Consequence Study of 

a Beyond-Design-Basis 

Earthquake Affecting 

the Spent Fuel Pool for 

a U.S. Mark I Boiling 

Water Reactor

Office of Nuclear Regulatory Research

NUREG-2161
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Consequence Study of 

a Beyond-Design-Basis 

Earthquake Affecting 

the Spent Fuel Pool for 

a U.S. Mark I Boiling 

Water Reactor

Manuscript Completed: October 2013

Date Published: September 2014

Prepared by:

Andrew Barto, Y. James Chang, Keith Compton,

Hossein Esmaili, Donald Helton, Andrew Murphy,

Andrew Nosek, Jose Pires, Fred Schofer, and 

Brian Wagner

Donald Algama, NRC Project Manager

Katie Wagner, Former NRC Project Manager

Office of Nuclear Regulatory Research

NUREG-2161
ii
iii

ABSTRACT

The U.S. Nuclear Regulatory Commission performed this consequence study to continue its 

examination of the risks and consequences of postulated spent fuel pool accidents. The study 

provides publicly available consequence estimates of a hypothetical spent fuel pool accident 

initiated by a low likelihood seismic event at a specific reference plant. The study compares 

high-density and low-density loading conditions and assesses the benefits of post 9/11 

mitigation measures. Past risk studies have shown that storage of spent fuel in a high-density 

configuration is safe and risk of a large release due to an accident is very low. This study’s 

results are consistent with earlier research conclusions that spent fuel pools are robust 

structures that are likely to withstand severe earthquakes without leaking. The NRC continues 

to believe, based on this study and previous studies that high density storage of spent fuel in 

pools protects public health and safety. The study’s results will inform a broader regulatory 

analysis of the spent fuel pools at U.S. nuclear reactors as part of the Japan Lessons-learned 

Tier 3 plan.
iv
v

FOREWORD

U.S. nuclear power plants are required to be designed with appropriate consideration of the 

most severe natural phenomena (e.g. floods, earthquakes, tornadoes) historically reported for 

their location and surrounding regions, with sufficient margin, to ensure that important safety 

functions can be performed. As part of our mission to protect public health and safety, the U.S. 

Nuclear Regulatory Commission (NRC) uses advanced computer modeling and other 

techniques to study more severe, and highly unlikely, events that go beyond what the plant was 

designed to withstand to estimate risk to the public and to explore and ensure safety margins.

On March 11, 2011, the Tohoku earthquake and subsequent tsunami in Japan resulted in 

significant damage to the site of the Fukushima Dai-ichi nuclear power station. Although the 

spent fuel pools and the used fuel assemblies stored in the pools remained intact at the plant, 

the event led to questions about the safe storage of spent fuel and whether the NRC should 

require the expedited transfer of spent fuel from pools to dry cask storage containers at U.S. 

nuclear power plants.

This report documents the Office of Nuclear Regulatory Research’s consequence study that 

continues our examination of the risks and consequences of postulated spent fuel pool 

accidents. A spent fuel pool’s robust concrete structure and stainless steel liner keep more than 

20 feet of water above the spent fuel stored within it ensuring ample cooling for the spent fuel 

and adequate radiation shielding for plant personnel. About every two years, some used fuel is 

removed from the reactor and placed into the spent fuel pool. The used fuel most recently 

removed from a reactor is radiologically and thermally “hot”. The hot fuel is distributed 

throughout the pool and is surrounded by older, cooler used fuel. After used fuel has cooled in 

the spent fuel pool for more than about five years, it has radiologically decayed such that it can 

be moved to dry storage casks for longer term storage.

This study compared potential accident consequences from a pool nearly filled with spent fuel 

and a pool in which fuel that has cooled sufficiently has been removed. The staff first evaluated 

whether a severe, though unlikely, earthquake would damage the spent fuel pool to the point of 

leaking. In order to assess the consequences that might result from a spent fuel pool leak, the 

study assumed seismic forces greater than the maximum earthquake reasonably expected to 

occur at the reference plant location. The NRC expects that the ground motion used in this 

study is more challenging for the spent fuel pool structure than that experienced at the 

Fukushima Daiichi nuclear power plant from the earthquake that occurred off the coast of Japan 

on March 11, 2011. That earthquake did not result in any spent fuel pool leaks. In the small 

likelihood that such an extreme earthquake caused a leak, the staff then analyzed where the 

leak would be expected, the size of the leak, and how the spent fuel could overheat and 

potentially release radioactive material into the environment. Finally, the staff analyzed what the 

public health and environmental effects of a radiological release would be in the area 

surrounding the plant. In order to estimate the hypothetical consequences, the staff analyzed 

scenarios where some preplanned and improvised mitigative actions by the emergency 

response organization were either not successful or not implemented.

The study results for the specific reference plant and earthquake analyzed are consistent with 

past studies’ conclusions that spent fuel pools are likely to withstand severe earthquakes 

without leaking. Past studies considered a wider range of earthquakes than this study. In the 

unlikely situation that a leak occurs, this study shows that for the scenarios and spent fuel pool 

studied, spent fuel is only susceptible to a radiological release within a few months after the fuel 
vi

is moved from the reactor into the spent fuel pool. After that time, the spent fuel is coolable by 

air for at least 72 hours. This study shows the likelihood of a radiological release from the spent 

fuel after the analyzed severe earthquake at the reference plant to be about one time in 10 

million years or lower. If a leak and radiological release were to occur, this study shows that the 

individual cancer fatality risk for a member of the public is several orders of magnitude lower 

than the Commission’s Quantitative Health Objective of two in one million (2x10-6

/year). For 

such a radiological release, this study shows public and environmental effects are generally the 

same or smaller than earlier studies.

The Office of Nuclear Reactor Regulation’s regulatory analysis for this study indicates that

expediting movement of spent fuel from the pool does not provide a substantial safety 

enhancement for the reference plant. The NRC plans to use the insights from this analysis to 

inform a broader regulatory analysis of the spent fuel pools at all U.S. operating nuclear reactors

as part of its Japan Lessons-learned Tier 3 plan. The NRC continues to believe, based on this 

study and previous studies that spent fuel pools continue to provide adequate protection of

public health and safety.
vii

CONTENTS

Page

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

FOREWORD.................................................................................................................................v

LIST OF FIGURES ..................................................................................................................... xv

LIST OF TABLES....................................................................................................................... xix

EXECUTIVE SUMMARY ..........................................................................................................xxiii

ACKNOWLEDGMENTS ...........................................................................................................xxxi

ABBREVIATIONS AND ACRONYMS.....................................................................................xxxiii

1. INTRODUCTION AND BACKGROUND................................................................................1

 1.1 Project Impetus...............................................................................................................3

 1.2 Technical Approach........................................................................................................4

 1.3 Site Specificity and Familiarization .................................................................................5

 1.4 Basic Scenario Development .........................................................................................6

 1.5 Rationale for Focusing on Consequences of a Seismic Hazard.....................................6

 1.6 Operating Cycle Phase Approach ..................................................................................9

 1.7 Overview of Past Studies .............................................................................................10

 1.8 Potential Follow-On Work and Related Activities .........................................................17

 1.9 Layout of Remainder of This Report.............................................................................18

2. MAJOR ASSUMPTIONS.....................................................................................................19

 2.1 Study Assumptions.......................................................................................................19

 2.2 Multi-Unit Considerations .............................................................................................28

 2.3 Inadvertent Criticality ....................................................................................................29

3. SEISMIC HAZARD CHARACTERIZATION.........................................................................31

 3.1 Basis for Probabilistic Estimates ..................................................................................31

 3.2 Characterization of the Event Likelihood ......................................................................34

 3.3 Characterization of the Ground Motion Response Spectra ..........................................36

4. STRUCTURAL ANALYSIS AND RELATED INITIAL DAMAGE CHARACTERIZATION.....39

 4.1 Damage States for the Spent Fuel Pool Structure .......................................................39

 4.1.1 Approach and Seismic Loads ............................................................................39

 4.1.2 Description of the Spent Fuel Pool Structure.....................................................47

 4.1.3 Finite Element Model Description ......................................................................51

 4.1.4 Finite Element Analysis Results for the Spent Fuel Pool...................................56
viii

 4.1.5 Damage States ...................................................................................................61

4.2 Other Damage States...................................................................................................66

4.3 Review of Spent Fuel Pool Performance under Recent Major 

 Earthquakes in Japan...................................................................................68

5. SCENARIO DELINEATION AND PROBABILISTIC CONSIDERATIONS...........................77

 5.1 Representative Operating Cycle Characterization .......................................................77

 5.2 Operating Cycle Phase Specification ...........................................................................78

 5.3 Treatment of Mitigation.................................................................................................80

 5.3.1 Approach Details and Assumptions...................................................................83

 5.3.2 Rationale for Producing Unmitigated Results ....................................................86

 5.4 Refueling Floor Dose Rate Analysis Using SCALE......................................................87

 5.5 Discussion of Repair and Recovery .............................................................................90

 5.6 Scenario Development .................................................................................................90

 5.6.1 Identification of Key Events ...............................................................................90

 5.6.2 Scenario Calculation Matrices ...........................................................................91

 5.6.3 Summary of Event Split Fractions .....................................................................91

6. ACCIDENT PROGRESSION ANALYSIS............................................................................95

 6.1 Modeling Spent Fuel Pools with MELCOR...................................................................95

 6.1.1 Overview and Experimental/Analytical Basis.....................................................95

 6.1.2 Heat Transfer Modeling within Spent Fuel Pool and to Surrounding Walls .......98

 6.1.3 Modeling of Mitigative Sprays............................................................................99

 6.1.4 Modeling of Fuel Collapse and Baseplate Failure ...........................................100

 6.1.5 Radionuclide Transport Modeling and Treatment of Hydrogen .......................101

 6.2 Description of MELCOR Models.................................................................................106

 6.2.1 High-Density Loading during Outage................................................................111

 6.2.2 High-Density Loading Postoutage ....................................................................115

 6.2.3 Low-Density Loading during Outage ...............................................................115

 6.2.4 Low-Density Loading Postoutage ....................................................................117

 6.3 MELCOR Analysis Results.........................................................................................118

 6.3.1 Sequences That Do Not Lead to a Release ....................................................118

 6.3.2 Sequences That Do Lead to a Release...........................................................129

 6.3.3 Source Terms for Offsite Consequence Analysis ............................................149

 6.3.4 Accumulation of Water Elsewhere in the Reactor Building..............................150

7. OFFSITE CONSEQUENCE ANALYSIS............................................................................153

 7.1 Offsite Consequence Modeling ..................................................................................153
ix

 7.1.1 Radiological Source Term ................................................................................153

 7.1.2 Atmospheric Modeling and Meteorology ..........................................................155

 7.1.3 Exposure, Dosimetry, and Health Effects Modeling..........................................156

 7.1.4 Emergency Response Modeling.......................................................................157

 7.1.5 Long-Term Protective Action Modeling.............................................................162

 7.2 Offsite Consequence Results .....................................................................................163

 7.2.1 Individual Early Fatality Risk.............................................................................166

 7.2.2 Individual Latent Cancer Fatality Risk ..............................................................166

 7.2.3 Land Contamination..........................................................................................170

 7.2.4 Displaced Individuals ........................................................................................172

 7.3 Offsite Consequence Comparison..............................................................................173

8. HUMAN RELIABILITY ANALYSIS.....................................................................................177

 8.1 Summary of Spent Fuel Pool Study Analysis Results Relevant to Human 

 Reliability Analysis......................................................................................................179

 8.1.1 High Level Scenarios Classification.................................................................179

 8.1.2 Key Factors Affecting Available Time for Mitigation.........................................180

 8.2 Staffing, Mitigation Equipment, Strategies, and Procedures ......................................183

 8.2.1 Staffing, Procedures, Training, and Response Time .......................................183

 8.2.2 Mitigation Equipment .......................................................................................186

 8.2.3 Mitigation Strategies ........................................................................................187

 8.3 Study Framework, Scope, and Approach...................................................................190

 8.3.1 Study Framework and Scope ..........................................................................190

 8.3.2 Approach to Human Error Probability Estimates .............................................191

 8.3.2.1 A Two-Phased Approach to HEP Estimates.......................................191

 8.3.2.2 Step 1—Identify the Time Required ...................................................192

 8.3.2.3 Step 2—Identify the Damage States and Available Time ..................193

 8.3.2.4 Step 3—Estimate Basic HEPs of a Single Unit Event........................194

 8.3.2.5 Step 4—Additional Feasibility Considerations ...................................198

 8.4 Discussion and Summary...........................................................................................200

9. CONSIDERATION OF UNCERTAINTY ............................................................................203

 9.1 Sensitivity to Hydrogen Combustion (MELCOR)........................................................203

 9.2 Sensitivity to 1x8 Fuel Assembly Pattern (MELCOR).................................................208

 9.3 Sensitivity to a Contiguous (Uniform) Fuel Pattern during an Outage 

 (MELCOR/MACCS2)..................................................................................................213

 9.4 Sensitivity to Multiunit or Concurrent Accident Events (MELCOR).............................223
x

 9.5 Sensitivity to Molten Core-Concrete Interaction (MELCOR/MACCS2).......................227

 9.6 Sensitivity to Radiative Heat Transfer (MELCOR)......................................................230

 9.7 Sensitivity to Land Contamination (MACCS2)............................................................230

 9.8 Sensitivity to Time Truncation (MELCOR/MACCS2)..................................................232

 9.9 Sensitivity to Reactor Building Leakage (MELCOR)...................................................234

10. ASSESSMENT OF PREVIOUS STUDIES OF SAFETY CONSEQUENCES 

 ASSOCIATED WITH LOADING, TRANSFER, AND LONG-TERM DRY STORAGE....235

 10.1 Introduction............................................................................................................235

 10.2 Previous Spent Fuel Pool Studies.........................................................................235

 10.2.1 Quantitative Comparison of Spent Fuel Pool Analytical Results ..............236

 10.2.2 Comparison of SFPS Results to previous Spent Fuel Pool Studies.........238

 10.3 Previous Dry Cask Storage Studies ......................................................................239

 10.3.1 Supplemental Analyses .............................................................................239

 10.3.2 Quantitative Comparison of Dry Cask Storage and SFPS 

 Analytical Results........................................................................ 244

 10.3.3 Comparison of SFPS Results to Previous and Supplemental 

 Cask Studies..............................................................................................244

 10.4 Summary of Assessment of Previous Studies.......................................................245

11. REGULATORY ANALYSIS SCREENING SUMMARY ..................................................247

 11.1 Decision Rationale ...............................................................................................247

 11.1.1 Comparison to the Safety Goal Policy Statement.....................................247

 11.1.2 Cost-Benefit Analysis................................................................................248

 11.2 Further Actions .....................................................................................................249

12. SUMMARY AND CONCLUSIONS.................................................................................251

 12.1 Summary ...............................................................................................................251

 12.2 Conclusions...........................................................................................................255

