.4E085.9E083.8E075.2E061.7E051.4E054.1E061.1E062.5E084.1E05 59.5% 

ELAP and

DC success 

and

RCIC-ST 

fails 

(Seqs. 225, 

226)

8.5E121.3E105.3E103.4E095.7E086.2E071.9E061.4E069.1E078.3E084.9E06 7.1% 

ELAP and

DC fails 

and

RCIC-ST 

success 

(Seqs. 227 

to 338) 

1.4E145.2E121.8E106.0E094.3E073.1E065.1E063.6E063.0E067.8E071.6E05 23.1% 

ELAP and

DC fails 

and

RCIC-ST 

fails 

(Seqs. 339, 

340)

1.2E164.7E141.6E125.5E114.7E091.1E077.0E071.2E062.5E062.6E067.2E06 10.3% 
F-24

Table F-24 Quad Cities seismic ELAP frequencies 

Frequency 

(per year) 

Bin 1 

0.03 

to

0.05 

g 

Bin 2 

0.05 

to

0.075 

g 

Bin 3 

0.075 

to

0.1 g 

Bin 4 

0.1 

to

0.15 

g 

Bin 5 

0.15 

to

0.3 g 

Bin 6 

0.3 

to

0.5 g 

Bin 7 

0.5 

to 

0.75 

g 

Bin 8 

0.75 

to

1 g 

Bin 9 

1 to 

1.5 g 

Bin 

10

1.5 

to

3 g Total % 

Seismic Bin 4.7E04

1.6E046.2E054.8E053.4E058.7E062.8E069.4E076.0E072.6E07 n/a n/a 

Total ELAP 3.7E11

5.6E103.1E093.1E086.8E072.4E062.1E069.0E076.0E072.6E077.0E06 100.0% 

ELAP and

DC success 

and

RCIC-ST 

success 

(Seqs. 1 to 

224)

3.7E115.5E103.1E092.9E086.1E071.9E061.3E063.6E078.7E081.9E094.4E06 62.6% 

ELAP and

DC success 

and

RCIC-ST 

fails 

(Seqs. 225, 

226)

3.3E135.0E122.8E112.7E106.7E097.0E081.8E071.2E077.3E086.2E094.6E07 6.6% 

ELAP and

DC fails 

and

RCIC-ST 

success 

(Seqs. 227 

to 338) 

5.9E151.6E124.2E111.2E096.4E083.7E075.1E073.1E072.4E075.9E081.6E06 22.4% 

ELAP and

DC fails 

and

RCIC-ST 

fails 

(Seqs. 339, 

340)

5.4E171.5E143.8E131.1E117.0E101.3E086.9E081.1E072.0E072.0E075.9E07 8.4% 
G-1

Table G-1 Plant damage state frequencies (per year) 

APPENDIX G

PLANT DAMAGE STATE FREQUENCIES 

1 E-HP-IS-ST-XX 4.3E-07 4.3E-07 4.3E-07

2 E-HP-IS-XX-XX 1.1E-07 1.1E-07 1.1E-07

3 E-LP-IS-ST-XX 1.0E-06 1.0E-06 1.0E-06

4 E-LP-IS-XX-XX 9.8E-07 9.8E-07 9.8E-07

5 L-HP-DW-LT-F 2.0E-09 3.2E-08 6.8E-09

6 L-HP-DW-LT-H 1.3E-09 2.1E-08 4.5E-09

7 L-HP-DW-LT-OK 8.0E-10 1.3E-08 2.7E-09

8 L-HP-DW-OK-F 6.6E-10 7.2E-08 6.6E-09

9 L-HP-DW-OK-H 4.3E-10 4.8E-08 4.3E-09

10 L-HP-DW-OK-OK 8.6E-11 9.4E-09 8.6E-10

11 L-HP-DW-ST-F 4.2E-09 6.5E-08 1.4E-08

12 L-HP-DW-ST-H 2.7E-09 4.3E-08 9.1E-09

13 L-HP-DW-ST-OK 1.6E-09 2.6E-08 5.4E-09

14 L-HP-IS-LT-F 2.1E-08 0.0E+00 2.1E-08

15 L-HP-IS-LT-H 1.4E-08 0.0E+00 1.4E-08

16 L-HP-IS-LT-OK 8.1E-09 0.0E+00 8.1E-09

17 L-HP-IS-OK-F 6.5E-08 0.0E+00 6.5E-08

18 L-HP-IS-OK-H 4.3E-08 0.0E+00 4.3E-08

19 L-HP-IS-OK-OK 8.5E-09 0.0E+00 8.5E-09

20 L-HP-IS-ST-F 4.2E-08 0.0E+00 4.2E-08

21 L-HP-IS-ST-H 2.8E-08 0.0E+00 2.8E-08

22 L-HP-IS-ST-OK 1.6E-08 0.0E+00 1.6E-08

23 L-HP-WW-LT-F 6.8E-09 2.3E-10 2.0E-09

24 L-HP-WW-LT-H 4.5E-09 1.5E-10 1.3E-09

25 L-HP-WW-LT-OK 2.7E-09 8.9E-11 8.0E-10

26 L-HP-WW-OK-F 6.6E-09 6.6E-10 6.6E-10 

27 L-HP-WW-OK-H 4.3E-09 4.3E-10 4.3E-10

28 L-HP-WW-OK-OK 8.6E-10 8.6E-11 8.6E-11

29 L-HP-WW-ST-F 1.4E-08 4.6E-10 4.2E-09

30 L-HP-WW-ST-H 9.1E-09 3.0E-10 2.7E-09

31 L-HP-WW-ST-OK 5.4E-09 1.8E-10 1.6E-09

32 L-MP-DW-LT-F 1.6E-08 2.6E-07 5.5E-08

33 L-MP-DW-LT-H 1.1E-08 1.7E-07 3.6E-08

34 L-MP-DW-LT-OK 6.5E-09 1.0E-07 2.2E-08

35 L-MP-DW-OK-F 6.8E-09 7.5E-07 6.8E-08

36 L-MP-DW-OK-H 4.5E-09 4.9E-07 4.5E-08

37 L-MP-DW-OK-OK 8.9E-10 9.8E-08 8.9E-09

38 L-MP-DW-ST-F 6.8E-09 1.1E-07 2.3E-08

Index Plant Damage State WWF Venting 

Strategy 

DWF Venting 

Strategy (Passive) 

DWF Venting 

Strategy (Manual) 
G-2

39 L-MP-DW-ST-H 4.5E-09 7.0E-08 1.5E-08

40 L-MP-DW-ST-OK 2.6E-09 4.2E-08 8.8E-09

41 L-MP-IS-LT-F 1.7E-07 0.0E+00 1.7E-07

42 L-MP-IS-LT-H 1.1E-07 0.0E+00 1.1E-07

43 L-MP-IS-LT-OK 6.5E-08 0.0E+00 6.5E-08

44 L-MP-IS-OK-F 6.7E-07 0.0E+00 6.7E-07

45 L-MP-IS-OK-H 4.4E-07 0.0E+00 4.4E-07

46 L-MP-IS-OK-OK 8.8E-08 0.0E+00 8.8E-08

47 L-MP-IS-ST-F 6.8E-08 0.0E+00 6.8E-08

48 L-MP-IS-ST-H 4.5E-08 0.0E+00 4.5E-08

49 L-MP-IS-ST-OK 2.7E-08 0.0E+00 2.7E-08

50 L-MP-WW-LT-F 5.5E-08 1.8E-09 1.6E-08

51 L-MP-WW-LT-H 3.6E-08 1.2E-09 1.1E-08

52 L-MP-WW-LT-OK 2.2E-08 7.2E-10 6.5E-09

53 L-MP-WW-OK-F 6.8E-08 6.8E-09 6.8E-09

54 L-MP-WW-OK-H 4.5E-08 4.5E-09 4.5E-09

55 L-MP-WW-OK-OK 8.9E-09 8.9E-10 8.9E-10

56 L-MP-WW-ST-F 2.3E-08 7.5E-10 6.8E-09

57 L-MP-WW-ST-H 1.5E-08 4.9E-10 4.5E-09

58 L-MP-WW-ST-OK 8.8E-09 2.9E-10 2.6E-09

59 M-HP-DW-LT-F 7.4E-10 1.2E-08 2.5E-09

60 M-HP-DW-LT-H 2.1E-09 3.3E-08 7.0E-09

61 M-HP-DW-LT-OK 1.2E-09 2.0E-08 4.2E-09

62 M-HP-DW-OK-F 3.6E-13 4.0E-11 3.6E-12

63 M-HP-DW-OK-H 1.0E-12 1.1E-10 1.0E-11

64 M-HP-DW-OK-OK 2.0E-13 2.2E-11 2.0E-12

65 M-HP-DW-ST-F 1.5E-09 2.4E-08 5.0E-09

66 M-HP-DW-ST-H 4.3E-09 6.7E-08 1.4E-08

67 M-HP-DW-ST-OK 2.5E-09 4.0E-08 8.5E-09

68 M-HP-IS-LT-F 7.5E-09 0.0E+00 7.5E-09

69 M-HP-IS-LT-H 2.1E-08 0.0E+00 2.1E-08

70 M-HP-IS-LT-OK 1.3E-08 0.0E+00 1.3E-08

71 M-HP-IS-LT-XX 3.3E-08 1.1E-09 3.3E-08

72 M-HP-IS-OK-F 3.6E-11 0.0E+00 3.6E-11

73 M-HP-IS-OK-H 1.0E-10 0.0E+00 1.0E-10

74 M-HP-IS-OK-OK 2.0E-11 0.0E+00 2.0E-11

75 M-HP-IS-OK-XX 5.8E-09 5.8E-10 5.8E-09

76 M-HP-IS-ST-F 1.5E-08 0.0E+00 1.5E-08

77 M-HP-IS-ST-H 4.3E-08 0.0E+00 4.3E-08

78 M-HP-IS-ST-OK 2.6E-08 0.0E+00 2.6E-08

79 M-HP-IS-ST-XX 6.8E-08 2.3E-09 6.8E-08

80 M-HP-WW-LT-F 2.5E-09 8.2E-11 7.4E-10

81 M-HP-WW-LT-H 7.0E-09 2.3E-10 2.1E-09

82 M-HP-WW-LT-OK 4.2E-09 1.4E-10 1.2E-09

83 M-HP-WW-OK-F 3.6E-12 3.6E-13 3.6E-13

Index Plant Damage State WWF Venting 

Strategy 

DWF Venting 

Strategy (Passive) 

DWF Venting 

Strategy (Manual) 
G-3

84 M-HP-WW-OK-H 1.0E-11 1.0E-12 1.0E-12

85 M-HP-WW-OK-OK 2.0E-12 2.0E-13 2.0E-13

86 M-HP-WW-ST-F 5.0E-09 1.7E-10 1.5E-09

87 M-HP-WW-ST-H 1.4E-08 4.7E-10 4.3E-09

88 M-HP-WW-ST-OK 8.5E-09 2.8E-10 2.5E-09

89 M-LP-DW-LT-F 1.1E-08 1.8E-07 3.8E-08

90 M-LP-DW-LT-H 3.3E-08 5.1E-07 1.1E-07

91 M-LP-DW-OK-F 2.0E-09 2.2E-07 2.0E-08

92 M-LP-DW-OK-H 5.8E-09 6.3E-07 5.8E-08

93 M-LP-DW-ST-F 1.0E-08 1.6E-07 3.4E-08

94 M-LP-DW-ST-H 2.9E-08 4.6E-07 9.7E-08

95 M-LP-IS-LT-F 1.2E-07 0.0E+00 1.2E-07

96 M-LP-IS-LT-H 3.3E-07 0.0E+00 3.3E-07

97 M-LP-IS-LT-XX 6.9E-08 2.3E-09 6.9E-08

98 M-LP-IS-OK-F 2.0E-07 0.0E+00 2.0E-07

99 M-LP-IS-OK-H 5.7E-07 0.0E+00 5.7E-07

100 M-LP-IS-OK-XX 1.2E-08 1.2E-09 1.2E-08

101 M-LP-IS-ST-F 1.0E-07 0.0E+00 1.0E-07

102 M-LP-IS-ST-H 2.9E-07 0.0E+00 2.9E-07

103 M-LP-IS-ST-XX 1.0E-07 3.3E-09 1.0E-07

104 M-LP-WW-LT-F 3.8E-08 1.3E-09 1.1E-08

105 M-LP-WW-LT-H 1.1E-07 3.6E-09 3.3E-08

106 M-LP-WW-OK-F 2.0E-08 2.0E-09 2.0E-09

107 M-LP-WW-OK-H 5.8E-08 5.8E-09 5.8E-09

108 M-LP-WW-ST-F 3.4E-08 1.1E-09 1.0E-08

109 M-LP-WW-ST-H 9.7E-08 3.2E-09 2.9E-08

110 M-MP-DW-LT-F 8.1E-09 1.3E-07 2.7E-08

111 M-MP-DW-LT-H 2.3E-08 3.6E-07 7.7E-08

112 M-MP-DW-LT-OK 1.4E-08 2.2E-07 4.6E-08

113 M-MP-DW-OK-F 4.0E-12 4.4E-10 4.0E-11

114 M-MP-DW-OK-H 1.1E-11 1.2E-09 1.1E-10

115 M-MP-DW-OK-OK 2.2E-12 2.5E-10 2.2E-11

116 M-MP-DW-ST-F 3.3E-09 5.2E-08 1.1E-08

117 M-MP-DW-ST-H 9.5E-09 1.5E-07 3.2E-08

118 M-MP-DW-ST-OK 5.7E-09 8.9E-08 1.9E-08

119 M-MP-IS-LT-F 8.2E-08 0.0E+00 8.2E-08

120 M-MP-IS-LT-H 2.3E-07 0.0E+00 2.3E-07

121 M-MP-IS-LT-OK 1.4E-07 0.0E+00 1.4E-07

122 M-MP-IS-LT-XX 3.7E-07 1.2E-08 3.7E-07

123 M-MP-IS-OK-F 3.9E-10 0.0E+00 3.9E-10

124 M-MP-IS-OK-H 1.1E-09 0.0E+00 1.1E-09

125 M-MP-IS-OK-OK 2.2E-10 0.0E+00 2.2E-10

126 M-MP-IS-OK-XX 6.4E-08 6.4E-09 6.4E-08

127 M-MP-IS-ST-F 3.4E-08 0.0E+00 3.4E-08

128 M-MP-IS-ST-H 9.6E-08 0.0E+00 9.6E-08

Index Plant Damage State WWF Venting 

Strategy 

DWF Venting 

Strategy (Passive) 

DWF Venting 

Strategy (Manual) 
G-4

129 M-MP-IS-ST-OK 5.7E-08 0.0E+00 5.7E-08

130 M-MP-IS-ST-XX 1.5E-07 5.0E-09 1.5E-07

131 M-MP-WW-LT-F 2.7E-08 9.0E-10 8.1E-09

132 M-MP-WW-LT-H 7.7E-08 2.6E-09 2.3E-08

133 M-MP-WW-LT-OK 4.6E-08 1.5E-09 1.4E-08

134 M-MP-WW-OK-F 4.0E-11 4.0E-12 4.0E-12

135 M-MP-WW-OK-H 1.1E-10 1.1E-11 1.1E-11

136 M-MP-WW-OK-OK 2.2E-11 2.2E-12 2.2E-12

137 M-MP-WW-ST-F 1.1E-08 3.7E-10 3.3E-09

138 M-MP-WW-ST-H 3.2E-08 1.1E-09 9.5E-09

139 M-MP-WW-ST-OK 1.9E-08 6.3E-10 5.7E-09

8.9E-06 8.4E-06 8.9E-06

Index Plant Damage State WWF Venting 

Strategy 

DWF Venting 

Strategy (Passive) 

DWF Venting 

Strategy (Manual) 
H-1

APPENDIX H

SOURCE TERM INFORMATION TO MAP EACH MELCOR CASE 

VARIATION TO A MACCS SOURCE TERM BIN 

Table H-1 Mark I source term detailed information

Case 

Cesium 

Release 

(%) 

Iodine

Release 

(%) 

Fraction of 

Cesium Release

Through Vented

Pathway 

Fraction of 

Iodine

Release 

Through 

Vented

Pathway 

Start of 

Release to 

Environme

nt (hours) 

Source

Term Bin 

1 1.93% 22.70% 

78.2% 85.5% 14.9

12

1DF10 0.57% 5.24% 7 

1DF100 0.44% 3.49% 7 

1DF1000 0.42% 3.32% 7 

1S1 0.65% 6.33%

55.4% 54.0% 23.3

7 

1S1DF10 0.33% 3.25% 7 

1S1DF100 0.29% 2.94% 6 

1S1DF1000 0.29% 2.91% 6 

2 3.06% 26.40% 

53.9% 83.0% 14.9

15

2DF10 1.58% 6.69% 10

2DF100 1.43% 4.72% 10

2DF1000 1.41% 4.52% 10

3 9.88% 30.20% 

11.5% 21.6% 9.8

17

3DF10 8.85% 24.32% 17

3DF100 8.75% 23.74% 17

3DF1000 8.74% 23.68% 17

4 1.86% 9.85%

27.3% 59.2% 18.1

10

4DF10 1.40% 4.60% 10

4DF100 1.36% 4.08% 10

4DF1000 1.35% 4.03% 10

5 1.39% 6.43%

27.6% 61.7% 24.2

10

5DF10 1.04% 2.86% 9 

5DF100 1.01% 2.50% 9 

5DF1000 1.01% 2.46% 9 

6 1.97% 20.30% 

68.5% 80.8% 13.9

12

6DF10 0.76% 5.54% 7 

6DF100 0.63% 4.06% 7 

6DF1000 0.62% 3.92% 7 

7 1.51% 19.40% 

98.7% 99.0% 14.9

11

7DF10 0.17% 2.12% 6 

7DF100 0.03% 0.39% 5 

7DF1000 0.02% 0.22% 4 

7dw 1.62% 17.70% 

100.0% 100.0% 14.9

11

7dwDF10 0.16% 1.77% 6 

7dwDF100 0.02% 0.18% 4 

7dwDF1000 0.002% 0.02% 2 
H-2

Case 

Cesium 

Release 

(%) 

Iodine 

Release 

(%) 

Fraction of 

Cesium Release 

Through Vented 

Pathway 

Fraction of 

Iodine Release 

Through 

Vented 

Pathway 

Start of 

Release to 

Environme

nt (hours) 

Source 

Term Bin 

8 1.49% 19.20% 

100.0% 100.0% 14.9

11

8DF10 0.15% 1.92% 6 

8DF100 0.01% 0.19% 4 

8DF1000 0.001% 0.02% 2 

9 0.61% 7.84%

100.0% 100.0% 14.4

7 

9DF10 0.06% 0.78% 5 

9DF100 0.01% 0.08% 3 

9DF1000 0.001% 0.01% 1 

10 0.72% 8.04%

100.0% 100.0% 16.3

7 

10DF10 0.07% 0.80% 5 

10DF100 0.01% 0.08% 3 

10DF1000 0.001% 0.01% 1 

11 0.61% 7.79%

99.7% 100.0% 14.4

7 

11DF10 0.06% 0.78% 5 

11DF100 0.01% 0.08% 3 

11DF1000 0.003% 0.01% 2 

12 1.34% 16.40% 

100.0% 100.0% 13.9

11

12DF10 0.13% 1.64% 6 

12DF100 0.01% 0.16% 4 

12DF1000 0.001% 0.02% 2 

13 0.35% 3.90%

100.0% 100.0% 24.2

7 

13DF10 0.04% 0.39% 5 

13DF100 0.004% 0.04% 3 

13DF1000 0.0004% 0.004% 1 

14 0.62% 6.03%

97.1% 95.9% 14.9

7 

14DF10 0.08% 0.83% 5 

14DF100 0.02% 0.31% 4 

14DF1000 0.019% 0.26% 4 

15 0.61% 5.83%

100.0% 100.0% 14.9

7 

15DF10 0.06% 0.58% 5 

15DF100 0.01% 0.06% 3 

15DF1000 0.001% 0.01% 1 

16 0.61% 5.88%

99.7% 99.9% 14.9

7 

16DF10 0.06% 0.60% 5 

16DF100 0.01% 0.07% 3 

16DF1000 0.003% 0.01% 2 
H-3

Case 

Cesium

Release 

(%) 

Iodine

Release 

(%) 

Fraction of 

Cesium

Release 

Through 

Vented 

Pathway 

Fraction of 

Iodine

Release 

Through 

Vented 

Pathway 

Start of 

Release to 

Environme

nt (hours) 

Source 

Term Bin 

18 0.41% 4.97%

98.8% 100.0% 16.3

7 

18DF10 0.05% 0.50% 5 

18DF100 0.01% 0.05% 3 

18DF1000 0.005% 0.005% 3 

21 1.58% 19.90% 

97.5% 97.5% 14.9

11

21DF10 0.19% 2.44% 6 

21DF100 0.06% 0.69% 5 

21DF1000 0.04% 0.52% 5 

22 1.78% 22.30% 

100.0% 100.0% 14.9

12

22DF10 0.18% 2.23% 6 

22DF100 0.02% 0.22% 4 

22DF1000 0.002% 0.02% 2 

22dw 2.82% 18.60% 

100.0% 100.0% 14.9

14

22dwDF10 0.28% 1.86% 6 

22dwDF100 0.03% 0.19% 4 

22dwDF1000 0.003% 0.02% 2 

24 0.84% 8.24%

99.8% 99.8% 16.3

7 

24DF10 0.09% 0.84% 5 

24DF100 0.01% 0.10% 4 

24DF1000 0.003% 0.03% 2 

24dw 2.20% 8.59%

100.0% 100.0% 16

13

24dwDF10 0.22% 0.86% 6 

24dwDF100 0.02% 0.09% 4 

24dwDF1000 0.002% 0.01% 2 

27 0.63% 6.03%

96.7% 95.9% 14.9

7 

27DF10 0.08% 0.83% 5 

27DF100 0.03% 0.31% 4 

27DF1000 0.02% 0.26% 4 

28 0.61% 5.78%

100.0% 100.0% 14.9

7 

28DF10 0.06% 0.58% 5 

28DF100 0.01% 0.06% 3 

28DF1000 0.001% 0.01% 1 

30 0.62% 5.88%

99.7% 100.0% 14.9

7 

30DF10 0.06% 0.59% 5 

30DF100 0.01% 0.06% 3 

30DF1000 0.002% 0.01% 2 
H-4

Case 

Cesium 

Release 

(%) 

Iodine 

Release 

(%) 