13. REFERENCES...............................................................................................................259

APPENDIX A: DETAILED EMERGENCY RESPONSE MODELS............................................ A-1

 A.1 Evacuation Model 1: WinMACCS response parameters for sequences where 

 PAGs are not exceeded beyond the EPZ. ............................................................ A-1

 A.2 Evacuation Model 2: WinMACCS response parameters for late release 
xi

 sequences where the PAG is exceeded beyond the EPZ..................................... A-3

 A.3 Evacuation Model 3: WinMACCS response parameters for early release 

 sequences where the PAG is exceeded beyond the EPZ..................................... A-6

APPENDIX B: A QUALITATIVE RISK COMPARISON OF SPENT FUEL STORAGE 

STRATEGIES ...........................................................................................................................B-1

 B.1 Introduction ............................................................................................................. B-1

 B.2 Spent Fuel Storage Strategies................................................................................ B-1

 B.3 Spent Fuel Storage Stages..................................................................................... B-1

 B.4 Risk of Spent Fuel Storage ..................................................................................... B-2

 B.5 Risk during Each Stage .......................................................................................... B-3

 B.6 Total Risk over Time ............................................................................................... B-5

 B.7 Availability of Information ........................................................................................ B-6

 B.7.1 Cask Risks (rcask,load and rcask,store)................................................................... B-7

 B.7.2 Number of Casks (Nload and Nstore)................................................................. B-7

 B.7.3 Pool Uncovery Frequency from Cask Drop Events 

 (Nload, Pdrop and Pdamage).................................................................................. B-8

 B.7.4 Pool Uncovery Frequency from Seismic Events (Fseismic).............................. B-9

 B.7.5 Pool Uncovery Frequency from Other Events (Fother).................................. B-10

 B.7.6 Pool Consequences (Cuncovery)..................................................................... B-10

 B.7.7 Other Spent Fuel Risk Considerations ........................................................ B-11

 B.8 Conclusions .......................................................................................................... B-11

APPENDIX C: CORRESPONDENCE ......................................................................................C-1

APPENDIX D: REGULATORY ANALYSIS AND BACKFITTING DISCUSSION TO 

 DETERMINE THE SAFETY BENEFIT OF EXPEDITED TRANSFER 

 OF SPENT FUEL AT A REFERENCE PLANT .................................................D-1

 D.1 INTRODUCTION ...................................................................................................D-3

 D.1.1 Statement of the Problem ............................................................................D-3

 D.1.2 Objective of Proposed Action.......................................................................D-4

 D.2 IDENTIFICATION AND PRELIMINARY ANALYSIS OF ALTERNATIVE 

 APPROACHES ......................................................................................................D-5

 D.2.1 Alternative 1 – Regulatory Baseline – Maintain the Existing Spent Fuel

 Storage Requirements.................................................................................D-5

 D.2.2 Alternative 2 – Low-density Spent Fuel Pool Storage..................................D-6

 D.3 ESTIMATION AND EVALUATION OF VALUES AND IMPACTS..........................D-6

 D.3.1 Identification of Affected Attributes...............................................................D-7

 D.3.2 Methodology Overview.................................................................................D-8
xii

 D.3.2.1 Analysis Model ...............................................................................D-9

 D.3.2.1.1 Baseline for the Analysis...............................................D-9

 D.3.2.1.2 Discount Rates............................................................D-10

 D.3.2.2 Data..............................................................................................D-10

 D.3.2.2.1 Spent Fuel Pool Initiator Release Frequency..............D-10

 D.3.2.2.2 Duration of On-site Spent Fuel Storage Risk ..............D-15

 D.3.2.2.3 Cost/Benefit Inflators...................................................D-15

 D.3.2.2.4 Dollar per Person-Rem Conversion Factor .................D-16

 D.3.2.2.5 Onsite Property Decontamination, Repair, and 

 Refurbishment Costs ..................................................D-16

 D.3.2.2.6 Replacement Energy Costs ........................................D-17

 D.3.2.2.7 Occupational Worker Exposure (Accident) .................D-17

 D.3.2.2.8 Long-Term Habitability Criteria ...................................D-21

 D.3.2.2.9 Other Key Data ...........................................................D-23

 D.3.2.3 Assumptions.................................................................................D-23

 D.3.2.3.1 Projected Number of Outages and Spent Fuel 

 Assemblies .................................................................D-23

 D.3.2.3.2 Dry Storage Capacity ..................................................D-24

 D.3.2.3.3 Fuel Assembly Decay Heat as a Function of 

 Burnup and Cooling Time ...........................................D-25

 D.3.2.3.4 Dry Storage Upfront Costs ..........................................D-25

 D.3.2.3.5 Incremental Costs Associated with Earlier 

 DSC Purchase and Loading .......................................D-26

 D.3.2.3.6 Incremental Annual ISFSI 

 Operating Costs..........................................................D-26

 D.3.2.3.7 Dry Storage Occupational Exposure 

 (Routine).....................................................................D-27

 D.3.2.3.8 Number of Dry Storage Casks ...................................D-28

 D.3.3 Sensitivity Analysis.....................................................................................D-29

 D.3.3.1 Present Value Calculations ..........................................................D-29

 D.3.3.2 Dollar per Person-Rem Conversion Factor..................................D-29

 D.3.3.3 Replacement Energy Costs .........................................................D-30

 D.3.3.4 Consequences Extending Beyond 50 Miles.................................D-30

 D.3.4 Alternative – Low-Density Spent Fuel Pool Storage ..................................D-31

 D.3.4.1 Public Health (Accident)...............................................................D-31

 D.3.4.2 Occupational Health (Accident)....................................................D-32

 D.3.4.3 Occupational Health (Routine).....................................................D-33

 D.3.4.4 Offsite Property ............................................................................D-34
xiii

 D.3.4.5 Onsite Property ...........................................................................D-35

 D.3.4.6 Industry Implementation..............................................................D-35

 D.3.4.7 Industry Operation.......................................................................D-36

 D.3.4.8 NRC Implementation...................................................................D-37

 D.3.4.9 NRC Operation............................................................................D-37

 D.3.4.10 Other Considerations ................................................................D-37

 D.3.4.10.1 Modeling Uncertainties...........................................D-37

 D.3.4.10.2 Cask Handling Risk................................................D-37

 D.3.4.10.3 Mitigating Strategies...............................................D-38

 D.3.4.10.4 Other Favorable Spent Fuel Loading 

 Configurations........................................................D-39

 D.4 PRESENTATION OF RESULTS ..........................................................................D-39

 D.4.1 Regulatory Analysis ...................................................................................D-39

 D.4.1.1 Summary Table............................................................................D-39

 D.4.1.2 Implementation and Operation Costs...........................................D-41

 D.4.1.3 Total Benefits and Cost Offsets ...................................................D-41

 D.4.1.4 Sensitivity Analysis.......................................................................D-41

 D.4.1.4.1 Dollar per Person-Rem Conversion Factor .................D-42

 D.4.1.4.2 Consequences Extending Beyond 50 Miles................D-42

 D.4.1.4.3 Combined Effect of Consequences Extending 

 Beyond 50 Miles and Dollar per Person-Rem 

 Conversion Factor ......................................................D-43

 D.4.2 Backfit Analysis ..........................................................................................D-44

 D.4.2.1 General Description of the Activity Required at the Reference 

 Plant to Complete the Backfit .....................................................D-45

 D.4.2.2 Potential Change in the Risk to the Public from the 

 Accidental Offsite Release of Radioactive Material ....................D-45

 D.4.2.3 Potential Impact on Radiological Exposure of 

 Facility Employees......................................................................D-45

 D.4.2.4 Installation and Continuing Costs Associated 

 with the Backfit, including the Cost of Facility Downtime 

or the Cost of Construction Delay...............................................D-46

 D.4.2.5 Potential Safety Impact of Changes in Plant 

 or Operational Complexity, including the Relationship to 

 Proposed and Existing Regulatory Requirements ......................D-46

 D.4.2.6 Estimated Resource Burden on the NRC 

 Associated with the Proposed Backfit and the Availability of 

Such Resources. ........................................................................D-46
xiv

 D.4.2.7 Potential Impact of Differences in Facility Type, Design, 

 or Age on the Relevancy and Practicality of the Proposed 

 Action..........................................................................................D-47

 D.4.2.8 Whether the Proposed Backfit is Interim or Final and, 

 if Interim, the Justification for Imposing the Proposed 

 Backfit on an Interim Basis ..........................................................D-47

 D.4.2.9 Other Information Relevant and Material to the 

 Proposed Backfit..........................................................................D-47

 D.4.3 Disaggregation ...........................................................................................D-48

 D.4.4 Safety Goal Evaluation...............................................................................D-49

 D.4.5 CRGR Results............................................................................................D-50

 D.5 DECISION RATIONALE.......................................................................................D-50

 D.5.1 Regulatory Analysis ...................................................................................D-50

 D.5.2 Backfit Analysis ...........................................................................................D-50

 D.5.3 Conclusion ..................................................................................................D-51

 D.6 REFERENCES .......................................................................................................D-52

APPENDIX E: SFPS PUBLIC COMMENTS SUMMARY.......................................................... E-1
xv

LIST OF FIGURES

Figure 1 Graphical overview of significant SFP-related activities ........................................10

Figure 2 Comparison of PGA exceedance frequencies at the reference plant....................32

Figure 3 Comparison of spectral exceedance frequencies at the reference plant (rock 

 hazard curves) .......................................................................................................32

Figure 4 Comparison of annual PGA exceedance frequencies for U.S. Mark I reactors 

 (USGS 2008 model) (rock hazard curves).............................................................33

Figure 5 Comparison of annual exceedance frequencies for 1 Hz spectral accelerations 

 for U.S. Mark I reactors (USGS 2008 model) (rock hazard curves).......................33

Figure 6 Comparison of seismic initiating event frequencies...............................................34

Figure 7 Input acceleration response spectrum and SSE (Horizontal Ground Motion) .......36

Figure 8 Response spectrum for 5-percent damping scaled to 1.0 g PGA: SSE, 

 NUREG/CR-4550 (NUREG-1150 PRA), and this study (GI-199) .........................37

Figure 9 Horizontal response spectrum (5-percent damping): SSE, SFP Study and 

 NUREG-1150 PRA (for 0.71g PGA) ......................................................................38

Figure 10 Vertical ISRS for 5-percent damping at Elevation 195 ft (bottom of the SFP) .......45

Figure 11 Horizontal ISRS for 5-percent damping midway between Elevation 195 ft 

 and Elevation 234 ft (midheight of the SFP) ..........................................................45

Figure 12 Vertical ISRS for 5-percent and 10-percent damping at Elevation 195 ft 

 (bottom of the SFP)................................................................................................46

Figure 13 Horizontal ISRS for 5-percent and 10-percent damping midway between 

 Elevation 195 ft and Elevation 234 ft (midheight of the SFP) ................................46

Figure 14 Schematic diagram of the reactor building foundation near the drywell ................47

Figure 15 SFP details in cutout of 3D CAD model.................................................................48

Figure 16 Cutouts of 3D CAD models of the reactor building and SFP.................................49

Figure 17 Finite element model of the SFP structure with labels for the floor and walls........49

Figure 18 Outline of detailed finite element model of the SFP liner representing 

 attachments to the SFP floor and walls (E and W walls) .......................................50

Figure 19 Cutouts of 3D finite element model showing components included in the model..51

Figure 20 Estimated peak vertical accelerations (in/sec2

) of the SFP floor from 

 response spectrum analysis and vertical ISRS as input (1.0g PGA and 

 5-percent damping)................................................................................................56

Figure 21 Estimated peak hydrodynamic pressures (psi) on the SFP floor from response 

 spectrum analysis and vertical ISRS as input 

 (1.0g PGA and 5-percent damping).......................................................................56

Figure 22 Contours of vertical displacements (mm) of the SFP floor and walls......................57

Figure 23 Contours of vertical displacement (mm) of the SFP walls ......................................58

Figure 24 Liner strains (overall response not fully accounting for strain concentrations) .......58

Figure 25 Region of concrete cracking initiation at the floor-wall junction ..............................59

Figure 26 SFP displacements (mm) with detailed liner insert.................................................60

Figure 27 Strain concentrations in the SFP liner ....................................................................60

Figure 28 Moderate leakage flow rate (through concrete cracks)...........................................64

Figure 29 Small leakage flow rate (through localized steel tears) ..........................................66

Figure 30 Vertical response spectra: Kashiwazaki-Kariwa Units 1 and 4 (foundation 

 level) and SFP study (free-field) .............................................................................72

Figure 31 Horizontal response spectra: Fukushima Daiichi Units 1 and 4 (foundation) 

 and SFP study (free-field).......................................................................................73
xvi

Figure 32 Vertical response spectra: Fukushima Daiichi Unit 4 (foundation) and SFP 

 study (free-field)......................................................................................................73

Figure 33 Vertical ISRS for Kashiwazaki-Kariwa Units 1 and 4 and for the SFP Study..........74

Figure 34 Illustration of SFP patterns ....................................................................................78

Figure 35 Cutaway depiction of a lifted assembly with water level at the top of the racks.....88

Figure 36 Approximate dose rate of elevation contours, water at top of fuel hardware,

 around the time of defueling (rem per hour). .........................................................89

Figure 37 MELCOR modeling of heat transfer paths.............................................................99

Figure 38 Spray model for SFP analysis .............................................................................100

Figure 39 Typical SFP rack cut away cross sections...........................................................107

Figure 40 MELCOR nodalization of the whole pool high density model ..............................108

Figure 41 MELCOR nodalization of the assemblies (only two rings shown)........................109

Figure 42 MELCOR reactor building model .........................................................................111

Figure 43 SFP and reactor connection model during outage ..............................................112

Figure 44 Layout of assemblies for OCP1 high density (1x4) model...................................113

Figure 45 MELCOR radial radiative coupling scheme .........................................................114

Figure 46 Layout of assemblies for OCP2 high-density (1x4) model...................................114

Figure 47 Layout of assemblies for OCP1 low-density model .............................................116

Figure 48 Layout of assemblies for OCP2 low-density model .............................................117

Figure 49 Water level for boiloff scenarios...........................................................................119

Figure 50 Water level for mitigated low-density OCP1 (small leak) scenario ......................120

Figure 51 Flow rates for mitigated low-density OCP1 (small leak) scenario........................121

Figure 52 Water level for unmitigated high-density moderate leak (OCP4).........................122

Figure 53 Ring 1 temperature for unmitigated high-density moderate leak (OCP4)............122

Figure 54 Water level for unmitigated high-density small leak (OCP4)................................123

Figure 55 Ring 1 temperature for unmitigated high-density small leak (OCP4)...................123

Figure 56 Ring 1 temperature for unmitigated low-density moderate leak (OCP4) .............124

Figure 57 Ring 1 temperature for unmitigated low-density small leak (OCP4) ....................124