Fraction of 

Cesium 

Release 

Through 

Vented 

Pathway 

Fraction of 

Iodine 

Release 

Through 

Vented 

Pathway 

Start of 

Release to 

Environme

nt (hours) 

Source 

Term Bin 

32 0.39% 4.85%

100.0% 100.0% 16.3

7 

32DF10 0.04% 0.49% 5 

32DF100 0.004% 0.05% 3 

32DF1000 0.0004% 0.01% 1 

41 4.58% 14.10% 

25.1% 46.0% 9.8

16

41DF10 3.55% 8.27% 13

41DF100 3.44% 7.68% 13

41DF1000 3.43% 7.63% 13

42 4.33% 10.70% 

26.6% 60.6% 9.8

16

42DF10 3.30% 4.87% 13

42DF100 3.19% 4.28% 13

42DF1000 3.18% 4.23% 13

43 4.77% 16.00% 

24.9% 40.5% 9.8

16

43DF10 3.70% 10.17% 13

43DF100 3.59% 9.58% 13

43DF1000 3.58% 9.53% 13

44 4.40% 10.80% 

27.0% 60.0% 9.8

16

44DF10 3.33% 4.97% 13

44DF100 3.22% 4.38% 13

44DF1000 3.21% 4.33% 13

45 0.90% 9.60%

100.0% 100.0% 14.8

7 

45DF10 0.09% 0.96% 5 

45DF100 0.01% 0.10% 3 

45DF1000 0.001% 0.01% 1 

46 0.98% 11.00% 

100.0% 100.0% 14.8

8 

46DF10 0.10% 1.10% 5 

46DF100 0.01% 0.11% 3 

46DF1000 0.001% 0.01% 2 

47 0.19% 1.24%

100.0% 100.0% 11.4

6 

47DF10 0.02% 0.12% 4 

47DF100 0.002% 0.01% 2 

47DF1000 0.0002% 0.001% 1 

48 0.24% 1.69%

100.0% 100.0% 11.4

6 

48DF10 0.02% 0.17% 4 

48DF100 0.002% 0.02% 2 

48DF1000 0.0002% 0.002% 1 
H-5

Case 

Cesium 

Release 

(%) 

Iodine 

Release 

(%) 

Fraction of 

Cesium 

Release 

Through 

Vented 

Pathway 

Fraction of 

Iodine 

Release 

Through 

Vented 

Pathway 

Start of 

Release to 

Environme

nt (hours) 

Source 

Term Bin 

49 0.53% 1.67%

99.6% 100.0% 7.3

7 

49DF10 0.06% 0.17% 5 

49DF100 0.01% 0.02% 3 

49DF1000 0.002% 0.002% 2 

50 0.21% 1.09%

99.8% 100.0% 7.3

6 

50DF10 0.02% 0.11% 4 

50DF100 0.003% 0.01% 2 

50DF1000 0.001% 0.001% 1 

51 0.93% 10.20% 

100.0% 100.0% 14.9

8 

51DF10 0.09% 1.02% 5 

51DF100 0.009% 0.10% 3 

51DF1000 0.0009% 0.01% 1 

52 15.90% 34.30% 

83.6% 77.0% 18.6

18

52DF10 3.93% 10.54% 13

52DF100 2.73% 8.16% 13

52DF1000 2.61% 7.93% 13

53 2.79% 29.10% 

100.0% 100.0% 18.6

15

53DF10 0.28% 2.91% 6 

53DF100 0.03% 0.29% 4 

53DF1000 0.003% 0.029% 2 
H-6

Table H-2 Mark II source term detailed information

Case

Cesium 

Release to 

Environment 

(%)

Iodine 

Release to 

Environment 

(%)

Start of 

Release to 

Environment 

(hours)

Fraction of Cesium 

and Iodine Release 

Through Vented 

Pathway

Source Term 

Bin

1 2.46% 19.81%

22.8 100%

8

1DF10 0.246% 1.981% 5

1DF100 0.025% 0.198% 4

1DF1000 0.002% 0.020% 3

3 1.09% 10.26%

14.3 100%

7

3DF10 0.109% 1.026% 5

3DF100 0.011% 0.103% 4

3DF1000 0.001% 0.010% 3

5 0.55% 4.94%

32.2 100%

6

5DF10 0.055% 0.494% 4

5DF100 0.006% 0.049% 3

5DF1000 0.001% 0.005% 2

10 0.23% 2.67%

22.2 100%

5

10DF10 0.023% 0.267% 4

10DF100 0.002% 0.027% 3

10DF1000 0.0002% 0.003% 2

11 0.04% 0.45%

20.3 100%

4

11DF10 0.004% 0.045% 3

11DF100 0.0004% 0.005% 2

11DF1000 0.00004% 0.0005% 1

24 0.71% 5.79%

30.5 100%

6

24DF10 0.071% 0.579% 4

24DF100 0.007% 0.058% 3

24DF1000 0.001% 0.006% 2

42 0.43% 3.65%

14.3 100%

6

42DF10 0.043% 0.365% 4

42DF100 0.004% 0.037% 3

42DF1000 0.0004% 0.004% 2

44 0.44% 3.68%

14.3 100%

6

44DF10 0.044% 0.368% 4

44DF100 0.004% 0.037% 3

44DF1000 0.0004% 0.004% 2
H-7

Case

Cesium 

Release to 

Environment 

(%)

Iodine 

Release to 

Environment 

(%)

Start of 

Release to 

Environment 

(hours)

Fraction of Cesium 

and Iodine Release 

Through Vented 

Pathway

Source Term 

Bin

45 2.29% 19.26%

18.3 100%

8

45DF10 0.229% 1.926% 5

45DF100 0.023% 0.193% 4

45DF1000 0.002% 0.019% 3

49 0.65% 7.62%

11.2 100%

6

49DF10 0.065% 0.762% 4

49DF100 0.006% 0.076% 3

49DF1000 0.001% 0.008% 2

51 2.29% 20.14%

16.6 100%

8

51DF10 0.229% 2.014% 5

51DF100 0.023% 0.201% 4

51DF1000 0.002% 0.020% 3

52 3.57% 28.67%

16.6 100%

9

52DF10 0.357% 2.867% 5

52DF100 0.036% 0.287% 4

52DF1000 0.004% 0.029% 3

I-1

APPENDIX I

MACCS SENSITIVITY CALCULATION RESULTS

Table I-1 Sensitivity calculation results for Mark I source terms with low 

 population site file (Hatch) 

Individual 

Early Fatality 

Risk

0-1.3 mi and 

beyond 0-10 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi

1 Base Case 6.63E-06 9.74E-07 3.33E-07 5.83E+03 1.32E+04 2.11E+07 2.16E+07 6.38E+00 7.22E+00 2.73E-01 2.73E-012 1-hour Evacuation Delay 6.63E-06 9.74E-07 3.33E-07 5.83E+03 1.32E+04 2.11E+07 2.16E+07 6.38E+00 7.22E+00 2.73E-01 2.73E-013 No Intermediate Phase 7.53E-06 1.06E-06 3.63E-07 5.91E+03 1.33E+04 2.05E+07 2.10E+07 6.42E+00 7.26E+00 3.08E+00 3.08E+004 1-year Intermediate Phase 5.82E-06 8.63E-07 2.95E-07 5.73E+03 1.30E+04 2.06E+07 2.11E+07 6.38E+00 7.22E+00 1.11E-03 1.11E-035 EPA Habitability Criterion (2 rem per year) 7.03E-06 9.84E-07 3.36E-07 5.84E+03 1.32E+04 2.04E+07 2.09E+07 6.38E+00 7.22E+00 0.00E+00 0.00E+006 No Intermediate Phase and EPA Hab. 7.84E-06 1.07E-06 3.64E-07 5.92E+03 1.33E+04 2.04E+07 2.09E+07 6.38E+00 7.22E+00 1.26E-01 1.26E-017 1-year Intermediate Phase and EPA Hab. 5.93E-06 8.65E-07 2.95E-07 5.74E+03 1.30E+04 2.04E+07 2.09E+07 6.38E+00 7.22E+00 1.11E-03 1.11E-038 Larger Nonevacuating Cohort 6.75E-06 9.76E-07 3.34E-07 5.83E+03 1.32E+04 2.07E+07 2.12E+07 6.38E+00 7.22E+00 2.73E-01 2.73E-01

9 Base Case 3.98E-04 1.17E-04 4.44E-05 2.33E+05 5.74E+05 1.12E+09 1.46E+09 3.73E+02 6.63E+02 7.19E+03 7.21E+03

10 1-hour Evacuation Delay 3.98E-04 1.17E-04 4.44E-05 2.33E+05 5.74E+05 1.12E+09 1.46E+09 3.73E+02 6.63E+02 7.19E+03 7.21E+0311 No Intermediate Phase 3.46E-04 1.16E-04 4.65E-05 2.33E+05 5.85E+05 7.11E+08 8.81E+08 5.39E+02 8.39E+02 1.76E+04 1.85E+0412 1-year Intermediate Phase 4.25E-04 1.09E-04 4.04E-05 2.25E+05 5.56E+05 1.54E+09 1.73E+09 3.21E+02 6.11E+02 3.43E+03 3.43E+0313 EPA Habitability Criterion (2 rem per year) 9.11E-04 1.55E-04 5.30E-05 2.67E+05 6.11E+05 4.27E+08 5.79E+08 2.96E+02 5.85E+02 1.51E+03 1.51E+0314 No Intermediate Phase and EPA Hab. 8.91E-04 1.67E-04 5.76E-05 2.78E+05 6.31E+05 2.98E+08 4.45E+08 3.26E+02 6.16E+02 3.70E+03 3.70E+0315 1-year Intermediate Phase and EPA Hab. 5.19E-04 1.24E-04 4.36E-05 2.38E+05 5.70E+05 6.62E+08 8.09E+08 3.21E+02 6.11E+02 3.43E+03 3.43E+0316 Larger Nonevacuating Cohort 4.02E-04 1.17E-04 4.44E-05 2.33E+05 5.74E+05 1.12E+09 1.46E+09 3.73E+02 6.63E+02 7.19E+03 7.21E+0317 Base Case 7.12E-04 3.32E-04 1.69E-04 5.65E+05 1.53E+06 4.22E+09 7.28E+09 1.22E+03 2.52E+03 4.19E+04 5.25E+04

18 1-hour Evacuation Delay 7.12E-04 3.32E-04 1.69E-04 5.65E+05 1.53E+06 4.22E+09 7.28E+09 1.22E+03 2.52E+03 4.19E+04 5.25E+0419 No Intermediate Phase 6.66E-04 3.07E-04 1.67E-04 5.55E+05 1.54E+06 2.80E+09 4.54E+09 1.67E+03 3.26E+03 7.22E+04 1.12E+0520 1-year Intermediate Phase 7.40E-04 3.26E-04 1.57E-04 5.52E+05 1.47E+06 6.79E+09 1.01E+10 9.72E+02 2.20E+03 2.51E+04 2.79E+0421 EPA Habitability Criterion (2 rem per year) 1.72E-03 5.41E-04 2.27E-04 7.50E+05 1.78E+06 1.76E+09 2.72E+09 8.12E+02 2.02E+03 1.33E+04 1.40E+0422 No Intermediate Phase and EPA Hab. 1.76E-03 5.59E-04 2.43E-04 7.75E+05 1.85E+06 1.14E+09 1.87E+09 9.40E+02 2.17E+03 2.31E+04 2.65E+0423 1-year Intermediate Phase and EPA Hab. 8.26E-04 3.81E-04 1.76E-04 6.01E+05 1.55E+06 3.71E+09 4.75E+09 9.72E+02 2.20E+03 2.51E+04 2.79E+0424 Larger Nonevacuating Cohort 7.17E-04 3.32E-04 1.69E-04 5.65E+05 1.53E+06 4.22E+09 7.28E+09 1.22E+03 2.52E+03 4.19E+04 5.25E+04

Mark I - Peach Bottom

0.00E+000.00E+000.00E+00

Run

Base Model

Site File Source Term Individual Latent Cancer Fatality Risk Population Dose (person-rem) Population Subject to Long-Term Protective Actions Offsite Cost ($ 2013) Type Land (sq mi) Exceeding Long-Term Habitability Criterion

Low - Hatch

Mk I - Low

 (Bin 3)

Mk I - Med

(Bin 10)

Mk I - High

(Bin 17)
I-2

Table I-2 Sensitivity calculation results for Mark I source terms with medium population site fil

e

 (Vermont Yankee)

Individual 

Early Fatality 

Risk

0-1.3 mi and 

beyond 0-10 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi

25 Base Case 1.01E-05 9.53E-07 2.99E-07 5.35E+03 1.56E+04 5.89E+07 5.94E+07 2.50E+00 3.14E+00 1.69E+00 1.69E+0026 1-hour Evacuation Delay 1.01E-05 9.53E-07 2.99E-07 5.35E+03 1.56E+04 5.89E+07 5.94E+07 2.50E+00 3.14E+00 1.69E+00 1.69E+0027 No Intermediate Phase 1.13E-05 1.04E-06 3.25E-07 5.61E+03 1.62E+04 5.56E+07 5.62E+07 2.56E+00 3.20E+00 2.76E+01 2.76E+0128 1-year Intermediate Phase 8.86E-06 8.42E-07 2.64E-07 5.02E+03 1.48E+04 5.65E+07 5.71E+07 2.49E+00 3.14E+00 2.35E-01 2.35E-0129 EPA Habitability Criterion (2 rem per year) 1.08E-05 9.70E-07 3.01E-07 5.40E+03 1.57E+04 5.50E+07 5.55E+07 2.49E+00 3.14E+00 1.26E-02 1.26E-0230 No Intermediate Phase and EPA Hab. 1.21E-05 1.06E-06 3.27E-07 5.67E+03 1.63E+04 5.48E+07 5.53E+07 2.50E+00 3.14E+00 2.70E-01 2.70E-0131 1-year Intermediate Phase and EPA Hab. 9.16E-06 8.49E-07 2.65E-07 5.04E+03 1.48E+04 5.48E+07 5.54E+07 2.49E+00 3.14E+00 2.35E-01 2.35E-0132 Larger Nonevacuating Cohort 1.03E-05 9.57E-07 2.99E-07 5.37E+03 1.56E+04 5.76E+07 5.81E+07 2.50E+00 3.14E+00 1.69E+00 1.69E+00

33 Base Case 5.00E-04 1.15E-04 4.29E-05 4.38E+05 1.36E+06 3.45E+09 5.25E+09 2.49E+02 4.80E+02 2.09E+04 2.11E+04

34 1-hour Evacuation Delay 5.00E-04 1.15E-04 4.29E-05 4.38E+05 1.36E+06 3.45E+09 5.25E+09 2.49E+02 4.80E+02 2.09E+04 2.11E+0435 No Intermediate Phase 4.41E-04 1.15E-04 4.63E-05 4.47E+05 1.45E+06 2.09E+09 2.50E+09 4.41E+02 6.83E+02 5.39E+04 6.15E+0436 1-year Intermediate Phase 5.30E-04 1.06E-04 3.83E-05 4.09E+05 1.25E+06 4.80E+09 5.45E+09 1.83E+02 4.13E+02 1.20E+04 1.20E+0437 EPA Habitability Criterion (2 rem per year) 1.19E-03 1.48E-04 4.81E-05 5.42E+05 1.48E+06 1.06E+09 1.29E+09 1.47E+02 3.77E+02 6.71E+03 6.71E+0338 No Intermediate Phase and EPA Hab. 1.16E-03 1.61E-04 5.27E-05 5.84E+05 1.60E+06 6.03E+08 7.82E+08 1.87E+02 4.17E+02 1.24E+04 1.24E+0439 1-year Intermediate Phase and EPA Hab. 6.31E-04 1.18E-04 4.02E-05 4.47E+05 1.29E+06 2.19E+09 2.37E+09 1.83E+02 4.13E+02 1.20E+04 1.20E+0440 Larger Nonevacuating Cohort 5.05E-04 1.15E-04 4.29E-05 4.39E+05 1.36E+06 3.45E+09 5.25E+09 2.49E+02 4.80E+02 2.09E+04 2.11E+04

41 Base Case 8.79E-04 3.49E-04 1.82E-04 1.28E+06 5.10E+06 1.51E+10 3.60E+10 9.96E+02 2.08E+03 1.30E+05 2.19E+05

42 1-hour Evacuation Delay 8.79E-04 3.49E-04 1.82E-04 1.28E+06 5.10E+06 1.51E+10 3.60E+10 9.96E+02 2.08E+03 1.30E+05 2.19E+0543 No Intermediate Phase 8.47E-04 3.23E-04 1.83E-04 1.24E+06 5.16E+06 9.89E+09 2.10E+10 1.51E+03 2.92E+03 2.44E+05 5.75E+0544 1-year Intermediate Phase 8.98E-04 3.37E-04 1.66E-04 1.23E+06 4.71E+06 2.39E+10 4.79E+10 7.08E+02 1.70E+03 7.66E+04 1.01E+0545 EPA Habitability Criterion (2 rem per year) 2.12E-03 5.44E-04 2.28E-04 1.87E+06 6.18E+06 5.16E+09 8.74E+09 5.07E+02 1.48E+03 3.99E+04 4.50E+0446 No Intermediate Phase and EPA Hab. 2.23E-03 5.67E-04 2.48E-04 1.97E+06 6.68E+06 3.03E+09 4.72E+09 6.68E+02 1.67E+03 7.14E+04 9.89E+0447 1-year Intermediate Phase and EPA Hab. 9.72E-04 3.92E-04 1.84E-04 1.40E+06 5.14E+06 1.31E+10 1.80E+10 7.08E+02 1.70E+03 7.66E+04 1.01E+0548 Larger Nonevacuating Cohort 8.88E-04 3.49E-04 1.82E-04 1.28E+06 5.10E+06 1.51E+10 3.60E+10 9.96E+02 2.08E+03 1.30E+05 2.19E+05

Mark I - Peach Bottom

0.00E+000.00E+000.00E+00

Run

Base Model

Site File Source Term Individual Latent Cancer Fatality Risk Population Dose (person-rem) Population Subject to Long-Term Protective Actions Offsite Cost ($ 2013) Type Land (sq mi) Exceeding Long-Term Habitability Criterion

Mk I - High

(Bin 17)

Medium - Vermont Yankee

Mk I - Low

 (Bin 3)

Mk I - Med

(Bin 10)
I-3

Table I-3 Sensitivity calculation results for Mark I source terms with high population site file (Peach Bottom) 

Individual 

Early Fatality 

Risk

0-1.3 mi and 

beyond 0-10 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi

49 Base Case 6.25E-06 7.16E-07 3.21E-07 1.65E+04 2.73E+04 9.81E+07 9.87E+07 9.80E+00 1.05E+01 5.65E-01 5.65E-0150 1-hour Evacuation Delay 6.25E-06 7.16E-07 3.21E-07 1.65E+04 2.73E+04 9.81E+07 9.87E+07 9.80E+00 1.05E+01 5.65E-01 5.65E-0151 No Intermediate Phase 7.12E-06 7.78E-07 3.49E-07 1.72E+04 2.84E+04 9.52E+07 9.58E+07 9.83E+00 1.05E+01 1.14E+01 1.14E+0152 1-year Intermediate Phase 5.47E-06 6.35E-07 2.84E-07 1.56E+04 2.59E+04 9.55E+07 9.61E+07 9.80E+00 1.05E+01 9.12E-02 9.12E-0253 EPA Habitability Criterion (2 rem per year) 6.60E-06 7.20E-07 3.22E-07 1.65E+04 2.73E+04 9.48E+07 9.54E+07 9.80E+00 1.05E+01 1.46E-02 1.46E-0254 No Intermediate Phase and EPA Hab. 7.37E-06 7.80E-07 3.49E-07 1.72E+04 2.84E+04 9.48E+07 9.54E+07 9.80E+00 1.05E+01 1.31E-01 1.31E-0155 1-year Intermediate Phase and EPA Hab. 5.56E-06 6.36E-07 2.85E-07 1.56E+04 2.59E+04 9.48E+07 9.54E+07 9.80E+00 1.05E+01 9.12E-02 9.12E-0256 Larger Nonevacuating Cohort 6.36E-06 7.17E-07 3.22E-07 1.65E+04 2.73E+04 9.64E+07 9.70E+07 9.80E+00 1.05E+01 5.65E-01 5.65E-01

57 Base Case 4.06E-04 9.78E-05 4.51E-05 1.36E+06 2.30E+06 9.90E+09 1.20E+10 4.80E+02 7.14E+02 5.14E+04 5.15E+04

58 1-hour Evacuation Delay 4.06E-04 9.78E-05 4.51E-05 1.36E+06 2.30E+06 9.90E+09 1.20E+10 4.80E+02 7.14E+02 5.14E+04 5.15E+0459 No Intermediate Phase 3.49E-04 9.80E-05 4.71E-05 1.37E+06 2.38E+06 6.16E+09 6.72E+09 6.20E+02 8.65E+02 1.60E+05 1.72E+0560 1-year Intermediate Phase 4.34E-04 8.97E-05 4.07E-05 1.26E+06 2.12E+06 1.30E+10 1.39E+10 4.36E+02 6.71E+02 2.02E+04 2.02E+0461 EPA Habitability Criterion (2 rem per year) 9.32E-04 1.21E-04 5.24E-05 1.62E+06 2.58E+06 2.64E+09 2.95E+09 4.16E+02 6.50E+02 7.01E+03 7.01E+0362 No Intermediate Phase and EPA Hab. 9.06E-04 1.32E-04 5.74E-05 1.75E+06 2.78E+06 1.60E+09 1.80E+09 4.41E+02 6.76E+02 2.39E+04 2.39E+0463 1-year Intermediate Phase and EPA Hab. 5.35E-04 1.01E-04 4.42E-05 1.39E+06 2.25E+06 4.41E+09 4.62E+09 4.36E+02 6.71E+02 2.02E+04 2.02E+0464 Larger Nonevacuating Cohort 4.11E-04 9.79E-05 4.51E-05 1.36E+06 2.30E+06 9.90E+09 1.20E+10 4.80E+02 7.14E+02 5.14E+04 5.15E+04