Figure 58 Water level for mitigated high-density moderate leak (OCP3).............................126

Figure 59 Water flow rates for mitigated high-density moderate leak (OCP3).....................126

Figure 60 Ring 1 clad temperatures for mitigated (simple flow regime active) 

 high-density moderate leak (OCP3).....................................................................127

Figure 61 Ring 1 clad temperatures for mitigated (simple flow regime inactive) 

 high-density moderate leak (OCP3).....................................................................127

Figure 62 Flow rates for mitigated high-density moderate leak (OCP3) with late 

 actuation of sprays...............................................................................................128

Figure 63 Ring 1 clad temperatures for mitigated (simple flow regime active) 

 high-density moderate leak (OCP3) with late actuation of sprays .......................128

Figure 64 Ring 1 clad temperatures for mitigated (simple flow regime inactive) 

 high-density moderate leak (OCP3) with late actuation of sprays .......................129

Figure 65 Water level for unmitigated high-density moderate leak (OCP1).........................131

Figure 66 SFP power for unmitigated high-density moderate leak (OCP1).........................132

Figure 67 Ring 1 clad temperature for unmitigated high-density moderate leak (OCP1) ....132

Figure 68 Midplane clad temperature for unmitigated high-density moderate leak (OCP1) 133

Figure 69 Midplane rack temperature for unmitigated high-density moderate leak (OCP1) 133

Figure 70 SFP wall liner temperature for unmitigated high-density moderate leak (OCP1) 134

Figure 71 Reactor building DF for unmitigated high-density moderate leak (OCP1)...........134

Figure 72 Cesium environmental release fraction for unmitigated high-density 

 moderate leak (OCP1).........................................................................................135

Figure 73 Ring 1 clad temperature for unmitigated low-density moderate leak (OCP1)......135

Figure 74 Midplane clad temperature for unmitigated low-density moderate leak (OCP1)..136
xvii

Figure 75 Cesium environmental release fraction for unmitigated low-density 

 moderate leak (OCP1).........................................................................................136

Figure 76 Water level for mitigated high-density moderate leak (OCP1).............................138

Figure 77 Ring 1 clad temperature for mitigated high density moderate leak (OCP1).........138

Figure 78 Cesium environmental release fraction for mitigated high density moderate 

 leak (OCP1) .........................................................................................................139

Figure 79 Ring 1 clad temperature for mitigated low density moderate leak (OCP1)..........139

Figure 80 Cesium environmental release fraction for mitigated low density moderate 

 leak (OCP1) .........................................................................................................140

Figure 81 Water level for unmitigated high-density small leak (OCP2)................................141

Figure 82 SFP power for unmitigated high-density small leak (OCP2)................................141

Figure 83 Ring 1 clad temperature for unmitigated high-density small leak (OCP2) ...........142

Figure 84 Midplane clad temperature for unmitigated high-density small leak (OCP2).......142

Figure 85 Reactor building mole fraction for unmitigated high-density small leak (OCP2)..143

Figure 86 Reactor building DF for unmitigated high-density small leak (OCP2)..................143

Figure 87 Cesium environmental release fraction for unmitigated high-density small 

 leak (OCP2) .........................................................................................................144

Figure 88 Ring 1 clad temperature for unmitigated low-density small leak (OCP2).............144

Figure 89 Midplane clad temperature for unmitigated low-density small leak (OCP2) ........145

Figure 90 Ring 1 clad temperature for unmitigated high-density moderate leak (OCP3) ....146

Figure 91 SFP wall liner temperature for unmitigated high-density moderate leak (OCP3) 147

Figure 92 Reactor building mole fractions for unmitigated high-density moderate leak 

 (OCP3) ................................................................................................................147

Figure 93 Ring 1 clad temperature for unmitigated low-density moderate leak (OCP3)......148

Figure 94 MELCOR default ruthenium release for unmitigated high-density moderate 

 leak (OCP3) .........................................................................................................148

Figure 95 MELCOR enhanced ruthenium release under air oxidizing conditions for 

 unmitigated high density moderate leak (OCP3) ................................................149

Figure 96 Individual Latent Cancer Fatality Risk (per year).................................................169

Figure 97 Conditional Individual LCF Risk for Different Dose Truncations and Distances ..170

Figure 98 Approximate dose rate of elevation contours, water at the top of fuel 

 hardware, around the time of defueling (rem per hour). ......................................182

Figure 99 The refueling floor temperature of OCP1 small leak scenarios. ..........................182

Figure 100 The refueling floor temperature of OCP 2 small leak scenarios. .........................183

Figure 101 Generalized guidance for SFP makeup and spray decisions ..............................190

Figure 102 The study framework ...........................................................................................191

Figure 103 Flow chart for estimating HEPs for a single reactor unit event ............................196

Figure 104 The gap release probability assessments given damage to multiple reactors 

 and SFPs. ............................................................................................................199

Figure 105 Reactor building mole fraction for unmitigated high-density moderate leak 

 (OCP2).................................................................................................................204

Figure 106 Reactor building mole fraction for unmitigated high-density moderate leak 

 (OCP2-S) .............................................................................................................204

Figure 107 SFP power for unmitigated high-density moderate leak (OCP2)..........................205

Figure 108 SFP power for unmitigated high-density moderate leak (OCP2-S) ......................205

Figure 109 Ring 1 clad temperature for unmitigated high-density moderate leak (OCP2) .....206

Figure 110 Ring 1 clad temperature for unmitigated high-density moderate leak (OCP2-S)..206

Figure 111 Cesium environmental release fraction for unmitigated high density moderate 

 leak (OCP2) ..........................................................................................................207

Figure 112 Cesium environmental release fraction for unmitigated high density 

 moderate leak (OCP2-S) ......................................................................................207
xviii

Figure 113 Layout of assemblies for OCP3 high density (1x8) model...................................208

Figure 114 Ring 1 clad temperature for unmitigated high-density small leak (OCP3; 1x8) ...209

Figure 115 Midplane clad temperature for unmitigated high-density small leak 

 (OCP3; 1x8).........................................................................................................210

Figure 116 Ring 1 clad temperature for unmitigated high-density small leak (OCP3; 1x4) ...210

Figure 117 Midplane clad temperature for unmitigated high-density small leak 

 (OCP3; 1x4)..........................................................................................................211

Figure 118 Ring 1 clad temperature for unmitigated high-density small leak (OCP2; 1x8) ...211

Figure 119 Ring 1 clad temperature for unmitigated high-density small leak (OCP2; 1x4) ...212

Figure 120 Environmental release fractions for unmitigated high-density small leak 

 (OCP2; 1x8)........................................................................................................212

Figure 121 Environmental release fractions for unmitigated high-density small leak 

 (OCP2; 1x4).........................................................................................................213

Figure 122 Layout of assemblies for OCP1 high-density (uniform) model ...........................214

Figure 123 Layout of assemblies for OCP2 high-density (uniform) model ............................214

Figure 124 Ring 1 clad temperature for unmitigated uniform high-density moderate leak 

 (OCP1)................................................................................................................215

Figure 125 Cesium environmental release fraction for unmitigated uniform high-density 

 moderate leak (OCP1)........................................................................................216

Figure 126 Ring 1 clad temperature for mitigated uniform high-density moderate leak 

 (OCP2) with 500 gpm injection ............................................................................217

Figure 127 Ring 1 clad temperature for mitigated uniform high-density moderate leak 

 (OCP2) with 200 gpm spray.................................................................................217

Figure 128 Ring 1 clad temperature for mitigated uniform high density moderate leak 

 (OCP2) with 500 gpm spray.................................................................................218

Figure 129 Ring 1 clad temperature for unmitigated high-density moderate leak (OCP3) ....224

Figure 130 Ring 1 clad temperature for unmitigated high-density moderate leak 

 (OCP3; early reactor building failure)...................................................................225

Figure 131 Ring 1 clad temperature for unmitigated high-density moderate leak (OCP3; 

 late reactor building failure)..................................................................................225

Figure 132 Ring 1 clad temperature for unmitigated high-density moderate leak (OCP4) ....226

Figure 133 Ring 1 clad temperature for unmitigated high-density moderate leak (OCP4; 

 late reactor building failure)..................................................................................226

Figure 134 SFP concrete floor temperature for unmitigated high-density moderate leak 

 (OCP3; late reactor building failure).....................................................................228

Figure 135 Cesium release fraction for unmitigated high-density moderate leak (OCP3; 

 late reactor building failure) with and without MCCI.............................................228

Figure 136 Land Contamination Sensitivity to Dose Criteria .................................................232

Figure 137 Atmospheric release fractions for unmitigated high density small leak 

 (OCP3) with a 96 hour time truncation.................................................................233

Figure 138 Affecting Likelihood of SFP Release from a Severe Seismic Event .....................255

Figure 139 Comparison of Population-Weighted Average Individual Latent Cancer Fatality 

 Risk Results for this Study to the NRC Safety Goal (plotted on 

 logarithmic scale).................................................................................................256

Figure 140 The Reference Plant’s Wind Rose........................................................................ A-3

Figure 141 Graphical representation of spent fuel risks ......................................................... B-4

Figure 142 Dose Rate in Vicinity of Fukushima Daiichi Nuclear Plant Site Main Gate 

 between March 11 and March 16, 2011 .............................................................D-18

Figure 143 Fukushima Daiichi Site Dose Rates between March 22 and March 23, 2011.... D-19
xix

LIST OF TABLES

Table 1 Frequency of SFP Fuel Uncovery (/yr) .......................................................................9

Table 2 Comparison of Fuel Age and Heat Load against Industry Averages ........................17

Table 3 Major Assumptions ...................................................................................................19

Table 4 Seismic Bins and Initiating Event Frequencies .........................................................34

Table 5 Comparison of Seismic Frequencies from Various Sources.....................................35

Table 6 Estimated Natural Frequencies of Vibration for the Peach Bottom Reactor 

 Building (Lambright et al., 1990)...............................................................................44

Table 7 Material Properties for the Nonlinear Finite Element Analyses.................................53

Table 8 Approximate Dead Loads on the SFP Floor in Terms of an Equivalent 

 Vertical Floor Pressure .............................................................................................54

Table 9 Approximate Peak Equivalent Seismic Loads in Terms of an Equivalent 

 Static Vertical Floor Pressure ...................................................................................54

Table 10 Fukushima Daiichi, Measured and Design (DBGM SS) PGAs at Foundation 

 Slab...........................................................................................................................70

Table 11 Onagawa, Measured and Design (DBGM SS) PGAs at Foundation Slab 

 (Tohoku, 2011 Earthquake) ......................................................................................70

Table 12 Fukushima Daiini, Measured and Design (DBGM SS) PGAs at Foundation Slab 

 (Tohoku, 2011 Earthquake) .......................................................................................70

Table 13 Tokai, Measured and Design (DBGM SS) PGAs at Foundation Slab 

 (Tohoku, 2011 Earthquake) ......................................................................................71

Table 14 Kashiwazaki-Kariwa, Measured and Design PGAs at Foundation Slab 

 (Chuetsu-Oki, 2007 Earthquake) ..............................................................................71

Table 15 Remaining Boundary and Initial Conditions ..............................................................77

Table 16 OCP Definition for the Modeled Operating Cycle .....................................................79

Table 17 Summary of Mitigation Assumptions.........................................................................83

Table 18 Scenario Breakdown per OCP..................................................................................91

Table 19 Refresher on the Seismic Hazard Estimates ............................................................91

Table 20 Refresher on ac Fragility...........................................................................................92

Table 21 Refresher on SFP Leakage Conditional Probabilities...............................................92

Table 22 Refresher on the OCP Fractional Contributions .......................................................92

Table 23 MELCOR Radionuclide Class Composition............................................................102

Table 24 Radionuclides Gap Inventories ...............................................................................103

Table 25 Distribution of Decay Heat in the Reactor and SFP for High Density Loading .......115

Table 26 Distribution of Decay Heat in the Reactor and SFP for Low Density Loading ........118

Table 27 Summary of Release Characteristics for High-Density Scenarios..........................151

Table 28 Summary of Release Characteristics for Low-Density Scenarios...........................152

Table 29 Release Category Types ........................................................................................154

Table 30 Release Category Tally...........................................................................................154

Table 31 Listing of Scenario-specific Release Sequences ....................................................155

Table 32 Summary of Evacuation Models .............................................................................160

Table 33 Overall Consequence Results ................................................................................165

Table 34 Dose-Response Model Results (LNT) and Dose Truncation Comparison..............168

Table 35 Average Land Interdiction* (square miles per event)..............................................171

Table 36 Average Number of Long-term Displaced Individuals* (per event).........................172

Table 37 Consequence1 Comparison – High (1x4) Density / Low Density Loading without 

 Successful 50.54(hh)(2) Mitigation .........................................................................173
xx

Table 38 Consequence Comparison – Unsuccessful/Successful Deployment of 

 50.54(hh)(2) Equipment ..........................................................................................174

Table 39 The scope and assumptions of the HRA study.......................................................178

Table 40 The SFPS Simulation Results.................................................................................179

Table 41 Approximate Time of Fuel Uncovery.......................................................................181

Table 42 The available time* .................................................................................................183

Table 43 Estimates of the Time Required for the Operator to Deploy SFP Makeup If Fire 

 Water Is Available ...................................................................................................193

Table 44 Estimates of the Time Required for the Operator to Deploy SFP Makeup If Fire 

 Water Is Not Available or If It Cannot Deliver Sufficient Flow .................................193

Table 45 Estimates of time required and time available for mitigation...................................193

Table 46 Time margin effects on human error probability in the SPAR-H HRA method for 

 cognitive activities in low power /shutdown operations...........................................194

Table 47 Human error probability estimates of a single unit event ........................................197

Table 48 The likelihood of gap release*.................................................................................198

Table 49 Scenario Specific Human Error Probability Estimates*...........................................200

Table 50 Summary of Release Characteristics for High-Density, Uniform Pattern................218

Table 51 Listing of Uniform Pattern Release Sequences ......................................................219

Table 52 Uniform Pattern Consequence Results...................................................................220

Table 53 Dose Truncation Comparison for Uniform Pattern..................................................221

Table 54 Consequence Comparison – High Density (1x4 and Uniform) Loading Without 

 Successful 50.54(hh)(2) Mitigation .........................................................................222

Table 55 Consequence Comparison – High (Uniform) Density / Low Density 

 Loading Without Successful 50.54(hh)(2) Mitigation ..............................................223

Table 56 Comparison of Low-Density OCP1 Release Fractions for a Concurrent 

 Reactor Accident.....................................................................................................227

Table 57 Comparison of Release Fractions with and without MCCI......................................229

Table 58 Consequence Comparison – Molten Core Concrete Interaction ............................229

Table 59 Low-density OCP2 Release Fraction Sensitivity to Ring-Ring Radiation................230

Table 60 Consequence Comparison – Land Contamination Sensitivity ................................231

Table 61 Consequence Comparison – Time Truncation Sensitivity ......................................233