65 Base Case 7.10E-04 2.95E-04 1.68E-04 3.83E+06 7.72E+06 4.69E+10 6.79E+10 1.36E+03 2.47E+03 4.17E+05 5.04E+05

66 1-hour Evacuation Delay 7.10E-04 2.95E-04 1.68E-04 3.83E+06 7.72E+06 4.69E+10 6.79E+10 1.36E+03 2.47E+03 4.17E+05 5.04E+0567 No Intermediate Phase 6.66E-04 2.72E-04 1.64E-04 3.66E+06 7.66E+06 3.04E+10 4.19E+10 1.73E+03 3.13E+03 7.71E+05 1.10E+0668 1-year Intermediate Phase 7.40E-04 2.87E-04 1.57E-04 3.72E+06 7.26E+06 7.49E+10 9.88E+10 1.16E+03 2.19E+03 2.29E+05 2.55E+0569 EPA Habitability Criterion (2 rem per year) 1.74E-03 4.62E-04 2.29E-04 5.68E+06 1.01E+07 1.57E+10 1.93E+10 1.04E+03 2.04E+03 1.10E+05 1.17E+0570 No Intermediate Phase and EPA Hab. 1.77E-03 4.84E-04 2.45E-04 5.97E+06 1.07E+07 9.12E+09 1.10E+10 1.14E+03 2.17E+03 2.07E+05 2.37E+0571 1-year Intermediate Phase and EPA Hab. 8.27E-04 3.38E-04 1.78E-04 4.27E+06 8.06E+06 3.95E+10 4.49E+10 1.16E+03 2.19E+03 2.29E+05 2.55E+0572 Larger Nonevacuating Cohort 7.15E-04 2.95E-04 1.68E-04 3.83E+06 7.72E+06 4.69E+10 6.79E+10 1.36E+03 2.47E+03 4.17E+05 5.04E+05

Mark I - Peach Bottom

0.00E+000.00E+000.00E+00

Run

Base Model

Site File Source Term Individual Latent Cancer Fatality Risk Population Dose (person-rem) Population Subject to Long-Term Protective Actions Offsite Cost ($ 2013) Type Land (sq mi) Exceeding Long-Term Habitability Criterion

High - Peach Bottom

Mk I - Low

 (Bin 3)

Mk I - Med

(Bin 10)

Mk I - High

(Bin 17)
I-4

Table I-4 Sensitivity calculation results for Mark II source terms with low population site file (Columbia) 

Individual 

Early Fatality 

Risk

0-1.3 mi and 

beyond 0-10 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi

73 Base Case 7.05E-07 1.64E-07 9.15E-08 1.86E+03 2.42E+03 1.86E+07 1.86E+07 4.10E-04 4.10E-04 0.00E+00 0.00E+0074 1-hour Evacuation Delay 7.05E-07 1.64E-07 9.15E-08 1.86E+03 2.42E+03 1.86E+07 1.86E+07 4.10E-04 4.10E-04 0.00E+00 0.00E+0075 No Intermediate Phase 7.55E-07 1.72E-07 9.58E-08 1.87E+03 2.43E+03 1.86E+07 1.86E+07 4.10E-04 4.10E-04 0.00E+00 0.00E+0076 1-year Intermediate Phase 6.34E-07 1.53E-07 8.54E-08 1.85E+03 2.40E+03 1.86E+07 1.86E+07 4.10E-04 4.10E-04 0.00E+00 0.00E+0077 EPA Habitability Criterion (2 rem per year) 7.05E-07 1.64E-07 9.15E-08 1.86E+03 2.42E+03 1.86E+07 1.86E+07 4.10E-04 4.10E-04 0.00E+00 0.00E+0078 No Intermediate Phase and EPA Hab. 7.55E-07 1.72E-07 9.58E-08 1.87E+03 2.43E+03 1.86E+07 1.86E+07 4.10E-04 4.10E-04 0.00E+00 0.00E+0079 1-year Intermediate Phase and EPA Hab. 6.34E-07 1.53E-07 8.54E-08 1.85E+03 2.40E+03 1.86E+07 1.86E+07 4.10E-04 4.10E-04 0.00E+00 0.00E+0080 Larger Nonevacuating Cohort 7.32E-07 1.64E-07 9.16E-08 1.86E+03 2.42E+03 1.84E+07 1.84E+07 4.10E-04 4.10E-04 0.00E+00 0.00E+00

81 Base Case 1.15E-04 3.29E-05 1.86E-05 1.06E+05 1.84E+05 3.63E+08 4.12E+08 2.52E+02 3.64E+02 5.31E+02 5.31E+02

82 1-hour Evacuation Delay 1.15E-04 3.29E-05 1.86E-05 1.06E+05 1.84E+05 3.63E+08 4.12E+08 2.52E+02 3.64E+02 5.31E+02 5.31E+0283 No Intermediate Phase 9.98E-05 3.38E-05 1.93E-05 1.06E+05 1.86E+05 2.54E+08 3.02E+08 2.58E+02 3.70E+02 2.92E+03 2.92E+0384 1-year Intermediate Phase 1.10E-04 3.04E-05 1.70E-05 1.03E+05 1.81E+05 3.63E+08 4.11E+08 2.51E+02 3.62E+02 1.54E+02 1.54E+0285 EPA Habitability Criterion (2 rem per year) 1.72E-04 3.76E-05 2.08E-05 1.10E+05 1.88E+05 1.88E+08 2.36E+08 2.50E+02 3.62E+02 2.62E+01 2.62E+0186 No Intermediate Phase and EPA Hab. 1.89E-04 4.16E-05 2.30E-05 1.13E+05 1.93E+05 1.71E+08 2.19E+08 2.51E+02 3.62E+02 9.77E+01 9.77E+0187 1-year Intermediate Phase and EPA Hab. 1.39E-04 3.30E-05 1.82E-05 1.05E+05 1.83E+05 1.86E+08 2.34E+08 2.51E+02 3.62E+02 1.54E+02 1.54E+0288 Larger Nonevacuating Cohort 1.18E-04 3.29E-05 1.86E-05 1.06E+05 1.84E+05 3.63E+08 4.12E+08 2.52E+02 3.64E+02 5.31E+02 5.31E+02

89 Base Case 4.40E-04 3.06E-04 1.84E-04 5.65E+05 8.87E+05 4.63E+09 6.12E+09 1.23E+03 2.25E+03 4.03E+04 4.34E+04

90 1-hour Evacuation Delay 4.40E-04 3.06E-04 1.84E-04 5.65E+05 8.87E+05 4.63E+09 6.12E+09 1.23E+03 2.25E+03 4.03E+04 4.34E+0491 No Intermediate Phase 3.65E-04 2.77E-04 1.72E-04 5.87E+05 9.16E+05 3.82E+09 5.09E+09 1.59E+03 3.26E+03 8.94E+04 1.17E+0592 1-year Intermediate Phase 5.20E-04 3.01E-04 1.80E-04 5.54E+05 8.72E+05 6.96E+09 9.09E+09 1.07E+03 2.03E+03 8.20E+03 8.22E+0393 EPA Habitability Criterion (2 rem per year) 9.87E-04 4.13E-04 2.42E-04 6.65E+05 1.00E+06 2.47E+09 3.25E+09 1.10E+03 2.07E+03 1.30E+04 1.31E+0494 No Intermediate Phase and EPA Hab. 8.80E-04 4.18E-04 2.54E-04 7.11E+05 1.07E+06 2.05E+09 2.65E+09 1.29E+03 2.38E+03 4.22E+04 4.86E+0495 1-year Intermediate Phase and EPA Hab. 6.17E-04 3.55E-04 2.16E-04 6.03E+05 9.40E+05 4.00E+09 4.81E+09 1.07E+03 2.03E+03 8.20E+03 8.22E+0396 Larger Nonevacuating Cohort 4.96E-04 3.06E-04 1.84E-04 5.65E+05 8.87E+05 4.63E+09 6.12E+09 1.23E+03 2.25E+03 4.03E+04 4.34E+04

Mark II - Limerick

0.00E+000.00E+000.00E+00

Run

Base Model

Site File Source Term Individual Latent Cancer Fatality Risk Population Dose (person-rem) Population Subject to Long-Term Protective Actions Offsite Cost ($ 2013) Type Land (sq mi) Exceeding Long-Term Habitability Criterion

Low - Columbia

Mk II - Low

 (Bin 2)

Mk II - Med

(Bin 5)

Mk II - High

(Bin 8)
I-5

Table I-5 Sensitivity calculation results for Mark II source terms with medium population site file (Susquehanna) 

Individual 

Early Fatality 

Risk

0-1.3 mi and 

beyond 0-10 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi

97 Base Case 8.47E-07 1.52E-07 4.46E-08 1.29E+03 2.43E+03 9.12E+07 9.12E+07 1.61E-03 1.61E-03 0.00E+00 0.00E+0098 1-hour Evacuation Delay 8.47E-07 1.52E-07 4.46E-08 1.29E+03 2.43E+03 9.12E+07 9.12E+07 1.61E-03 1.61E-03 0.00E+00 0.00E+0099 No Intermediate Phase 9.10E-07 1.60E-07 4.68E-08 1.33E+03 2.50E+03 9.12E+07 9.12E+07 1.64E-03 1.64E-03 0.00E+00 0.00E+00

100 1-year Intermediate Phase 7.63E-07 1.41E-07 4.15E-08 1.24E+03 2.32E+03 9.12E+07 9.12E+07 1.61E-03 1.61E-03 0.00E+00 0.00E+00101 EPA Habitability Criterion (2 rem per year) 8.49E-07 1.52E-07 4.46E-08 1.29E+03 2.43E+03 9.12E+07 9.12E+07 1.61E-03 1.61E-03 0.00E+00 0.00E+00102 No Intermediate Phase and EPA Hab. 9.10E-07 1.60E-07 4.68E-08 1.33E+03 2.50E+03 9.12E+07 9.12E+07 1.61E-03 1.61E-03 0.00E+00 0.00E+00103 1-year Intermediate Phase and EPA Hab. 7.63E-07 1.41E-07 4.15E-08 1.24E+03 2.32E+03 9.12E+07 9.12E+07 1.61E-03 1.61E-03 0.00E+00 0.00E+00104 Larger Nonevacuating Cohort 8.79E-07 1.53E-07 4.47E-08 1.30E+03 2.43E+03 8.83E+07 8.83E+07 1.61E-03 1.61E-03 0.00E+00 0.00E+00

105 Base Case 1.08E-04 2.82E-05 9.18E-06 1.46E+05 3.60E+05 8.43E+08 9.58E+08 1.00E+02 2.03E+02 3.37E+03 3.37E+03

106 1-hour Evacuation Delay 1.08E-04 2.82E-05 9.18E-06 1.46E+05 3.60E+05 8.43E+08 9.58E+08 1.00E+02 2.03E+02 3.37E+03 3.37E+03107 No Intermediate Phase 9.03E-05 2.86E-05 9.75E-06 1.48E+05 3.75E+05 4.93E+08 5.58E+08 1.24E+02 2.26E+02 1.12E+04 1.12E+04108 1-year Intermediate Phase 1.06E-04 2.61E-05 8.23E-06 1.38E+05 3.37E+05 9.91E+08 1.06E+09 9.47E+01 1.97E+02 1.48E+03 1.48E+03109 EPA Habitability Criterion (2 rem per year) 1.81E-04 3.38E-05 1.00E-05 1.66E+05 3.81E+05 2.86E+08 3.51E+08 9.16E+01 1.94E+02 3.71E+02 3.71E+02110 No Intermediate Phase and EPA Hab. 1.86E-04 3.66E-05 1.09E-05 1.76E+05 4.02E+05 2.06E+08 2.71E+08 9.48E+01 1.97E+02 1.42E+03 1.42E+03111 1-year Intermediate Phase and EPA Hab. 1.35E-04 2.86E-05 8.60E-06 1.47E+05 3.45E+05 3.77E+08 4.42E+08 9.47E+01 1.97E+02 1.48E+03 1.48E+03112 Larger Nonevacuating Cohort 1.11E-04 2.83E-05 9.20E-06 1.46E+05 3.61E+05 8.40E+08 9.55E+08 1.00E+02 2.03E+02 3.37E+03 3.37E+03

113 Base Case 3.93E-04 2.61E-04 1.08E-04 1.16E+06 3.29E+06 1.42E+10 4.04E+10 7.48E+02 1.70E+03 1.21E+05 1.48E+05

114 1-hour Evacuation Delay 3.93E-04 2.61E-04 1.08E-04 1.16E+06 3.29E+06 1.42E+10 4.04E+10 7.48E+02 1.70E+03 1.21E+05 1.48E+05115 No Intermediate Phase 3.34E-04 2.42E-04 1.13E-04 1.27E+06 3.64E+06 1.10E+10 2.81E+10 1.33E+03 2.93E+03 3.17E+05 8.95E+05116 1-year Intermediate Phase 4.63E-04 2.56E-04 1.05E-04 1.13E+06 3.20E+06 2.31E+10 6.01E+10 4.67E+02 1.37E+03 3.16E+04 3.18E+04117 EPA Habitability Criterion (2 rem per year) 9.24E-04 3.53E-04 1.34E-04 1.50E+06 3.92E+06 6.60E+09 1.49E+10 4.95E+02 1.40E+03 3.88E+04 3.92E+04118 No Intermediate Phase and EPA Hab. 8.30E-04 3.68E-04 1.54E-04 1.71E+06 4.59E+06 4.75E+09 8.41E+09 7.93E+02 1.83E+03 1.35E+05 2.42E+05119 1-year Intermediate Phase and EPA Hab. 5.43E-04 3.10E-04 1.33E-04 1.31E+06 3.83E+06 1.18E+10 1.91E+10 4.67E+02 1.37E+03 3.16E+04 3.18E+04120 Larger Nonevacuating Cohort 4.59E-04 2.63E-04 1.08E-04 1.17E+06 3.30E+06 1.42E+10 4.04E+10 7.48E+02 1.70E+03 1.21E+05 1.48E+05

Mark II - Limerick

0.00E+000.00E+000.00E+00

Run

Base Model

Site File Source Term Individual Latent Cancer Fatality Risk Population Dose (person-rem) Population Subject to Long-Term Protective Actions Offsite Cost ($ 2013) Type Land (sq mi) Exceeding Long-Term Habitability Criterion

Medium - Susquehanna

Mk II - Low

 (Bin 2)

Mk II - Med

(Bin 5)

Mk II - High

(Bin 8)
I-6

Table I-6 Sensitivity calculation results for Mark II source terms with high population site file (Limerick) 

Individual 

Early Fatality 

Risk

0-1.3 mi and 

beyond 0-10 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi

121 Base Case 1.15E-06 1.81E-07 6.35E-08 4.34E+03 5.45E+03 3.81E+08 3.81E+08 5.25E-03 5.25E-03 0.00E+00 0.00E+00

122 1-hour Evacuation Delay 1.15E-06 1.81E-07 6.35E-08 4.34E+03 5.45E+03 3.81E+08 3.81E+08 5.25E-03 5.25E-03 0.00E+00 0.00E+00123 No Intermediate Phase 1.24E-06 1.91E-07 6.68E-08 4.55E+03 5.69E+03 3.81E+08 3.81E+08 5.27E-03 5.27E-03 1.60E-01 1.60E-01124 1-year Intermediate Phase 1.04E-06 1.68E-07 5.90E-08 4.04E+03 5.10E+03 3.81E+08 3.81E+08 5.25E-03 5.25E-03 0.00E+00 0.00E+00125 EPA Habitability Criterion (2 rem per year) 1.15E-06 1.81E-07 6.36E-08 4.34E+03 5.45E+03 3.81E+08 3.81E+08 5.25E-03 5.25E-03 0.00E+00 0.00E+00126 No Intermediate Phase and EPA Hab. 1.24E-06 1.91E-07 6.68E-08 4.55E+03 5.69E+03 3.81E+08 3.81E+08 5.25E-03 5.25E-03 0.00E+00 0.00E+00127 1-year Intermediate Phase and EPA Hab. 1.04E-06 1.68E-07 5.90E-08 4.05E+03 5.10E+03 3.81E+08 3.81E+08 5.25E-03 5.25E-03 0.00E+00 0.00E+00128 Larger Nonevacuating Cohort 1.20E-06 1.83E-07 6.40E-08 4.36E+03 5.47E+03 3.67E+08 3.67E+08 5.25E-03 5.25E-03 0.00E+00 0.00E+00

129 Base Case 1.35E-04 3.39E-05 1.21E-05 6.89E+05 8.88E+05 4.25E+09 4.38E+09 1.30E+02 2.21E+02 1.54E+04 1.54E+04

130 1-hour Evacuation Delay 1.35E-04 3.39E-05 1.21E-05 6.89E+05 8.88E+05 4.25E+09 4.38E+09 1.30E+02 2.21E+02 1.54E+04 1.54E+04131 No Intermediate Phase 1.12E-04 3.51E-05 1.27E-05 7.13E+05 9.24E+05 2.32E+09 2.41E+09 1.53E+02 2.44E+02 5.12E+04 5.12E+04132 1-year Intermediate Phase 1.34E-04 3.13E-05 1.10E-05 6.46E+05 8.30E+05 4.57E+09 4.65E+09 1.24E+02 2.15E+02 6.97E+03 6.97E+03133 EPA Habitability Criterion (2 rem per year) 2.34E-04 3.98E-05 1.38E-05 7.83E+05 9.82E+05 1.22E+09 1.31E+09 1.21E+02 2.12E+02 2.42E+03 2.42E+03134 No Intermediate Phase and EPA Hab. 2.40E-04 4.30E-05 1.50E-05 8.38E+05 1.05E+06 8.64E+08 9.46E+08 1.24E+02 2.15E+02 7.50E+03 7.50E+03135 1-year Intermediate Phase and EPA Hab. 1.72E-04 3.37E-05 1.18E-05 6.86E+05 8.69E+05 1.82E+09 1.91E+09 1.24E+02 2.15E+02 6.97E+03 6.97E+03136 Larger Nonevacuating Cohort 1.39E-04 3.40E-05 1.21E-05 6.93E+05 8.92E+05 4.24E+09 4.37E+09 1.30E+02 2.21E+02 1.54E+04 1.54E+04

137 Base Case 4.70E-04 3.17E-04 1.25E-04 6.11E+06 8.26E+06 8.56E+10 1.10E+11 8.53E+02 1.68E+03 7.21E+05 7.41E+05

138 1-hour Evacuation Delay 4.70E-04 3.17E-04 1.25E-04 6.11E+06 8.26E+06 8.56E+10 1.10E+11 8.53E+02 1.68E+03 7.21E+05 7.41E+05139 No Intermediate Phase 4.09E-04 2.86E-04 1.20E-04 6.72E+06 9.14E+06 7.07E+10 8.56E+10 1.40E+03 2.81E+03 1.85E+06 2.29E+06140 1-year Intermediate Phase 5.51E-04 3.09E-04 1.22E-04 5.84E+06 7.95E+06 1.34E+11 1.63E+11 5.81E+02 1.37E+03 1.49E+05 1.49E+05141 EPA Habitability Criterion (2 rem per year) 1.09E-03 4.26E-04 1.62E-04 7.89E+06 1.03E+07 3.92E+10 4.55E+10 6.11E+02 1.40E+03 2.16E+05 2.16E+05142 No Intermediate Phase and EPA Hab. 9.89E-04 4.49E-04 1.76E-04 9.25E+06 1.21E+07 3.02E+10 3.29E+10 8.97E+02 1.80E+03 8.11E+05 8.74E+05143 1-year Intermediate Phase and EPA Hab. 6.35E-04 3.77E-04 1.49E-04 6.91E+06 9.36E+06 6.92E+10 7.39E+10 5.81E+02 1.37E+03 1.49E+05 1.49E+05144 Larger Nonevacuating Cohort 5.64E-04 3.66E-04 3.20E-04 6.16E+06 8.30E+06 8.56E+10 1.10E+11 8.53E+02 1.68E+03 7.21E+05 7.41E+05

0.00E+000.00E+000.00E+00

Mark II - Limerick

Run

Base Model

Site File Source Term Individual Latent Cancer Fatality Risk Population Dose (person-rem) Population Subject to Long-Term Protective Actions Offsite Cost ($ 2013) Type Land (sq mi) Exceeding Long-Term Habitability Criterion

High - Limerick

Mk II - Low

 (Bin 2)

Mk II - Med

(Bin 5)

Mk II - High

(Bin 8)
J-1

APPENDIX J

OFFSITE CONSEQUENCE RESULTS FOR COMPOSITE 

MELCOR CASES 

Table J-1 Build-up of composite MELCOR cases and MACCS bins: Conditional 

consequences for individual early fatality, individual latent cancer fatality, and 

population dose

C101 101 0 1.28E-04 4.85E-05 2.54E-05 817,000 1,481,000

21 11 0 1.35E-04 6.41E-05 3.43E-05 1,110,000 2,030,000

24 7 0 1.21E-04 3.28E-05 1.64E-05 524,000 932,000

C101DF10 102 0 4.99E-05 9.73E-06 4.71E-06 162,100 288,500

21DF10 6 0 7.95E-05 1.61E-05 7.79E-06 253,000 450,000

24DF10 5 0 2.03E-05 3.36E-06 1.62E-06 71,200 127,000

C101DF100 103 0 1.88E-05 2.86E-06 1.32E-06 59,800 102,300

21DF100 5 0 2.03E-05 3.36E-06 1.62E-06 71,200 127,000

24DF100 4 0 1.72E-05 2.35E-06 1.01E-06 48,400 77,600

C101DF1000 104 0 1.11E-05 1.78E-06 8.53E-07 38,340 67,630

21DF1000 5 0 2.03E-05 3.36E-06 1.62E-06 71,200 127,000

24DF1000 2 0 1.90E-06 1.90E-07 8.69E-08 5,480 8,260

C102DF10 105 0 3.72E-05 7.00E-06 3.38E-06 123,143 219,286

21DF10 6 0 7.95E-05 1.61E-05 7.79E-06 253,000 450,000

22DF10 6 0 7.95E-05 1.61E-05 7.79E-06 253,000 450,000

24DF10 5 0 2.03E-05 3.36E-06 1.62E-06 71,200 127,000

27DF10 5 0 2.03E-05 3.36E-06 1.62E-06 71,200 127,000

28DF10 5 0 2.03E-05 3.36E-06 1.62E-06 71,200 127,000

30DF10 5 0 2.03E-05 3.36E-06 1.62E-06 71,200 127,000

32DF10 5 0 2.03E-05 3.36E-06 1.62E-06 71,200 127,000

C102DF100 106 0 1.30E-05 1.79E-06 8.02E-07 37,986 63,100

21DF100 5 0 2.03E-05 3.36E-06 1.62E-06 71,200 127,000

22DF100 4 0 1.72E-05 2.35E-06 1.01E-06 48,400 77,600

24DF100 4 0 1.72E-05 2.35E-06 1.01E-06 48,400 77,600

27DF100 4 0 1.72E-05 2.35E-06 1.01E-06 48,400 77,600

28DF100 3 0 6.25E-06 7.16E-07 3.21E-07 16,500 27,300

30DF100 3 0 6.25E-06 7.16E-07 3.21E-07 16,500 27,300

32DF100 3 0 6.25E-06 7.16E-07 3.21E-07 16,500 27,300

MELCOR Case MACCS

Bin 

Individual 

Early 

Fatality 

Risk 

Individual Latent Cancer Fatality 

Risk 

Population Dose 

(person-rem) 