Table 62 Comparison of consequence results from current and previous spent fuel pool 

 analyses..................................................................................................................237

Table 63 Comparison of Source Terms from Current and Previous SFP analyses...............238

Table 64 Modeled Inventory for Supplemental Reanalysis....................................................240

Table 65 Particle Size Information .........................................................................................241

Table 66 Parameters and Results from NUREG-1864, Table E.3.........................................243

Table 67 Supplemental Reanalysis with SFPS Input Deck ...................................................243

Table 68 Comparison of consequence results from SFPS, NUREG-1864, and DCSS 

 supplemental analyses ...........................................................................................244

Table 69 Evacuation Model 1: EPZ Evacuation..................................................................... A-2

Table 70 Evacuation Model 2: Evacuation for PAGs exceeded beyond the EPZ 

 (SFP release after 40 hours.).................................................................................. A-5

Table 71 Evacuation Model 3: Evacuation for PAGs exceeded beyond the EPZ 

 (SFP release after 8 hours)..................................................................................... A-7

Table 72 Seismic Bins and Initiating Event Frequencies .....................................................D-11

Table 73 Frequency of Spent Fuel Pool Fuel Uncovery for Seismic Events .......................D-12

Table 74 Release Frequencies for Spent Fuel Pool Initiators..............................................D-14

Table 75 Spent Fuel Pool Release Frequency Estimates ...................................................D-14

Table 76 Release Frequency Comparison Between Inclusion of All Initiator Event 

 Classes to the Seismic Bin No. 3 Event................................................................D-15
xxi

Table 77 Consumer Price Index – All Urban Consumers Inflator ........................................D-15

Table 78 Onsite Property Decontamination, Repair, and Refurbishment Costs..................D-17

Table 79 Average Accident Occupational Exposure at Fukushima Dai-ichi Nuclear 

 Power Plant from March to May 2011...................................................................D-19

Table 80 Estimated Immediate Accident Occupational Monthly Exposure at Fukushima ... D-20

Table 81 Immediate Accident Occupational Exposure for a Spent Fuel Pool Fire ..............D-20

Table 82 Long-Term Accident Occupational Exposure for a Spent Fuel Pool Fire .............D-21

Table 83 Long-Term Habitability Criterion ...........................................................................D-22

Table 84 Habitability Criterion Scaling Factors ....................................................................D-22

Table 85 Economic Consequences Scaling Factors as a Function of Habitability Criteria.. D-23

Table 86 Number of Spent Fuel Assemblies Remaining through Operating License 

 Expiration ..............................................................................................................D-24

Table 87 Representative Sampling of Commercially Available BWR Spent Fuel Dry 

 Storage Technology..............................................................................................D-24

Table 88 Canister Storage Capacity Based on Heat Rate Limitations ................................D-25

Table 89 Amortized DSC Upfront Costs ..............................................................................D-26

Table 90 Incremental Unit Cost Estimates...........................................................................D-26

Table 91 Incremental Occupational Dose (Routine) Estimates ...........................................D-27

Table 92 Regulatory Baseline Loading of Dry Storage Casks.............................................D-28

Table 93 Low-density Spent Fuel Pool Case Loading of Dry Storage Casks......................D-29

Table 94 Summary of Public Health (Accident) for Low-density Spent Fuel Pool 

 Storage [All Initiators]............................................................................................D-31

Table 95 Sensitivity Analyses of Public Health (Accident) Benefits for Low-density 

 Spent Fuel Pool Storage for All Initiating Events (within 50 miles) .......................D-32

Table 96 Sensitivity Analyses of Public Health (Accident) Benefits for Low-density 

 Spent Fuel Pool Storage for All Initiating Events (extending beyond 50 miles) ... D-32

Table 97 Occupational Health (Accident) Benefits for Low-density Spent Fuel Pool 

 Storage Considering All Initiating Events..............................................................D-32

Table 98 Occupational Health (Routine) Costs for Low-density Spent Fuel Pool 

 Storage .................................................................................................................D-33

Table 99 Offsite Property Cost Offsets for Low-density Spent Fuel Pool Storage..............D-34

Table 100 Summary of Onsite Property Cost Offsets for Low-density Spent Fuel Pool 

 Storage ................................................................................................................D-35

Table 101 Industry Implementation Cost Model for Low-density Spent Fuel Pool Storage.. D-36

Table 102 Industry Implementation Costs for Low-density Spent Fuel Pool Storage........... D-36

Table 103 Summary of Net Benefits for Low-density Spent Fuel Pool Storage 

 Considering All Initiator Events (within 50 miles).................................................D-40

Table 104 Summary of Total Implementation and Operation Costs for Low-density 

 Spent Fuel Pool Storage for All Initiator Events...................................................D-41

Table 105 Summary of Total Benefits and Cost Offsets for Low-Density Spent Fuel Pool 

 Storage for All Initiator Events .............................................................................D-41

Table 106 Dollar Per Person-Rem Sensitivity Analysis of Net Benefits for Low-density 

 Spent Fuel Pool Storage Considering All Initiating Events (within 50 miles) ....... D-42

Table 107 Consequences Extending Beyond 50 Miles Sensitivity Analysis of Net 

 Benefits for Low-density Spent Fuel Pool Storage Considering All 

 Initiating Events....................................................................................................D-43

Table 108 Combined Sensitivity Analysis that Analyzes Consequences Beyond 50 

 Miles using a Revised Dollar per Person-Rem Conversion Factor on the Net 

 Benefits for Low-density Spent Fuel Pool Storage for All Initiator Events............ D-43

Table 109 Public Health (Accident) Person-Rem Averted ....................................................D-45

Table 110 Facility Employee Exposure.................................................................................D-45
xxii

Table 111 Installation and Continuing Costs Associated with the Backfit.............................D-46

Table 112 Summary of Backfitting Net Benefits for Low-density Spent Fuel Pool 

 Storage for All Initiator Events (within 50 miles) ..................................................D-47

Table 113 Summary of Cost Offsets for Onsite and Offsite Property ...................................D-47

Table 114 Combined Sensitivity Analysis of the Backfitting Net Benefits for 

 Low-density Spent Fuel Pool Storage for All Initiator Events (extending 

 analysis beyond 50 miles and using a Revised Dollar per Person-Rem 

 Conversion Factor) ..............................................................................................D-48

Table 115 Summary of Combined Sensitivity Analysis Cost Offsets for Onsite and 

 Offsite Property ....................................................................................................D-48
xxiii

EXECUTIVE SUMMARY

The U.S. Nuclear Regulatory Commission performed this consequence study to continue its 

examination of the risks and consequences of postulated spent fuel pool accidents. Pertinent 

research conducted over the last several decades is summarized in NUREG-1353, “Regulatory 

Analysis for the Resolution of Generic Issue 82, Beyond Design Basis Accidents in Spent Fuel 

Pools, April 1989; in NUREG/CR-6451, “A Safety and Regulatory Assessment of Generic BWR 

[boiling water reactor] and PWR [pressurized water reactor] Permanently Shutdown Nuclear 

Power Plants,” April 1997 and in NUREG-1738, “Technical Study of Spent Fuel Pool Accident 

Risk at Decommissioning Nuclear Power Plants,” February 2001. The purpose of this 

consequence study was to determine if accelerated transfer of older, colder spent fuel from the 

spent fuel pool at a reference plant to dry cask storage significantly reduces risks to public 

health and safety. The specific reference plant used for this study is a GE Type 4 BWR with a 

Mark I containment. 

The study’s results will help inform a broader regulatory analysis of the spent fuel pools at U.S. 

nuclear reactors as part of the Japan Lessons-learned Tier 3 plan. This study’s results are

consistent with earlier research studies’ conclusions that spent fuel pools are robust structures 

that are likely to withstand severe earthquakes without leaking cooling water and potentially 

uncovering the spent fuel. The study shows the likelihood of a radiological release from the 

spent fuel after the analyzed severe earthquake at the reference plant to be about one time in 

10 million years or lower. In addition, the regulatory analysis included with this study does not 

support accelerated spent fuel transfer to casks for the reference plant. .

This study presents detailed analyses using state-of-the-art, validated, deterministic methods 

and assumptions, as well as probabilistic insights where practical. Previous studies have shown 

that earthquakes present the dominant risk for spent fuel pools, so this analysis considered a 

severe earthquake with ground motion stronger than the maximum earthquake reasonably 

expected to occur for the reference plant. The NRC expects that the ground motion used in this 

study is more challenging for the spent fuel pool structure than that experienced at the 

Fukushima Daiichi nuclear power plant from the earthquake that occurred off the coast of Japan 

on March 11, 2011. That earthquake did not result in any spent fuel pool leaks. This beyonddesign-basis earthquake severity was selected to challenge the spent fuel pool integrity. The 

study considered two spent fuel configurations:

• A relatively full pool where the hottest spent fuel assemblies are surrounded by four cooler

fuel assemblies in a 1×4 pattern throughout the pool (referred to as the high-density loading 

scenario), and;

• A minimally loaded pool where all spent fuel with at least 5 years of pool cooling has been 

removed so the hottest fuel assemblies are surrounded by additional water (referred to as 

the low-density loading scenario).

Limited sensitivity analyses of a 1x8 spent fuel configuration and a uniform configuration were 

also performed to better understand the potential effect of spent fuel configurations on the 

results.

Additionally, the study evaluated the potential benefits of strategies required in Title 10, Code of 

Federal Regulations (10 CFR), Part 50.54 (hh)(2) following the September 11, 2001, attacks. 
xxiv

These “mitigation measures” are intended to maintain spent fuel pool cooling in the event of a 

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

The study evaluated 10 CFR 50.54(hh)(2) mitigation measures by analyzing each scenario 

twice – with and without credit for mitigation. The study shows that successful mitigation 

reduces the likelihood of a release. The likelihood of a spent fuel pool release was equally low 

for both high- and low-density fuel loading. This is because high- and low-density fuel loading 

contains the same amount of new, hotter spent fuel recently moved from the reactor to the 

spent fuel pool. In the unlikely event of an earthquake-induced spent fuel pool leak, the 

likelihood of fuel heatup leading to a release was more strongly affected by the fuel loading 

pattern rather than the total amount of fuel in the pool. In other words, the use of favorable fuel 

patterns such as the 1x4 pattern promotes natural circulation air coolability and reduces the 

likelihood of a release from a completely drained pool. Analysis also shows that for the 

scenarios and spent fuel pool studied, spent fuel is only susceptible to a radiological release 

within a few months after the fuel is moved from the reactor into the spent fuel pool. After that 

time, the spent fuel is coolable by air for at least 72 hours.

The study considered scenarios where some preplanned and improvised mitigative actions 

were either not successful or not implemented before three days, at which time the analysis was 

terminated. In addition to the 10 CFR 50.54(hh)(2) mitigation measures, the site emergency 

response organization would request support from the offsite response organizations to 

implement improvised additional mitigative measures, such as pumping water into the spent fuel 

pool using a fire truck. Analysis of these additional mitigative measures was beyond the scope 

of this study. Additionally, this study does not consider the post-Fukushima mitigation required 

by NRC in Orders EA-12-051 and EA-12-049 and currently being implemented by all operating

U.S. nuclear power plants which should serve to further reduce spent fuel pool accident risk by 

increasing the capability of nuclear power plants to mitigate beyond-design-basis external 

events. 

Figure ES-1 illustrates the study results in terms of the likelihood of a leak and magnitude of 

release from the spent fuel pool (SFP) for the severe, low likelihood earthquake considered in 

this study.
xxv

Figure ES-1: Likelihood of a leak and magnitude of releases from beyond design basis earthquake

This study considered a severe earthquake expected to occur once in 60,000 years; the pool is 

expected to remain intact during more likely, less severe earthquakes. The structural analysis 

of the pool shows the spent fuel pool in this study has a 90% probability of surviving the severe 

earthquake with no liner leakage (or conversely, a 10% probability of damaging the liner at the 

SFP wall/floor junction such that leakage will occur). The specific conditions for liner failure vary 

according to site conditions and spent fuel pool design. NUREG-1353 predicted the likelihood 

of liner failure from all potential earthquakes to be between about two and six times in a million 

years. NUREG-1738 predicted the likelihood of liner failure from all potential earthquakes to be 

between about two times in a million years and two times in 10 million years. This study 

considered an earthquake with ground motion roughly four to eight times stronger than that 

used in the plant design and predicted a liner failure likelihood of about two times in a million 

years. 

The study examined how an accident is expected to proceed if the pool liner is damaged, 

concluding that pool leaks are somewhat less likely to release radioactive material to the 

environment than in previous studies. Depending on the size of the pool liner leak, releases 

could start anywhere from eight hours to several days after the leak starts assuming 10 CFR 

50.54(hh)(2) mitigation measures are unsuccessful. In the event of an earthquake, releases are 

considered very unlikely for several reasons:
xxvi

• The study finds liner damage is the only way to cause a radiological release in less than 3 

days for the scenarios and spent fuel pool studied. Other possible outcomes provide time to 

prevent a release by taking emergency actions. This is consistent with earlier studies. 

• The time period of susceptibility for a release of radioactive materials during the operating 

cycle is short. This study’s detailed accident progression modeling differs from earlier work 

in showing that for the severe earthquake analyzed, draining the pool after liner failure is 

less likely to lead to a release. Because spent fuel can be effectively cooled by water, 

steam, or air, the likelihood of fuel overheating to the point of radiological release depends 

on several factors: how much residual heat the fuel generates, the fuel loading pattern, and 

the timing, location, and size of the liner leakage. If 10 CFR 50.54(hh)(2) mitigation 

measures aren’t successful, releases could occur the first few months after the fuel came 

out of the reactor (or 8% of the reactor’s two-year operating cycle). If 10 CFR 50.54(hh)(2) 

mitigation measures are successful, releases could only occur the first several days after the 

fuel came out of the reactor (a factor of twenty reduction in the likelihood of release).

In the unlikely event an earthquake induced liner failure does occur, this study predicts the 

largest releases would come from high-density loading cases without 10 CFR 50.54(hh)(2) 

mitigation measures. However, for each high-density loading release case, the corresponding 

low-density loading case also resulted in a release. The low-density cases generally resulted in 

a smaller release due to the smaller inventory of radioactive materials and the lower potential for 

hydrogen combustion. For the high-density cases, the releases are limited to a few percent of 

the cesium inventory, except for a few cases that predicted hydrogen combustion and resulted 

in releases of one to two orders of magnitude higher than the other cases. In these cases, the 

spent fuel heats up in a steam environment leading to oxidation of zirconium and releasing 

hydrogen gas into the reactor building. The mixing and reaction of hydrogen and oxygen leads 

to a hydrogen combustion and substantially damages the reactor building. That damage could 

breach structures that would retain radioactive material, along with allowing more oxygen into 

the building, potentially increasing the severity of the spent fuel fire. The study included a 

sensitivity analysis for a 1x8 loading pattern (hotter fuel surrounded by 8 cooler assemblies in a 

repeating pattern) which also resulted in smaller radioactive releases because the hotter 

assembly transfers its heat to the cooler assemblies resulting in lower peak fuel temperatures

Following the evaluation of successful and unsuccessful mitigation cases, a limited-scope 

human reliability analysis was performed to estimate the likelihood of successful operator 

actions implementing 10 CFR 50.54(hh)(2) mitigation measures to prevent fuel 

damage. Assumptions included post-earthquake on-site portable mitigation equipment 

required by 10 CFR 50.54(hh)(2) is available, minimum plant staffing are available for 

implementing spent fuel pool mitigation, and the work area is accessible to perform 

mitigation. The structural and accident progression analyses show that at least 99% of the time, 

the earthquake would not result in spent fuel overheating even without mitigative actions for the 

first seven days following the accident. For the remaining times, mitigative actions are needed 

to prevent fuel damage and the calculated mitigation success rates range from about 25% to 

95% depending on plant conditions and assuming that the refueling floor is accessible. There

are two exceptions where mitigation will be ineffective under the moderate leak scenarios: (1) if 

the earthquake occurs at the beginning of a refueling outage when the spent fuel is too hot for 

the assumed mitigation, and (2) if the earthquake occurs when spent fuel is relatively hot and 

the reactor and spent fuel pool are hydraulically disconnected resulting in insufficient time to 

deploy mitigation and natural cooling mechanisms cannot prevent fuel damage.
xxvii

The study’s analyses shows that a release from a spent fuel pool accident after the severe 

earthquake at the reference plant could occur about one time in 10 million years or lower. The 

factors leading to this low likelihood, as discussed above, are summarized in Figure ES-2.

Figure ES-2: Factors Affecting Likelihood of SFP Release from a Severe Seismic Event

The study then estimated consequences to the public of a low likelihood spent fuel pool 

accident release. The releases of radioactive material are generally comparable to past studies. 

Despite the fairly large releases for certain predicted accident progressions, consequence 

analysis of all scenarios indicated zero early fatalities from acute radiation effects because 

protective actions were modeled to be effective in limiting doses to the public. The study also 

showed that the risk of an individual dying from cancer from the radioactive release is very low. 

When including the very low likelihood of a release, the risk in the analyzed scenarios that an 

average individual within 10 miles receives a fatal latent cancer is between about two in a trillion 

and five in a hundred billion per year. The risks are similar between different loading or 

mitigation scenarios because of modeled offsite protective actions that include evacuation, 

sheltering, relocation, and decontamination. Additionally, these individual risks are dominated by 

long-term exposures to very lightly contaminated areas for which doses are small enough for 

the areas to be considered habitable.

In order to do a regulatory analysis to inform whether low density loading should be required at 

the reference plant, cost estimates of potential protective measures are considered along with 

other parameters in a cost-benefit analysis. The study shows that, while public health effects 

from these low likelihood spent fuel pool releases are expected to be very low for all the 

scenarios studied, offsite protective measures in the form of population relocation and land 

interdiction may be extensive. High-density loading releases without 10 CFR 50.54(hh)(2) 

mitigation measures are calculated to result in release frequency-weighted land interdiction 

values of 0.001 mi2 per year and 0.5 displaced individuals per year which are arrived at by 
xxviii

multiplying the estimated frequency and the estimated consequence. While the amount of land 

interdiction can be large, the fraction expected to be permanently interdicted is small if a release 

were to occur. For low-density loading or with successful deployment of 10 CFR 50.54(hh)(2) 

mitigation measures, considerably less land interdiction and displaced individuals are predicted. 

Comparisons of the calculated individual latent cancer fatality (LCF) risk within 10 miles to the 

NRC Safety Goal are provided in Figure ES-3 to give context that may help the reader to 

understand the contribution to cancer risks from the accident scenarios that were studied. The 

NRC Safety Goal for latent cancer fatality risk from nuclear power plant operation (i.e., 2x10-6 or 

two in one million per year) is set 1,000 times lower than the sum of cancer fatality risks 

resulting from all other causes (i.e., ~2x10-3 or two in one thousand per year).

Figure ES-3: Comparison of Population-Weighted Average Individual Latent Cancer Fatality Risk Results for 

this Study to the NRC Safety Goal (plotted on logarithmic scale) 

Comparing the study results to the NRC Safety Goal does involve important limitations. First, 

the safety goal is intended to encompass all accident scenarios on a nuclear power plant site, 

including both reactors and spent fuel. This study does not examine all scenarios that would 

need to be considered in a probabilistic risk assessment for a spent fuel pool, although seismic 

contributors are considered the most important contributors to spent fuel pool risk. Also, this 

study represents a mix of limited probabilistic considerations with a deterministic treatment of 

mitigating features. All analytical techniques, both deterministic and probabilistic, have inherent 

limitations of scope and method and also have uncertainty of varying degrees and types. As a 

result, comparison of the scenario-specific calculated individual LCF risk to the NRC Safety 

Goal is incomplete. However, it is intended to show how multiple spent fuel pool scenarios’ risk 

results in the one in a trillion (10-12) to one in 10 billion (10-10) per year LCF range) are low. While 
xxix

the results of this study are scenario-specific and related to a single spent fuel pool, staff 

concludes that since these risks are several orders of magnitude smaller than the 2x10-6 (two in 

one million) individual LCF risk that corresponds to the safety goal for latent cancer fatalities, it 

is unlikely that the results here would contribute significantly to a risk that would challenge the 

Commission’s safety goal policy (NRC, 1986).

In conclusion, past SFP risk studies have shown that high-density spent fuel storage is safe and 

risk of a release due to an accident is low. This study is consistent with earlier research 

conclusions that spent fuel pools are robust structures that are likely to withstand severe 

earthquakes without leaking. This study estimated that the likelihood of a radiological release 

from the spent fuel pool resulting from the selected severe seismic event analyzed in this study 

is on the order of one time in 10 million years or lower. For the hypothetical releases studied, 

no early fatalities attributable to radiation exposure were predicted and individual latent cancer 

fatality risks are projected to be low, but extensive protective actions may be needed.

The study results demonstrated that in a high-density loading configuration, dispersing hotter 

fuel throughout the pool or successful mitigation generally prevented or reduced the size of 

potential releases. Low-density loading reduced the size of potential releases but did not affect 

the likelihood of a release. When a release is predicted to occur, early and latent fatality risks 

for individual members of the public do not vary significantly between the scenarios studied 

because protective actions, including relocation of the public and land interdiction, were 

modeled to be effective in limiting exposure. The beneficial effects in the reduction of offsite 

consequences between a high-density loading scenario and a low-density loading scenario are 

primarily associated with the reduction in the potential extent of land contamination and 

associated protective actions. The regulatory analysis for this study indicates that expediting

movement of spent fuel from the pool does not provide a substantial safety enhancement for the 

reference plant. The NRC plans to use the insights from this analysis to help inform a broader 

regulatory analysis of the spent fuel pools at U.S. nuclear reactors as part of its Japan Lessonslearned Tier 3 plan. The NRC continues to believe, based on this study and previous studies

that high density storage of spent fuel in pools protects public health and safety.
xxx
xxxi

ACKNOWLEDGMENTS

The authors would like to acknowledge the following individuals who directly or indirectly 

contributed to this work. The following individuals directly contributed in the form of analyses 

and support:

Jon Ake, U.S. Nuclear Regulatory Commission

Jeffrey Mitman, U.S. Nuclear Regulatory Commission

Antonios Zoulis, U.S. Nuclear Regulatory Commission

Christopher Cahill, U.S. Nuclear Regulatory Commission

Nate Bixler, Sandia National Laboratories

Hernando Candra, U.S. Nuclear Regulatory Commission

Ian Gauld, Oak Ridge National Laboratory

Andrew Goldmann, Sandia National Laboratories

Larry Humphries, Sandia National Laboratories

Joe Jones, Sandia National Laboratories

Thomas Miler, Oak Ridge National Laboratory

Bruce Patton, Oak Ridge National Laboratory

Joel Piper, Rotatee from the U.S. Department of Homeland Security

Randy Sullivan, U.S. Nuclear Regulatory Commission

Kenneth C. Wagner, Dycoda, LLC

The following individuals from other offices of the U.S. Nuclear Regulatory Commission (NRC)

provided support and advice throughout the conduct of the study:

Eric Bowman

Rick Ennis

Sam Hansell

John Hughey

Steven Jones

Eric Powell

Wayne Schmidt

Eric Schrader

Bret Tegeler

Kent Wood

Adam Ziedonis

Finally, the authors wish to acknowledge the guidance and support of NRC management, and in 

particular Brian Sheron, Director of the Office of Nuclear Regulatory Research, who provided 

the initial direction for the study.
xxxii
xxxiii

ABBREVIATIONS AND ACRONYMS

ac alternating current

ACRS Advisory Committee on Reactor Safeguards

AEF annual exceedance frequency

ANL Argonne National Laboratory

BEIR biological effects of ionizing radiation

BEF biological effectiveness factor

Bq Becquerel

BWR boiling-water reactor

C Celsius

CEC Commission of the European Communities

CFD computational fluid dynamics

CFR Code of Federal Regulations

Ci curies

Cs cesium

CSCM continuous surface cap model

CV control volume

DBE design basis earthquake

dc direct current

DDREF dose and dose rate effectiveness factor

DF decontamination factor

DLTEVA delay to evacuation

DLTSHL delay to shelter

DOE U.S. Department of Energy

DURBEG duration of beginning phase

DURMID duration of middle phase

E East

EAL emergency action levels

EAS emergency alert system

EPA U.S. Environmental Protection Agency

EPRI Electric Power Research Institute

EPZ emergency planning zone

ESPEED speed (WinMACCS input variable)

ETE evacuation time estimate

FAQ frequently asked questions

FEMA Federal Emergency Management Agency

FGR federal guidance report

FSAR final safety analysis report

GE General Electric

GEIS generic environmental impact statement

GI generic issue

GNF Global Nuclear Fuel

gpm gallons per minute

GSI Generic Safety Issue

GWD gigawatt-day

HCLPF high confidence of low probability of failure

HEP human error probability

hr hour

HPS Health Physics Society

HRA human reliability analysis
xxxiv

I iodine

ICE inadvertent criticality event

ICRP International Commission on Radiological Protection

INL Idaho National Laboratory

IPEEE individual plant evaluation for external events

ISFSI independent spent fuel storage installation

ISRS in-structure response spectra

K Kelvin

KI potassium iodide

LCF latent cancer fatality

LLNL Lawrence Livermore National Laboratories

LNT linear no-threshold

LOOP loss of offsite power

MACCS2 MELCOR Accident Consequence Code System, Version 2

MCCI molten core-concrete interaction

MCi megacuries

MPC multi-purpose container

MTU metric tons of uranium

MW megawatts

MWD megawatt days

N North

NCRP National Council on Radiation Protection and Measurements

NAS National Academy of Sciences

NRC Nuclear Regulatory Commission

OCP operating cycle phase

ORNL Oak Ridge National Laboratory

ORO offsite response organization

OSC operational support center

PAG protective action guides

PBAPS Peach Bottom Atomic Power Station

PGA peak ground acceleration

PPG pool performance guidelines

PWR pressurized water reactor

PRA probabilistic risk assessment

QHO quantitative health objectives

RB reactor building

REM Roentgen Equivalent Man

RHR residual heat removal

S South

SAE site area emergency

SBO station blackout

SIP shelter in place

SOARCA State of the Art Reactor Consequence Analyses

SNL Sandia National Laboratories

SFP spent fuel pool

SFPS Spent Fuel Pool Study

SSC structures, systems, and components

SSE safe shutdown earthquake

TSC technical support center

TSG technical support guideline

TR technical report (EPRI technical reports)
xxxv

USGS United States Geological Survey

W West

1

1. INTRODUCTION AND BACKGROUND

All operating commercial nuclear reactors in the United States are of the light-water reactor 

design. They utilize upright fuel assemblies (roughly 12 feet in length) with low-enriched 

uranium oxide fuel (less than 5-percent uranium-235). The fuel assemblies which are 

composed of numerous fuel rods (typically 80-100 rods for boiling-water reactor fuel and 200–

300 rods for pressurized-water reactor fuel) are placed in the reactor for two to three operating

cycles. Each operating cycle typically lasts 18 to 24 months. At the end of their “life,” the 

assemblies are placed in large pools of water near the reactor that are roughly 12 meters (m)

(40 feet (ft)) deep. For facilities licensed to operate an independent spent fuel storage 

installation (ISFSI), the fuel assemblies are later loaded into casks and moved to the ISFSI as 

necessary to accommodate future core offloads. The casks are drained of water and inerted 

with helium during the loading process. This situation leads to the vernacular terms of “wet 

storage” (to describe storage in the spent fuel pool (SFP)) and “dry storage” (to describe storage 

in casks).

SFPs in the United States were originally designed to store one to two reactor cores worth of 

spent fuel, so that the fuel could “cool down” (become less thermally and radioactively “hot”) 

before its movement to a reprocessing facility or permanent geological repository. Owing to the 

abandonment of spent fuel reprocessing as well as delays in the identification, licensing and 

construction of a repository, U.S. nuclear power plants “re-racked” their SFPs in the 1980s and 

1990s to allow for the storage of larger numbers of spent nuclear fuel assemblies (i.e., roughly 

four reactor cores worth for the plant studied in this study). Throughout this time (including 

present day), the U.S. Nuclear Regulatory Commission (NRC) has maintained that SFPs 

provide adequate protection of the public health and safety in either low-density or high-density 

storage configurations. The basis for this position is discussed later in this section.

Stakeholders have periodically challenged the NRC’s position that SFPs provide adequate 

protection of public health and safety. To understand the basis for these challenges, it’s first 

necessary to understand two basic facts about spent nuclear fuel:

(1) Thermal and radioactivity loads associated with freshly discharged fuel necessitate the 

need for wet storage.

(2) All spent nuclear fuel, regardless of age (i.e., time since discharge from the reactor),

produces both heat and radiation.

The list below presents some less-obvious considerations from the perspective of the benefits

and disadvantages associated with transitioning from high-density storage to low-density 

storage. The list is subdivided into two parts—those considerations that are covered within this 

study and those that are not.

This study includes the following considerations:

• Removal of older fuel from the SFP will decrease the inventory of longer lived 

radionuclides, such as cesium-137, present in the SFP.

• Removal of older fuel will result in less radioactive material would be present in the pool 

if a radioactive release occurred, which would be expected to reduce potential offsite 

consequences.
2

• Removal of older fuel reduces the overall heat load in the pool while decreasing the 

amount of metal mass to act as a heat sink should the fuel become uncovered, which 

can have competing effects on accident timing depending on the type of accident (e.g., a 

boiloff event versus a complete draindown).