0-1.3 mi

and beyond 0-10 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi
J-2

C102DF1000 107 0 6.30E-06 9.10E-07 4.19E-07 19,897 33,449

21DF1000 5 0 2.03E-05 3.36E-06 1.62E-06 71,200 127,000

22DF1000 2 0 1.90E-06 1.90E-07 8.69E-08 5,480 8,260

24DF1000 2 0 1.90E-06 1.90E-07 8.69E-08 5,480 8,260

27DF1000 4 0 1.72E-05 2.35E-06 1.01E-06 48,400 77,600

28DF1000 1 0 4.65E-07 4.57E-08 2.06E-08 1,620 2,380

30DF1000 2 0 1.90E-06 1.90E-07 8.69E-08 5,480 8,260

32DF1000 1 0 4.65E-07 4.57E-08 2.06E-08 1,620 2,380

C103 108 0 1.21E-04 3.28E-05 1.64E-05 524,000 932,000

27 7 0 1.21E-04 3.28E-05 1.64E-05 524,000 932,000

30 7 0 1.21E-04 3.28E-05 1.64E-05 524,000 932,000

32 7 0 1.21E-04 3.28E-05 1.64E-05 524,000 932,000

C104 109 0 3.49E-04 9.94E-05 4.87E-05 1,540,000 2,690,000

1 12 0 2.91E-04 1.01E-04 5.23E-05 1,720,000 3,090,000

4 10 0 4.06E-04 9.78E-05 4.51E-05 1,360,000 2,290,000

5 10 0 4.06E-04 9.78E-05 4.51E-05 1,360,000 2,290,000

6 12 0 2.91E-04 1.01E-04 5.23E-05 1,720,000 3,090,000

C104DF10 110 0 2.51E-04 5.96E-05 2.79E-05 862,000 1,468,500

1DF10 7 0 1.21E-04 3.28E-05 1.64E-05 524,000 932,000

4DF10 10 0 4.06E-04 9.78E-05 4.51E-05 1,360,000 2,290,000

5DF10 9 0 3.55E-04 7.50E-05 3.35E-05 1,040,000 1,720,000

6DF10 7 0 1.21E-04 3.28E-05 1.64E-05 524,000 932,000

C104DF100 111 0 2.51E-04 5.96E-05 2.79E-05 862,000 1,468,500

1DF100 7 0 1.21E-04 3.28E-05 1.64E-05 524,000 932,000

4DF100 10 0 4.06E-04 9.78E-05 4.51E-05 1,360,000 2,290,000

5DF100 9 0 3.55E-04 7.50E-05 3.35E-05 1,040,000 1,720,000

6DF100 7 0 1.21E-04 3.28E-05 1.64E-05 524,000 932,000

C104DF1000 112 0 2.51E-04 5.96E-05 2.79E-05 862,000 1,468,500

1DF1000 7 0 1.21E-04 3.28E-05 1.64E-05 524,000 932,000

4DF1000 10 0 4.06E-04 9.78E-05 4.51E-05 1,360,000 2,290,000

5DF1000 9 0 3.55E-04 7.50E-05 3.35E-05 1,040,000 1,720,000

6DF1000 7 0 1.21E-04 3.28E-05 1.64E-05 524,000 932,000

C105 113 0 2.13E-04 8.26E-05 4.33E-05 1,415,000 2,560,000

22 12 0 2.91E-04 1.01E-04 5.23E-05 1,720,000 3,090,000

23 11 0 1.35E-04 6.41E-05 3.43E-05 1,110,000 2,030,000

MELCOR Case MACCS

Bin 

Individual 

Early 

Fatality 

Risk 

Individual Latent Cancer Fatality 

Risk 

Population Dose 

(person-rem) 

0-1.3 mi

and beyond 0-10 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi
J-3

C106 114 0 1.71E-04 6.55E-05 3.43E-05 1,116,000 2,020,500

21 11 0 1.35E-04 6.41E-05 3.43E-05 1,110,000 2,030,000

22 12 0 2.91E-04 1.01E-04 5.23E-05 1,720,000 3,090,000

23 11 0 1.35E-04 6.41E-05 3.43E-05 1,110,000 2,030,000

24 7 0 1.21E-04 3.28E-05 1.64E-05 524,000 932,000

C106DF10 115 0 6.47E-05 1.29E-05 6.25E-06 207,550 369,250

21DF10 6 0 7.95E-05 1.61E-05 7.79E-06 253,000 450,000

22DF10 6 0 7.95E-05 1.61E-05 7.79E-06 253,000 450,000

23DF10 6 0 7.95E-05 1.61E-05 7.79E-06 253,000 450,000

24DF10 5 0 2.03E-05 3.36E-06 1.62E-06 71,200 127,000

C106DF100 116 0 1.80E-05 2.60E-06 1.16E-06 54,100 89,950

21DF100 5 0 2.03E-05 3.36E-06 1.62E-06 71,200 127,000

22DF100 4 0 1.72E-05 2.35E-06 1.01E-06 48,400 77,600

23DF100 4 0 1.72E-05 2.35E-06 1.01E-06 48,400 77,600

24DF100 4 0 1.72E-05 2.35E-06 1.01E-06 48,400 77,600

C106DF1000 117 0 6.50E-06 9.83E-07 4.70E-07 21,910 37,945

21DF1000 5 0 2.03E-05 3.36E-06 1.62E-06 71,200 127,000

22DF1000 2 0 1.90E-06 1.90E-07 8.69E-08 5,480 8,260

23DF1000 2 0 1.90E-06 1.90E-07 8.69E-08 5,480 8,260

24DF1000 2 0 1.90E-06 1.90E-07 8.69E-08 5,480 8,260

C107 118 0 1.21E-04 3.28E-05 1.64E-05 524,000 932,000

25 7 0 1.21E-04 3.28E-05 1.64E-05 524,000 932,000

26 7 0 1.21E-04 3.28E-05 1.64E-05 524,000 932,000

28 7 0 1.21E-04 3.28E-05 1.64E-05 524,000 932,000

30 7 0 1.21E-04 3.28E-05 1.64E-05 524,000 932,000

C107DF10 119 0 2.03E-05 3.36E-06 1.62E-06 71,200 127,000

25DF10 5 0 2.03E-05 3.36E-06 1.62E-06 71,200 127,000

26DF10 5 0 2.03E-05 3.36E-06 1.62E-06 71,200 127,000

28DF10 5 0 2.03E-05 3.36E-06 1.62E-06 71,200 127,000

30DF10 5 0 2.03E-05 3.36E-06 1.62E-06 71,200 127,000

C107DF100 120 0 6.25E-06 7.16E-07 3.21E-07 16,500 27,300

25DF100 3 0 6.25E-06 7.16E-07 3.21E-07 16,500 27,300

26DF100 3 0 6.25E-06 7.16E-07 3.21E-07 16,500 27,300

28DF100 3 0 6.25E-06 7.16E-07 3.21E-07 16,500 27,300

30DF100 3 0 6.25E-06 7.16E-07 3.21E-07 16,500 27,300

MELCOR Case MACCS

Bin 

Individual 

Early 

Fatality 

Risk 

Individual Latent Cancer Fatality 

Risk 

Population Dose 

(person-rem) 

0-1.3 mi

and beyond 0-10 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi
J-4

C107DF1000 121 0 8.24E-07 8.18E-08 3.72E-08 2,585 3,850

25DF1000 1 0 4.65E-07 4.57E-08 2.06E-08 1,620 2,380

26DF1000 1 0 4.65E-07 4.57E-08 2.06E-08 1,620 2,380

28DF1000 1 0 4.65E-07 4.57E-08 2.06E-08 1,620 2,380

30DF1000 2 0 1.90E-06 1.90E-07 8.69E-08 5,480 8,260

C108 122 0 1.28E-04 4.01E-05 2.02E-05 652,000 1,156,000

1 prior to CR 60% of 12 0 1.75E-04 6.06E-05 3.14E-05 1,032,000 1,854,000

4 prior to CR 20% of 10 0 8.12E-05 1.96E-05 9.02E-06 272,000 458,000

5 prior to CR 20% of 10 0 8.12E-05 1.96E-05 9.02E-06 272,000 458,000

6 prior to CR 60% of 12 0 1.75E-04 6.06E-05 3.14E-05 1,032,000 1,854,000

C108DF10 123 0 7.44E-05 1.85E-05 8.85E-06 277,200 480,100

1DF10 prior to 

CR 60% of 7 0 7.26E-05 1.97E-05 9.84E-06 314,400 559,200

4DF10 prior to 

CR 20% of 10 0 8.12E-05 1.96E-05 9.02E-06 272,000 458,000

5DF10 prior to 

CR 20% of 9 0 7.10E-05 1.50E-05 6.70E-06 208,000 344,000

6DF10 prior to 

CR 60% of 7 0 7.26E-05 1.97E-05 9.84E-06 314,400 559,200

C108DF100 124 0 7.44E-05 1.85E-05 8.85E-06 277,200 480,100

1DF100 prior to 

CR 60% of 7 0 7.26E-05 1.97E-05 9.84E-06 314,400 559,200

4DF100 prior to 

CR 20% of 10 0 8.12E-05 1.96E-05 9.02E-06 272,000 458,000

5DF100 prior to 

CR 20% of 9 0 7.10E-05 1.50E-05 6.70E-06 208,000 344,000

6DF100 prior to 

CR 60% of 7 0 7.26E-05 1.97E-05 9.84E-06 314,400 559,200

C108DF1000 125 0 7.44E-05 1.85E-05 8.85E-06 277,200 480,100

1DF1000 prior 

to CR 60% of 7 0 7.26E-05 1.97E-05 9.84E-06 314,400 559,200

4DF1000 prior 

to CR 20% of 10 0 8.12E-05 1.96E-05 9.02E-06 272,000 458,000

5DF1000 prior 

to CR 20% of 9 0 7.10E-05 1.50E-05 6.70E-06 208,000 344,000

6DF1000 prior 

to CR 60% of 7 0 7.26E-05 1.97E-05 9.84E-06 314,400 559,200

MELCOR Case MACCS

Bin 

Individual 

Early 

Fatality 

Risk 

Individual Latent Cancer Fatality 

Risk 

Population Dose 

(person-rem) 