• Removal of older fuel will increase the area available for air circulation (natural 

circulation) should the pool become completely drained (the effect of this is somewhat 

limited by the nature of spent fuel racks as discussed later in this report).

• Removal of older fuel will increase the volume available for cooling water (note that this 

is mathematically a small effect with the older fuel comprising on the order of 5-percent 

of the total pool volume—because most of the pool is occupied by water, not fuel).

1

This study does not explicitly address the following considerations, though some are discussed 

further in APPENDIX B: 

• Discharging large amounts of fuel (and thus greatly increasing the amount of fuel 

contained in the ISFSI) would increase the number of casks required to store the 

existing spent fuel inventory.

• Expedited discharging of fuel from the SFP to dry storage increases the frequency of 

postulated cask drops, which in turn increases the frequency of causing damage to the 

pool or cask that could lead to a radioactive release.

• Expedited discharging of fuel increases occupational doses for workers involved with the 

management and transfer of the spent fuel.

• Earlier movement of fuel into casks that are not currently approved for shipping or longterm storage may require that fuel to be repackaged later for shipment to the eventual 

long-term repository or interim storage site.

Issues related to design-basis accidents and risk posed by dry cask storage have received, and 

continue to receive, attention from various stakeholders. Issues related to the existing dry cask 

storage infrastructure, worker dose, and economics are discussed in (NAC, 2011) and (EPRI, 

2012). Section 1.6 of this report provides more information on each of these studies.

The first set of bulleted considerations is generally advantages associated with expedited fuel 

movement to casks, while the latter set of bulleted considerations is generally disadvantages. 

The agency’s position—that spent fuel storage in either pools or casks is safe—is based on a 

number of past studies and regulatory activities that are discussed later in this chapter. By 

investigating the pros, we are informing ongoing discussions as to whether fuel movement from 

spent fuel pools to dry cask storage should be expedited and if any of the “pros” are more 

 1 The additional water can have a non-intuitive negative impact in certain situations. For a leak at the bottom 

of the SFP, the additional water at the elevation of the fuel causes it to take longer to “clear” the baseplate 

(i.e., for the level of the receding water to drop below the bottom of the baseplate). In situations where 

natural circulation of air under and up through the racks is effective for preventing fuel heatup, this actually 

temporarily inhibits cooling of the fuel. While this does require a specific set of conditions to be relevant, it is 

raised here because it does actually arise in one of the scenarios realized later in this report.
3

compelling than past studies suggest. If they are, then the issue can be addressed more 

holistically.

1.1 Project Impetus

Various risk studies (most recently NUREG-1738, “Technical Study of Spent Fuel Pool Accident 

Risk at Decommissioning Nuclear Power Plants,” issued February 2001) have shown that 

storage of spent fuel in a high-density configuration in SFPs is safe and that the risk is low. 

These studies used simplified and sometimes bounding assumptions and models for 

characterizing the likelihood and consequences of beyond-design-basis SFP accidents. As part 

of NRC’s security assessments after the events of September 11, 2001, SFP modeling using 

detailed thermal-hydraulic and severe accident progression models integrated into the MELCOR 

code were developed and applied to assess the realistic heatup of spent fuel under various pool 

draining conditions. Moreover, in conjunction with these post-9/11 security assessments, the 

NRC issued a new regulation, 10 CFR 50.54(hh)(2), that requires reactor licensees to develop 

and implement guidance and strategies intended, in part, to maintain or restore SFP cooling 

capabilities following certain beyond design basis events.

Recently, the agency has restated its views on the safety of spent fuel stored in high-density 

configurations in a response to Petition for Rulemaking (PRM)-51-10 and PRM-51-12 as well as 

the revision to NUREG-1437, Revision 1, “Generic Environmental Impact Statement for License 

Renewal of Nuclear Plants - Draft Report for Comment,” issued July 2009. However, this 

position relies in part on the findings of the aforementioned security assessments, which are not 

publicly available. The renewed interest in spent fuel storage engendered from the changes in 

the path forward of the planned geologic repository and from the events in Japan following the 

March 2011 earthquake has rekindled interest in capturing the consequences from postulated 

accidents associated with high-density SFP storage in an updated safety study. 

The spent fuel pool study’s (SFPS) primary objective was to determine if accelerated transfer of 

spent fuel from the spent fuel pool to dry cask storage provides a substantial safety 

enhancement for the reference plant. The insights from this analysis will inform a broader 

regulatory analysis of the SFPs at U.S. nuclear reactors as part of the Japan Lessons-learned 

Tier 3 plan. NRC analyzes low likelihood (beyond the design basis) events to estimate risk to 

the public and to explore and ensure safety margins. The results of the study will be used to 

inform the evaluation of what future regulatory actions the NRC might undertake, including 

whether expedited transfer of spent fuel from spent fuel pools into dry cask storage is justified. 

To help inform whether regulatory action needs to be taken in this area, the NRC has prepared 

an example of a regulatory analysis of the reference plant studied in this report (see APPENDIX 

D). A regulatory analysis is an analytical tool used by NRC decision-makers to help determine 

whether the NRC should implement a proposed regulatory action. The regulatory analysis is 

intended to inform NRC decision makers whether there is a substantial increase in the overall 

protection of the public health and safety, and whether the direct and indirect costs of 

implementation are justified in view of a potential substantial increase in protection. For the 

example regulatory analysis, the Spent Fuel Pool Study (SFPS) results are used as quantitative 

inputs to the safety goal screening criteria in accordance with the NRC regulatory analysis 

guidelines (NUREG/BR-0058), wherein the quantitative health objectives are used as a 

surrogate of the safety goal. The example regulatory analysis also contains estimates of 

benefits and costs, which are quantified when possible, together with a conclusion as to whether 

the proposed regulatory action is cost-beneficial. “Cost-beneficial” means that the benefits of 

the proposed action are equal to, or exceed, the costs of the proposed action. Accident 

consequences such as land interdiction and population relocation reported in this study are 
4

used to estimate the costs resulting from an accident (e.g., costs of interdiction measures, such 

as decontamination, cleanup, and evacuation) as part of the cost-benefit analysis.

Other aspects of SFP risk that have not been informed by this or past studies, may be 

addressed by future studies, such as the site Level 3 probabilistic risk assessment (PRA), as 

documented in SECY-11-0089, “Options for Proceeding with Future Level 3 Probabilistic Risk 

Assessment Activities,” dated July 7, 2011, and the associated staff requirements 

memorandum; or will be addressed through other inputs to the regulatory decision-making 

process, as needed.

1.2 Technical Approach

Two broad situations are considered in this study, which represent the following: 

(1) A condition representative of the following: (i) high-density loading in the SFP using a 

1x4 pattern (see Figure 34 for an illustration of what is meant by this terminology), (ii) a 

relatively full SFP, and (iii) current regulatory requirements relating to fuel configuration 

and preventive/mitigative capabilities; and 

(2) A condition where fuel with more than 5 years of cooling has been moved to dry cask 

storage (i.e., low-density loading in the SFP and current applicable regulatory 

requirements with respect to fuel configuration and preventive/mitigative capabilities).

For purposes of obtaining a near-term perspective on the issue, a single site and single 

assumed operating cycle are used. The site characterization (e.g., seismic response, decay 

heat, radionuclide inventory) is based on readily available information that primarily stemmed 

from sources such as the study reported in NUREG-1150, “Severe Accident Risks: An 

Assessment of Five Nuclear Power Plants,” issued December 1990; seismic information 

developed by the U.S. Geological Survey (USGS); the post-9/11 security assessments2

; and the 

State-of-the-Art Reactor Consequence Assessment (SOARCA) described in NUREG-1935. 

Later in the project, the licensee provided additional information that generally corroborates the 

assumptions made in this study. 

A BWR plant was chosen for this analysis for a mix of reasons including availability of computer 

models for a BWR plant, a perception of greater external stakeholder interest in elevated 

(relative to grade) SFPs 3

, and the fact that the nuclear reactors that felt the higher tsunami 

waves and stronger ground motions from the March 11, 2011, Tohoku earthquake, which 

includes those at Fukushima Daiichi, were all BWR reactors. In the context of a seismic event, 

the elevation of the pool will affect the transmission of seismic loads through the structure, can 

potentially inhibit accessibility for taking mitigative action, and can potentially lead to flooding of 

safety-related equipment, if the pool and surrounding structures are significantly damaged. The 

selection of a BWR design is not intended to suggest that these designs are more vulnerable to 

 2 The post 9/11-security assessments included consideration of SFPs and resulted in the collection of 

information and the development of computer models that provided a convenient starting point for the 

current study. 3 SFPs at pressurized water reactor and BWR/6s (which have Mark III containments) are generally at or near 

grade elevation, with many being partially below grade. In the Mark I and Mark II designs, the SFP is 

oriented such that the top of the SFP is at the same elevation as the top of the primary containment vessel, 

which results in them being well above grade.
5

SFP accidents. In reality, there are differences between the major design types (PWRs versus 

BWRs) that make each more or less susceptible to SFP accidents on a scenario-specific basis. 

Similarly, the selection of a site that has a separate SFP for each reactor (as opposed to a 

shared pool) is also not intended to suggest that these situations are inherently more 

vulnerable.

1.3 Site Specificity and Familiarization

This study is intentionally based on plant-specific information for a particular site, as opposed to 

attempting to define a generic site that might bound a set of plants. This approach was taken 

because it provides the best context for examining SFP accident progression and release 

phenomenology in a realistic fashion, for the purpose of providing a better understanding of the 

factors that affect the characterization of SFP beyond-design basis accidents. The decision to 

proceed in this manner was deliberately made in reaction to persistent criticisms regarding the 

realism of past studies (due to their goal of broad applicability in order to support their intended 

purposes). Because this study strives to be site-specific, it does not account for the variability in 

design and operation across the operating fleet, but rather, represents one point within that 

spectrum.

In almost all situations where plant-specific design and operational information is used, it is 

based on Unit #3 of the Peach Bottom Atomic Power Station (PBAPS), circa 

2011. Nevertheless, the SFPS makes some assumptions that are not representative of PBAPS 

because either (a) insufficient information was available at the time the modeling decision was 

made or (b) the PBAPS situation was viewed to be atypical. Regarding the former exception, 

the initial phase of the work was expedited to achieve early insights. Some modeling 

assumptions were confirmed in parallel to ongoing work, and in instances where newer 

information provided additional perspective on the modeling assumption, this is noted in the 

report. Regarding the second exception, the major example of this is that the study assumed 

the fuel is configured in a 1x4 pattern rather than in the 1x8 pattern used at PBAPS as 

discussed further in Section 5.1. In some situations, the 1x8 pattern is predicted to have a 

beneficial effect on the amount of radiation released (Section 9.2). Additionally, sensitivity 

analyses presented in Chapter 9 explore the effect of some important parameters on the study 

results. Due to these exceptions, the analysis contained in this report is best described as being 

performed for a “reference plant” which is largely based on PBAPS.

PBAPS has two General Electric (GE) Type 4 BWRs with Mark I containments, Units #2 and #3. 

This study uses Unit #3 when unit-specific information is required. Unit #1 is no longer in 

operation. Units #2 and #3 each have a dedicated SFP, and the pools do not share a common 

refueling floor, as is the case with some plants of this design. Most other aspects of the reactor, 

SFP, and reactor building are similar to BWR designs of this vintage. Two small power uprates 

have been approved for this site (1995 and 2002), with an extended power uprate submittal 

currently under review (as of January 2013).

Regarding the SFPs, the existing high-density racks were placed in service in 1986, and were 

designed and manufactured by Westinghouse Electric Corporation. As of 2010, the Unit #3 

SFP contained 2,945 assemblies, while the Unit #2 SFP contained 2,844. Both SFPs maintain 

enough open locations to allow for an emergency full core offload, if needed. The site also has 

an ISFSI for dry cask storage, utilizing the TN-68 cask design.

Finally, with respect to emergency preparedness, the site is located in a State (Pennsylvania) 

that has State-specific protective action guidelines. Detailed site-specific information relevant 
6

for this study is covered in the remainder of this report, including figures that show the reactor 

building layout, SFP layout, etc.

1.4 Basic Scenario Development

The following key aspects of the way this study is conducted should be mentioned at this point.

• A large seismic event is the only initiator considered.

• As mentioned previously, both the current situation (a high-density loading configuration 

in the pool) and an alternate situation (a low-density loading configuration in the pool) 

are analyzed. A situation in which the pool has been re-racked to a low-density rack 

configuration is not considered, because such a situation would be inefficient in terms of 

regulatory benefit given that much of the benefit of this situation could be achieved by 

storing less fuel in the existing racks (it should be mentioned that BWR fuel is 

channeled, which reduces the benefit of cross-flow if the pool were to become drained).

• The study focuses on the SFP, not the reactor, though for instances in which the two are 

hydraulically connected, both are considered to a certain extent.

• In estimating the likelihood and consequences of radiological release, the study does not 

attempt to quantify the likelihood of successful deployment of mitigation, but rather treats 

every scenario considering both the case with successful mitigation deployment and the 

case with unsuccessful mitigation deployment (also referred to as mitigated and 

unmitigated later in the report)4

. These results are then used to drive a human reliability 

analysis (Chapter 8) which provides information about what plant conditions impact 

mitigative reliability, and what range of likelihoods are expected.

• All portions of the operating cycle are considered.

• Detailed computer modeling is used to predict the plant’s response to the event, in terms 

of structural response, accident progression, mitigation effectiveness (when credited), 

and offsite consequences.

In cases in which the above represent limitations on the study’s scope or results, these are 

justified in this report. In particular, Chapter 2 of this report provides the study’s key limitations 

and assumptions.

1.5 Rationale for Focusing on Consequences of a Seismic Hazard

This section seeks to provide context regarding the suite of potential initiating events that can 

lead to an SFP accident, and why the consequences of a seismic event is the focus of this 

study. 

This study is a limited-scope consequence assessment that utilizes probabilistic insights. By 

looking at these probabilistic aspects, the results can be placed in better context, by means of 

the limited treatment of relative likelihood. While these elements provide some of the benefits of 

 4 Note that the shorthand of “mitigated” and “unmitigated” still refers to whether mitigative actions are 

successfully deployed, not whether the accident itself leads to a release.
7

an actual risk assessment, there are several elements of a risk assessment that are specifically 

not performed. These include the following:

• failure modes and effects analysis (except for SSCs specifically discussed in this 

Chapter)

• data analysis and component reliability (e.g., consideration of random failures)

• effects of dependencies

• HRA as part of the accident progression and recovery; a limited scope HRA is performed

in (Chapter 8

• system fault tree and sequence event tree development and quantification

Even so, this study does attempt to bring probabilistic insights to bear. In terms of inputs to the 

study, these include the following:

• risk information from past studies for selecting the scenarios studied

• initiating event likelihood

• initiating event timing effects (e.g., the relative likelihood of having an event during the 

various operating cycle phases and the likely configurations incurred) 

• relative likelihood of damage state characteristics and conditional probabilities 

associated with offsite consequence analysis (e.g., meteorological sampling in MACCS2 

analysis)

In terms of assessing the results, the consideration of probabilistic insights uses the above 

inputs (and simple algebraic combination) to quantify different figures of merit to put the results

in context. 