0-1.3 mi

and beyond 0-10 mi 0-50 mi 0-100 mi 0-50 mi 0-100 mi
J-5

Table J-2 Build-up of composite MELCOR cases and MACCS Bins: Conditional 

consequences for offsite cost, land exceeding long-term habitability criterion, 

and population subject to long-term protective action

C101 101 4,350,000,000 6,705,000,000 238 517 27750 36200

21 11 5,960,000,000 9,720,000,000 286 673 40,500 55,800

24 7 2,740,000,000 3,690,000,000 190 361 15,000 16,600

C101DF10 102 685,000,000 815,000,000 78.5 120 1779 1779

21DF10 6 1,150,000,000 1,390,000,000 116 175 3,440 3,440

24DF10 5 220,000,000 240,000,000 41 65 118 118

C101DF100 103 180,500,000 190,500,000 32 44 62.5 62.5

21DF100 5 220,000,000 240,000,000 41 65 118 118

24DF100 4 141,000,000 141,000,000 23 23 7 7

C101DF1000 104 149,850,000 159,850,000 21 33 59 59

21DF1000 5 220,000,000 240,000,000 41 65 118 118

24DF1000 2 79,700,000 79,700,000 1 1 0 0

C102DF10 105 485,714,286 568,571,429 62.4 96.4 1,067.1 1,067.1

21DF10 6 1,150,000,000 1,390,000,000 116 175 3,440 3,440

22DF10 6 1,150,000,000 1,390,000,000 116 175 3,440 3,440

24DF10 5 220,000,000 240,000,000 41 65 118 118

27DF10 5 220,000,000 240,000,000 41 65 118 118

28DF10 5 220,000,000 240,000,000 41 65 118 118

30DF10 5 220,000,000 240,000,000 41 65 118 118

32DF10 5 220,000,000 240,000,000 41 65 118 118

C102DF100 106 133,900,000 137,014,286 20.0 23.9 20.3 20.3

21DF100 5 220,000,000 240,000,000 41 65 118 118

22DF100 4 141,000,000 141,000,000 23 23 7 7

24DF100 4 141,000,000 141,000,000 23 23 7 7

27DF100 4 141,000,000 141,000,000 23 23 7 7

28DF100 3 98,100,000 98,700,000 10 11 1 1

30DF100 3 98,100,000 98,700,000 10 11 1 1

32DF100 3 98,100,000 98,700,000 10 11 1 1

MELCOR

Case 

MACCS

Bin 

Conditional Offsite Consequences 

Offsite Cost ($ 2013) 

Land (sq mi) 

Exceeding LongTerm Habitability 

Criterion 

Population Subject to 

Long-Term Protective 

Actions 

0-50 mi 0-100 mi 0-50 mi 0-100

mi 0-50 mi 0-100 mi
J-6

C102DF1000 107 108,271,429 111,128,571 9.6 13.0 25.0 25.0

21DF1000 5 220,000,000 240,000,000 41 65 118 118

22DF1000 2 79,700,000 79,700,000 1 1 0 0

24DF1000 2 79,700,000 79,700,000 1 1 0 0

27DF1000 4 141,000,000 141,000,000 23 23 7 7

28DF1000 1 78,900,000 78,900,000 0 0 - -

30DF1000 2 79,700,000 79,700,000 1 1 0 0

32DF1000 1 78,900,000 78,900,000 0 0 - -

C103 108 2,740,000,000 3,690,000,000 190.0 361.0 15,000.0 16,600.0

27 7 2,740,000,000 3,690,000,000 190 361 15,000 16,600

30 7 2,740,000,000 3,690,000,000 190 361 15,000 16,600

32 7 2,740,000,000 3,690,000,000 190 361 15,000 16,600

C104 109 11,450,000,000 14,700,000,000 514 878 57,950 65,600

1 12 13,000,000,000 17,400,000,000 549 1,040 64,500 79,700

4 10 9,900,000,000 12,000,000,000 479 715 51,400 51,500

5 10 9,900,000,000 12,000,000,000 479 715 51,400 51,500

6 12 13,000,000,000 17,400,000,000 549 1,040 64,500 79,700

C104DF10 110 5,667,500,000 6,995,000,000 303 467 29,150 29,975

1DF10 7 2,740,000,000 3,690,000,000 190 361 15,000 16,600

4DF10 10 9,900,000,000 12,000,000,000 479 715 51,400 51,500

5DF10 9 7,290,000,000 8,600,000,000 351 429 35,200 35,200

6DF10 7 2,740,000,000 3,690,000,000 190 361 15,000 16,600

C104DF100 111 5,667,500,000 6,995,000,000 303 467 29,150 29,975

1DF100 7 2,740,000,000 3,690,000,000 190 361 15,000 16,600

4DF100 10 9,900,000,000 12,000,000,000 479 715 51,400 51,500

5DF100 9 7,290,000,000 8,600,000,000 351 429 35,200 35,200

6DF100 7 2,740,000,000 3,690,000,000 190 361 15,000 16,600

C104DF1000 112 5,667,500,000 6,995,000,000 303 467 29,150 29,975

1DF1000 7 2,740,000,000 3,690,000,000 190 361 15,000 16,600

4DF1000 10 9,900,000,000 12,000,000,000 479 715 51,400 51,500

5DF1000 9 7,290,000,000 8,600,000,000 351 429 35,200 35,200

6DF1000 7 2,740,000,000 3,690,000,000 190 361 15,000 16,600

C105 113 9,480,000,000 13,560,000,000 418 857 52,500 67,750

22 12 13,000,000,000 17,400,000,000 549 1,040 64,500 79,700

23 11 5,960,000,000 9,720,000,000 286 673 40,500 55,800

MELCOR 

Case 

MACCS 

Bin 

Conditional Offsite Consequences 

Offsite Cost ($ 2013) 

Land (sq mi) 

Exceeding LongTerm Habitability 

Criterion 

Population Subject to 

Long-Term Protective 

Actions 

0-50 mi 0-100 mi 0-50 mi 0-100

mi 0-50 mi 0-100 mi
J-7

C106 114 6,915,000,000 10,132,500,000 328 687 40,125 51,975

21 11 5,960,000,000 9,720,000,000 286 673 40,500 55,800

22 12 13,000,000,000 17,400,000,000 549 1,040 64,500 79,700

23 11 5,960,000,000 9,720,000,000 286 673 40,500 55,800

24 7 2,740,000,000 3,690,000,000 190 361 15,000 16,600

C106DF10 115 917,500,000 1,102,500,000 97 148 2,610 2,610

21DF10 6 1,150,000,000 1,390,000,000 116 175 3,440 3,440

22DF10 6 1,150,000,000 1,390,000,000 116 175 3,440 3,440

23DF10 6 1,150,000,000 1,390,000,000 116 175 3,440 3,440

24DF10 5 220,000,000 240,000,000 41 65 118 118

C106DF100 116 160,750,000 165,750,000 28 34 35 35

21DF100 5 220,000,000 240,000,000 41 65 118 118

22DF100 4 141,000,000 141,000,000 23 23 7 7

23DF100 4 141,000,000 141,000,000 23 23 7 7

24DF100 4 141,000,000 141,000,000 23 23 7 7

C106DF1000 117 114,775,000 119,775,000 11 17 30 30

21DF1000 5 220,000,000 240,000,000 41 65 118 118

22DF1000 2 79,700,000 79,700,000 1 1 0 0

23DF1000 2 79,700,000 79,700,000 1 1 0 0

24DF1000 2 79,700,000 79,700,000 1 1 0 0

C107 118 2,740,000,000 3,690,000,000 190 361 15,000 16,600

25 7 2,740,000,000 3,690,000,000 190 361 15,000 16,600

26 7 2,740,000,000 3,690,000,000 190 361 15,000 16,600

28 7 2,740,000,000 3,690,000,000 190 361 15,000 16,600

30 7 2,740,000,000 3,690,000,000 190 361 15,000 16,600

C107DF10 119 220,000,000 240,000,000 41 65 118 118

25DF10 5 220,000,000 240,000,000 41 65 118 118

26DF10 5 220,000,000 240,000,000 41 65 118 118

28DF10 5 220,000,000 240,000,000 41 65 118 118

30DF10 5 220,000,000 240,000,000 41 65 118 118

MELCOR 

Case 

MACCS

Bin 

Conditional Offsite Consequences 

Offsite Cost ($ 2013) 

Land (sq mi) 

Exceeding LongTerm Habitability 

Criterion 

Population Subject to 

Long-Term Protective 

Actions 

0-50 mi 0-100 mi 0-50 mi 0-100

mi 0-50 mi 0-100 mi
J-8

MELCOR 

Case 

MACCS 

Bin 

Conditional Offsite Consequences 

Offsite Cost ($ 2013) 

Land (sq mi) 

Exceeding LongTerm Habitability 

Criterion 

Population Subject to 

Long-Term Protective 

Actions 

0-50 mi 0-100 mi 0-50 mi

0-100

mi 0-50 mi 0-100 mi

C107DF100 120 98,100,000 98,700,000 10 11 1 1 

25DF100 3 98,100,000 98,700,000 10 11 1 1 

26DF100 3 98,100,000 98,700,000 10 11 1 1 

28DF100 3 98,100,000 98,700,000 10 11 1 1 

30DF100 3 98,100,000 98,700,000 10 11 1 1 

C107DF1000 121 79,100,000 79,100,000 0 0 0 0 

25DF1000 1 78,900,000 78,900,000 0 0 - - 

26DF1000 1 78,900,000 78,900,000 0 0 - - 

28DF1000 1 78,900,000 78,900,000 0 0 - - 

30DF1000 2 79,700,000 79,700,000 1 1 0 0 

C108 122 4,890,000,000 6,420,000,000 213 384 24,490 29,060 

1 prior to CR 60% of 

12 7,800,000,000 10,440,000,000 329 624 38,700 47,820 

4 prior to CR 20% of 

10 1,980,000,000 2,400,000,000 96 143 10,280 10,300 

5 prior to CR 20% of 

10 1,980,000,000 2,400,000,000 96 143 10,280 10,300 

6 prior to CR 60% of 

12 7,800,000,000 10,440,000,000 329 624 38,700 47,820 

C108DF10 123 1,681,500,000 2,137,000,000 99 166 8,830 9,315 

1DF10 prior to 

CR 60% of 7 1,644,000,000 2,214,000,000 114 217 9,000 9,960 

4DF10 prior to 

CR

20% of 

10 1,980,000,000 2,400,000,000 96 143 10,280 10,300 

5DF10 prior to 

CR 20% of 9 1,458,000,000 1,720,000,000 70 86 7,040 7,040 

6DF10 prior to 

CR 60% of 7 1,644,000,000 2,214,000,000 114 217 9,000 9,960 

C108DF100 124 1,681,500,000 2,137,000,000 99 166 8,830 9,315 

1DF100 prior 

to CR 60% of 7 1,644,000,000 2,214,000,000 114 217 9,000 9,960 

4DF100 prior 

to CR 

20% of 

10 1,980,000,000 2,400,000,000 96 143 10,280 10,300 

5DF100 prior 

to CR 20% of 9 1,458,000,000 1,720,000,000 70 86 7,040 7,040 

6DF100 prior 

to CR 60% of 7 1,644,000,000 2,214,000,000 114 217 9,000 9,960 

C108DF1000 125 1,681,500,000 2,137,000,000 99 166 8,830 9,315 

1DF1000 prior 

to CR 60% of 7 1,644,000,000 2,214,000,000 114 217 9,000 9,960 

4DF1000 prior 

to CR 

20% of 

10 1,980,000,000 2,400,000,000 96 143 10,280 10,300 

5DF1000 prior 

to CR 20% of 9 1,458,000,000 1,720,000,000 70 86 7,040 7,040 

6DF1000 prior 

to CR 60% of 7 1,644,000,000 2,214,000,000 114 217 9,000 9,960 
Washington, D.C. 20555-0001

Same as above

Technical

March 2018



NUREG-2206 Technical Basis for the Containment Protection and Release Reduction Rulemaking for 

Boiling Water Reactors with Mark I and Mark II Containments

March 2018