The inclusion of probabilistic aspects within the current study allows the study to consider some 

aspects of likelihood, but will not support definitive statements on risk. To elaborate, this study 

focuses on a specific portion of the overall risk profile, that of large seismic events between 0.5 

and 1g. In comparing the results of this study to those of previous studies, one can corroborate 

or challenge the continued applicability of prior estimates for this piece of the risk profile. Since 

large seismic events have been shown in the past to be a prominent contributor to risk, this 

comparison helps to predict whether a comprehensive risk assessment would be expected to 

result in an overall decrease or increase in the estimated risk. Using this approach, the results 

of this study can draw supportable, but not definitive, conclusions about overall consequences 

and risk.

For the present study, because of (1) the relative simplicity of the SFP and its supporting 

infrastructure as compared to a reactor and its supporting infrastructure and (2) the much lower 

assembly decay heats, the majority of potential SFP accident risk is believed to emanate from 

either of the following two events:

(1) events that have the potential to cause a sizable leak in the SFP
8

(2) events that might preclude operator action to cool or inject water into the pool for an 

extended period of time (i.e., days)

When one considers the various possible initiators, the first criterion points to the following:

(1) very large (i.e., well beyond the design-basis) seismic events (Note that these events 

almost certainly initiate a loss-of-offsite power and may fail emergency on-site power.)

(2) heavy load (e.g., cask) drops

(3) inadvertent aircraft crashes

In addition to these, the second criterion also points to the following:

(4) extended loss-of-offsite power (LOOP) events caused by severe weather (e.g., severe 

storms, hurricanes, tornados), within design-basis seismic events or other grid upsets, 

with concurrent loss of emergency onsite alternating current (ac) power (either because 

of the same event or because of coincidental hardware failures)

(5) lack of accessibility caused by a reactor accident that has released radioactive material 

outside of primary containment (or an accident involving the other SFP)

Note that sabotage events have been excluded from the scope of this study.

Items #(1) (seismically-induced station blackout), #(2) (cask drops), and #(4) (extended LOOPs) 

have been considered in most other SFP studies, and are discussed further below. For item 

#(3), past studies (namely NUREG-1738 (NRC, 2001)) have concluded that the risk of this 

initiator is bounded by other initiators for both PWRs and BWRs, based on quantitative 

estimates of likelihood and expected damage (see Section 3.5.2 of that study). Item #(5) 

(effects of a concurrent reactor accident) generally have not been studied in prior efforts. The 

frequency and consequences of a reactor accident is not considered and the effect of a reactor 

accident on a spent fuel pool scenario is partially considered here, but not rigorously (see 

Section 2.2 of this study for more information).

Past studies have had generally similar conclusions about the relative contribution to risk from 

the various initiating events considered. Table 1 summarizes fuel uncovery frequencies from 

NUREG-1353, “Regulatory Analysis for the Resolution of Generic Issue 82, “Beyond Design

Basis Accidents in Spent Fuel Pools,”” issued April 1989, and NUREG-1738. For both NUREG1738 and NUREG-1353, seismic events were the largest contributor to the frequency of fuel 

uncovery.
9

Table 1 Frequency of SFP Fuel Uncovery (/yr)

Initiating Event Class NUREG-1353 (1989)

(BWR, best-estimate1) NUREG-1738 (2001)

Seismic events 7x10-6 2x10-6 (LLNL)

2x10-7 (EPRI)2

Cask / heavy load drop 3x10-8 2x10-7

LOOP – severe weather - 1x10-7

LOOP – other - 3x10-8

Internal fire - 2x10-8

Loss of pool cooling 6x10-8 1x10-8

Loss of coolant inventory 1x10-8 3x10-9

Inadvertent aircraft impacts 6x10-9 3x10-9

Missiles – general 1x10-8 -

Missiles - tornado - < 1x10-9

Pneumatic seal failures 3x10-8 - 1 These numbers have not been multiplied by the stated conditional probability of having a Zirconium fire of 0.25.

2 NUREG-1738 presented results for the two different seismic hazard models in wide use at the time (the Electric 

Power Research Institute (EPRI) and Lawrence Livermore National Labs (LLNL) models).

For these reasons, a seismic event was judged to be the logical focus of this limited-scope 

consequences assessment. Based on a review of the seismic hazard for the particular site 

studied, and consideration of seismic hazard binning from contemporary seismic PRA 

methodologies, a specific range of ground motions was chosen for this study (see Chapter 3). 

This range of ground motions represents a good compromise between more likely events that 

would not be expected to lead to any consequences and less likely events that would lead to 

greater consequences (risk is the product of the likelihood times the consequences).

1.6 Operating Cycle Phase Approach

During a given operating cycle, the spent fuel pool:

• will change configuration from an isolated pool to a pool that is hydraulically connected 

to the reactor vessel (and back again)―these configurations will be referred to as poolreactor configurations to distinguish from the different spent fuel loading configurations; 

• may have spent fuel temporarily offloaded from the reactor;

• will have spent fuel permanently offloaded from the reactor;

• will likely have spent fuel moved around within the SFP (as part of complying with 

regulatory requirements related to heat distribution, criticality, and neutron absorber 

monitoring)

• may have older spent fuel offloaded into storage casks and transferred to an ISFSI;

• will experience changes in the peak assembly fission product decay power (of interest 

for draindown events and spray mitigation) because of the above considerations as well 

as radioactive decay; and
10

• will experience changes in the total decay power of all assemblies (of interest for pool 

heatup/boiling and makeup mitigation) because of the above considerations as well as 

radioactive decay.

To rigorously represent these changing conditions, the study breaks up the operating cycle into 

numerous small periods of time or operating cycle phases (OCPs). However, the number of 

OCPs considered is nearly a linear multiplier on the amount of resources needed because each 

period of time requires its own set of accident progression and consequence analyses. Past 

studies have taken the approach of selecting specific points in time of interest, and comparing 

results for those specific times. This study takes a similar approach, but places more emphasis 

on the definition of these times as quasi-steady representations of the portion of the operating 

cycle that they represent. This approach allows for more accurate representation of the 

annualized frequencies of offsite consequences. The specific selection of these phases is 

described further in Section 5.2 of this report.

1.7 Overview of Past Studies

A number of past studies have been performed to look at various aspects of spent fuel and SFP

safety, security, and risk. The major regulatory activities are shown in Figure 1. A more 

comprehensive chronicling of these past studies, as well as other aspects of general interest 

pertinent to the current effort, are briefly described in the ensuing text.

Figure 1 Graphical overview of significant SFP-related activities

In March 1979, the NRC issued NUREG/CR-0649, “Spent Fuel Heatup Following Loss of Water 

During Storage,” which provided an analysis of spent fuel heatup following a hypothetical 

accident involving drainage of the storage pool (NRC, 1979). The report included analysis to 

assess the effect of decay time, fuel element design, storage rack design, packing density, room 

ventilation, drainage level, and other variables on the heatup characteristics of spent fuel stored 

in an SFP and to predict the conditions under which clad failure would occur. The report 
11

concluded that the likelihood of clad failure caused by rupture or melting following a complete 

drainage is extremely dependent on the storage configuration and the spent fuel decay period. 

Furthermore, the minimum prerequisite decay time to preclude clad failures may vary from less 

than 10 days for some storage configurations to several years for others. The potential for 

reducing this critical decay time either by making reasonable design modifications or by 

providing effective emergency countermeasures was found to be significant. Note that this 

study considered both low-density racking and mitigative accessibility.

In the late 1980s, work related to Generic Issue (GI)-82, “Beyond Design Basis Accidents in 

Spent Fuel Pools,” culminated in the publishing of several related reports: NUREG/CR-4982, 

“Severe Accidents in Spent Fuel Pools in Support of Generic Safety Issue 82,” issued July 1987 

(NRC, 1987), NUREG/CR-5281, “Value/Impact Analysis of Accident Preventive and Mitigative 

Options for Spent Fuel Pools,” issued March 1989 (NRC, 1989a), and NUREG-1353, 

“Regulatory Analysis for the Resolution of Generic Issue 82, ‘Beyond Design Basis Accidents in 

Spent Fuel Pools,’” (NRC, 1989b). In particular, NUREG/CR-5281 investigated options 

including limited low-density re-racking of spent fuel, installation of water sprays above the SFP, 

and installation of redundant cooling, makeup systems, or both. The results of these studies 

indicated that the measures were, in general, not likely to be cost effective because of the low 

likelihood of an SFP accident that could result in a significant radiological release and the high 

cost of proposed modifications. The report goes on to conclude that these insights are largely 

contingent upon compliance with guidelines developed for licensees to ensure the safe handling 

of heavy loads in the vicinity of SFPs, thus reducing the likelihood of the structural failure of the 

pool and rapid loss of water inventory resulting from a cask drop event. 

The latter report (NUREG-1353), which draws from the preceding reports, concludes that if the 

decay heat level is high enough to heat the fuel rod cladding to about 900 degrees Celsius (C) 

the oxidation becomes self-sustaining, resulting in a Zircaloy cladding fire. The conditional 

probability of a Zircaloy cladding fire given a complete loss of water was found to be 1.0 for 

PWRs and 0.25 for BWRs in high-density configurations based on differences in assumed rack 

geometry. The conditional probability of a Zircaloy cladding fire given a complete loss of water 

in low-density storage racks is estimated to be at least a factor of five less than for the highdensity configurations. The report goes on to state that although most of the SFP risk is derived 

from beyond-design-basis earthquakes, this risk is no greater than the risk from core damage 

accidents caused by seismic events beyond the safe-shutdown earthquake (SSE). Therefore, 

reducing SFP risk resulting from events beyond the SSE would still leave at least a comparable 

risk from core damage accidents. As a result of this conclusion, the results justified the decision 

that no regulatory action was needed.

In 1996, an NRC-sponsored and issued an Idaho National Laboratories (INL) study entitled, 

“Loss of Spent Fuel Pool Cooling PRA: Model and Results,” (INL, 1996). This study considered 

a dual-unit plant and the following initiators:

• loss of SFP cooling

• LOOP

• loss of SFP water inventory (did not include heavy load drops)

• loss of primary (reactor) coolant

• seismic events

The results of this study indicated that, for the plant studied, the annual probability of SFP 

boiling is 5x10-5 and the annual probability of internal plant flooding associated with SFP 
12

accidents is 1x10-3

. Qualitative arguments are provided to show that the likelihood of core 

damage from SFP boiling accidents is low for most U.S. commercial nuclear power plants. The 

INL study also showed that, depending on the design characteristics of a given plant, the 

likelihood of either (1) core damage from SFP-associated flooding or (2) spent fuel damage from 

pool dryout may not be negligible. Section 6.3.4 further discusses this issue.

The next year, three additional reports were issued: (1) NUREG-1275, Volume 12, “Operating 

Experience Feedback Report: Assessment of Spent Fuel Cooling,” (NRC, 1997a), (2) “Followup Activities on the Spent Fuel Pool Action Plan,” (NRC, 1997b), and (3) NUREG/CR-6451, “A 

Safety and Regulatory Assessment of Generic BWR and PWR Permanently Shutdown Nuclear 

Power Plants,” (NRC, 1997c). The first of these reports concluded that the typical plant may 

need improvements in SFP instrumentation, operator procedures and training, and configuration 

control. (Note that this is the conclusion stated in the report, and has not been placed in the 

regulatory context of balance-of-plant activities since the issuance of that report.) The staff 

determined that loss of SFP coolant inventory greater than 1 foot occurred at a rate of about 

one event per 100 reactor years. Loss of SFP cooling with a temperature increase greater than 

20 degrees Fahrenheit (F) occurred at a rate of approximately three events per 1,000 reactor 

years. The primary cause of these events was found to be human error. The report also 

concluded that utilities' efforts to reduce outage duration resulted in full core offloads occurring 

earlier in outages. This increased fuel pool heat load was felt to be important because it 

reduces the time available to recover from a loss of SFP cooling event early in the outage.

In the second of these reports (known as the Spent Fuel Pool Action Plan), the staff performed 

probabilistic screening analyses and found that, in most cases, event frequencies for sequences 

associated with identified SFP design issues were sufficiently low that further analyses were not 

warranted. In one instance in which the probabilistic screening criteria were met, the staff 

performed a deterministic evaluation of the issue using plant-specific information and found that 

safety enhancements were not warranted.

The third report (NUREG/CR-6451) presents a regulatory assessment for generic BWR and 

PWR plants that have permanently ceased operation. In addition to an assessment of 

regulatory requirements in the context of decommissioning, this study looked at the potential 

offsite consequences for four phases of decommissioning (hot fuel in the SFP, cold fuel in the 

SFP, all fuel in dry cask storage, and no spent fuel onsite). The following conclusions are based 

on an assumption that for the second configuration (cold fuel in the SFP), a zirconium fire would 

not occur, and that consequences are driven by an accident where a single fuel assembly is 

dropped during movement within the SFP (akin to the design-basis fuel handling accident). The 

report concluded that, “Since the estimated consequences of the Configuration 1 representative 

accident sequence approximate those of a core damage accident, it is recommended that all 

offsite and onsite emergency planning requirements remain in place during this period, with the 

exception of the Emergency Response Data System requirements of Part 50, Appendix E. 

Subject to plant specific confirmation, the offsite emergency preparedness (EP) requirements 

are expected to be eliminated for Configuration 2, on the basis of a generic boundary dose 

calculation. Part 50 offsite EP requirements can also be eliminated for Configurations 3 and 4 

because the spent fuel has been transferred to an ISFSI (subject to Part 72 requirements) or 

transported offsite.”

Several years later, the NRC re-visited these issues by conducting an SFP risk study for 

decommissioning plants to look at the relaxation of emergency preparedness requirements, and 

in 2001 the final version was issued as NUREG-1738 (NRC, 2001). The results of the study 

indicated that the risk at SFPs is low and well within the Commission's quantitative health 
13

objectives (QHOs). The risk was found to be low because of the very low likelihood of a 

zirconium fire, even though the consequences from a zirconium fire could be serious. The 

report found that the event sequences important to risk at decommissioning plants were limited 

to large earthquakes and cask drop events. This report represented a significant undertaking, 

and remains one of the prominent studies cited in NRC decision-making on SFPs. However, 

there are some important conservatisms associated with this study that need to be considered if 

it is applied outside of its intended context (e.g., exemption requests from NRC requirements for 

offsite emergency preparedness for decommissioning reactors). These conservatisms include: 

(1) the use of assumed and often bounding configurations, (2) simplified treatment of the 

thermal-hydraulic response, (3) simplifying assumptions regarding the pool failure leakage rate 

for large seismic event and cask drop (i.e., instantaneous draindown), and (4) emergency 

preparedness response representative of a decommissioned site.

On the heels of the aforementioned study, the agency also released NUREG/CR-6441 in March 

2002, entitled, “Analysis of Spent Fuel Heatup Following Loss of Water in a Spent Fuel Pool: A 

Users’ Manual for the Computer Code SHARP,” (NRC, 2002a). This document included an 

analysis of spent fuel heatup, using “representative” design parameters and fuel loading 

assumptions. Sensitivity calculations were also performed to study the effect of fuel burnup, 

building ventilation rate, baseplate hole size, partial filling of the racks, and the amount of 

available space to the edge of the pool. The spent fuel heatup was found to be strongly 

affected by the total decay heat production in the pool, the availability of open spaces for airflow, 

and the building ventilation rate. Note that the SFP analyses performed by the NRC after this 

time did not rely on this computer code. Rather, they relied on the use of the MELCOR 

computer code (owing to its mechanistic treatment of severe accident phenomena), with 

supporting analysis using the COBRA-SFS, FLOW3D and Fluent codes, along with confirmatory 

experiments at Sandia National Laboratories (SNL).

In response to the events of September 11, 2001, the NRC undertook studies (referred to 

hereafter as security assessments) of spent fuel storage in pools and casks. While this work 

was underway, Robert Alvarez et al. published the paper, “Reducing the Hazards from Stored 

Spent Power-Reactor Fuel in the United States,” dated April 21, 2003 (hereafter referred to as 

the 2003 Alvarez paper) (Alvarez et al., 2003). In response, the NRC issued a review of the 

paper (also in 2003) which concluded that the assessment performed of possible SFP accidents 

stemming from potential terrorist attacks in the 2003 Alvarez paper did not address such events 

in a realistic manner (NRC, 2003a). The NRC response went on to state that, in many cases, 

the authors of the 2003 Alvarez paper relied on studies that made overly conservative 

assumptions or were based on simplified and very conservative models. The NRC concluded 

that the fundamental recommendation of the 2003 Alvarez paper, namely that all spent fuel 

more than 5 years old be placed in dry casks through an expedited 10-year program costing 

many billions of dollars, was not justified.

Continued discussions on the issue of SFP safety and security led to a 2004-2005 National 

Academies study, documented in “Safety and Security of Commercial Spent Nuclear Fuel 

Storage,” issued in 2006 (NAS, 2006). This study was Congressionally mandated (e.g., see 

[Congress, 2005]). The National Academies committee was briefed on numerous occasions by 

the NRC staff regarding past and ongoing studies related to the subject topic. The study 

resulted in a classified report and the aforementioned publicly available report. The publicly 

available report documented numerous findings and recommendations, many of which were 

addressed as part of the NRC’s continued activities in this area (e.g., site-specific assessments 

of licensee response to develop strategies to maintain or restore SFP cooling capabilities). 
14

The NRC's initial response to the study was documented in a letter from the NRC Chairman 

(Nils Diaz) to Senator Peter Domenici, dated March 14, 2005 (NRC, 2005a). In that response, 

NRC expressed its appreciation for the insights of the National Academies committee, noting 

that many of the conclusions mirrored the NRC’s conclusions from prior work, which guided 

NRC initiatives. However, the NRC disagreed with some of the conclusions from the National 

Academies study, including the finding that the NRC might determine that the earlier movement 

of spent fuel from pools to dry cask storage would be prudent, depending on the outcome of 

plant-specific vulnerability analysis. “The Commission views the results of security 

assessments completed to date as clearly showing that storage of spent fuel in both SFP and in 

dry storage casks provides reasonable assurance that public health and safety, the 

environment, and the common defense and security will be adequately protected. The NRC will 

continue to evaluate the results of the ongoing plant-specific assessments and, based upon new 

information, would evaluate whether any change to its spent fuel storage policy is warranted.” 

The NRC’s position on each finding or recommendation that it disagreed with is contained in the

report to Congress that accompanied the March 2005 letter.

In parallel to the National Academies study, the NRC continued performing the aforementioned 

security assessments, which were completed in 2006-2008. While the results of these studies 

are not publicly available because of their nature (i.e., containing sensitive information that could 

be useful to an adversary), the conclusions of the studies were integrated into the NRC’s 

regulatory licensing and oversight processes (e.g., 10 CFR 50.54(hh)(2) as a result of the 

Power Reactor Security Rulemaking). Activities related to the development of new securityrelated requirements were later documented in a memorandum to the NRC Commission 

entitled, “Documentation of Evolution of Security Requirements at Commercial Nuclear Power 

Plants with Respect to Mitigation Measures for Large Fires and Explosions,” dated February 4, 

2010 (NRC, 2010).

Also in parallel to the above activities, the agency conducted a pilot PRA for dry cask storage 

documented in NUREG-1864, “A Pilot Probabilistic Risk Assessment of a Dry Cask Storage 

System at a Nuclear Power Plant,” issued March 2007 (NRC, 2007). The report’s analysis 

indicates that dry cask storage risk is solely from latent cancer fatalities, and no prompt fatalities 

are expected. Dry cask storage risk was found to be dominated by accident sequences 

occurring in three stages of the handling phase. These involved the drop of the transfer cask 

through the equipment hatch (termed Stage 18) and drops of the multipurpose canister (MPC) 

into the storage overpack (Stages 20 and 21). The aggregated risk values were quite low. The 

estimated aggregate risk was an individual probability of a latent cancer fatality of 1.8x10-12

during the first year of service, and 3.2x10-14 per year during subsequent years of storage. Note 

that when insufficient information was available, "conservative bounding assumptions or 

estimates" were used. Other limitations of the study included no consideration of uncertainty 

and conservative assumptions about the translation of failure modes to leak sizes.

Two other documents of regulatory interest were issued in 2008 and 2009. The first was the 

denial of two PRMs, as documented in SECY-08-0036, “Denial of Two Petitions for Rulemaking 

Concerning the Environmental Impacts of High-Density Storage of Spent Nuclear Fuel in Spent 

Fuel Pools (PRM-51-10 and PRM-51-12),” dated March 7, 2008, and the associated staff 

requirements memorandum (NRC, 2008a). These documents describe the NRC's denial of 

PRMs filed by the Attorney General of the Commonwealth of Massachusetts and the Attorney 

General for the State of California, which presented nearly identical issues and requests for 

rulemaking concerning the environmental impacts of high-density storage of spent nuclear fuel 

in SFPs.
15

The second document is the issuance in 2009 of the draft report for comment of Revision 1 to 

the NRC’s Generic Environmental Impact Statement (GEIS) on License Renewal 

(NUREG-1437, Revision 1 (NRC, 2009). This document reevaluated SFP environmental 

considerations related to SFPs by considering information developed since the original license 

renewal GEIS was issued in 1996 (NRC, 1996). The update concluded that the environmental 

impacts from accidents at SFPs (as quantified in NUREG-1738) can be comparable to those 

from reactor accidents at full power (as estimated in NUREG-1150 (NRC, 1990)). The updated 

GEIS goes on to state that subsequent analyses performed, and mitigative measures employed, 

since 2001 have further lowered the risk of SFP accidents; and even the conservative estimates 

from NUREG-1738 are much less than the impacts from full power reactor accidents as 

estimated in the original 1996 GEIS. As a result of these considerations, the update concludes 

that the environmental impacts stated in the 1996 GEIS bound the impact from SFP accidents.

Finally, in July 2011, the NRC issued, “Recommendations for Enhancing Reactor Safety in the 

21st Century: The Near-Term Task Force Review of Insights from the Fukushima Daiichi 

Accident” (NRC, 2011a). This report makes two sets of conclusions and recommendations 

related to spent fuel pool safety. The first occurs in the section of the report on prolonged loss 

of ac power. In this section, the task force stated the following:

The Commission’s [station blackout] SBO requirements provide assurance that each 

nuclear power plant can maintain adequate core cooling and maintain containment 

integrity for its approved coping period (typically 4 or 8 hours) following an SBO. 

Also, if available, the equipment used for compliance with 10 CFR 50.54(hh)(2) 

would provide additional ability to cool either the core or the spent fuel pool and 

mitigate releases from primary and secondary containment during a prolonged SBO. 

The implementing guidance for SBO focuses on high winds and heavy snowfalls in 

assessing potential external causes of loss of offsite power, but does not consider 

the likelihood of loss of offsite power from other causes such as earthquakes and 

flooding. Also, the SBO rule does not require the ability to maintain reactor coolant 

system integrity (i.e., PWR reactor coolant pump seal integrity) or to cool spent 

fuel....

The Task Force concludes that revising 10 CFR 50.63 to expand the coping 

capability to include cooling the spent fuel, preventing a loss-of-coolant accident, and 

preventing containment failure would be a significant benefit.

The task force went on to recommend orders requiring reasonable protection of the equipment 

provided pursuant to 10 CFR 50.54(hh)(2) and the acquisition of additional sets of equipment as 

needed to address multiunit events. The task force also recommended a rulemaking 

to revise 10 CFR 50.63 to require each operating and new reactor licensee to 

(1) establish a minimum coping time of 8 hours for a loss of all ac power, (2) 

establish the equipment, procedures, and training necessary to implement an 

“extended loss of all ac” coping time of 72 hours for core and spent fuel pool 

cooling and for reactor coolant system and primary containment integrity as 

needed, and (3) preplan and prestage offsite resources to support uninterrupted 

core and spent fuel pool cooling, and reactor coolant system and containment 

integrity as needed, including the ability to deliver the equipment to the site in the 

time period allowed for extended coping, under conditions involving significant 

degradation of offsite transportation infrastructure associated with significant 

natural disasters.
16

The second set of conclusions and recommendations is included in the section of the report on 

SFP Safety, where the task force concluded the following:

clear and coherent requirements to ensure that the plant staff can understand the 

condition of the spent fuel pool and its water inventory and coolability and to provide 

reliable, diverse, and simple means to cool the spent fuel pool under various 

circumstances are essential to maintaining defense-in-depth. 

The task force goes on to recommend orders addressing: (1) SFP instrumentation, (2) safetyrelated ac power for SFP makeup, (3) technical specification revision regarding onsite ac power 

for SFP makeup and instrumentation, and (4) a seismically-qualified spray capability. The task 

force also recommended rulemaking or licensing actions (or both) to require the above actions.

The U.S. nuclear industry has also undertaken various studies related to spent fuel storage and 

transportation. Examples include the following:

• Electric Power Research Institute (EPRI) TR-1003011, “Dry Cask Storage Probabilistic 

Risk Assessment Scoping Study,” issued in 2002

• EPRI TR-1009691, “Probabilistic Risk Assessment (PRA) of Bolted Storage Casks: 

Updated Quantification and Analysis Report,” issued in 2004

• EPRI TR-1021049, “Impacts Associated with Transfer of Spent Nuclear Fuel from Spent 

Fuel Storage Pools to Dry Storage After Five Years of Cooling,” issued in 2010

• EPRI TR-1025206, “Impacts Associated with Transfer of Spent Nuclear Fuel from Spent 

Fuel Storage Pools to Dry Storage After Five Years of Cooling, Revision 1,” issued in 

2012

The last two reports are of particular interest for the present effort. EPRI TR-1021049 

assesses the cost and risk impacts (from a worker dose perspective) associated with transfer of 

spent nuclear fuel from SFPs to dry storage after 5 years of cooling. The report concludes that 

expedited fuel movement would result in an increase cost to the U.S. nuclear industry of $3.6 

billion, with the increase primarily related to the additional capital costs for new casks and 

construction costs for the dry storage facilities. The report goes on to conclude that early 

movement of spent fuel into dry storage would have “significant radiological impacts.” These 

impacts are stated in terms of worker radiation exposure, and are estimated to be 507 personrem over 60 years as a result of the additional handling of spent fuel. With respect to SFP 

accidents, the report estimates that an additional 711 dry storage packages would have to be 

handled, as compared to the case without expedited fuel movement, thus increasing the risks 

associated with cask movement (based on a need to reduce the number of assemblies in some 

casks when loading more recently-discharged fuel to maintain overall heat load limits). A report 

prepared by NAC International entitled, “NAC White Paper on Establishing a Balanced 

Perspective on Wet and Dry Storage of Used Fuel at U.S. Reactors,” dated July 7, 2011, makes 

similar arguments with respect to the impacts of expediting fuel movement (NAC, 2011).

The updated EPRI study, EPRI TR-1025206 (EPRI, 2012), revised the 2010 study to evaluate 

the dose and cost impacts of accelerating transfer of spent fuel considering two scenarios – one 

where the campaign takes 10 years and one where it takes 15 years. The report also adds 

estimates of the reduction in Cs-134 and Cs-137 inventories in the SFP due to accelerated 

transfer of spent fuel. The updated report estimates the worker doses to be much higher than 

the 2010 study projected (3 to 4 times higher), while the costs are roughly equivalent. The 
17

reduction in Cs-134 and Cs-137 inventories reported range from 43% to 53%. None of the 

industry studies attempt to calculate offsite consequences associated with postulated SFP 

accidents, which is a significant difference between those studies and the study documented in 

this report.

Regarding the amount of fuel older than five years, and its associated decay heat, the table 

below compares industry averages reported in the NAC study with those from the study 

presented in this report.

Table 2 Comparison of Fuel Age and Heat Load against Industry Averages

Mass as a % of all fuel Heat generation as a % of all fuel

Time since 

discharge (yrs.)

Industry 

average This study

Industry 

average This study

< 5 22% 18% 58% 58-90%

5-9 22% 27% 22% 6-22%

10-14 16% 18% 9% 2-8%

15-19 15% 19% 6% 1-7%

20-24 10% 17% 3% 1-4%

25-29 6% 1% 1% 0-1%

30-34 4% - <1% -

Remainder 4% - <1% -

The NAC white paper and the latter (2012) EPRI study, make the case that heat load 

distributions like the ones in Table 2 support the notion that moving fuel older than 5 years has 

only modest effects on the overall SFP heat load (and thereby the cooling requirements and 

mitigative time available for beyond-design-basis SFP accidents). The values in the table for 

the site studied here highlight the caution that accompanies treating the heat load as a point 

estimate (the range of values in this study represent snapshots during the operation cycle). 

That said, the values from the reference plant (across the representative operating cycle) only 

strengthen the argument that the SFP heat load is driven by the fuel less than five years old.

1.8 Potential Follow-On Work and Related Activities

It is important to recognize that there are several ongoing activities that have a peripheral 

relationship to this study. These include, but are not limited to, the following:

• NRC Order EA-12-049, “Issuance of Order to Modify Licenses with Regard to 

Requirements for Mitigation Strategies for Beyond-Design-Basis External Events” (NRC, 

2012g)

• NRC Order EA-12-051, “Issuance of Order to Modify Licenses with Regard to Reliable 

Spent Fuel Pool Instrumentation” (NRC, 2012h)

• 10 CFR 50.54 (f) letters to licensees to review seismic hazards (NTTF Recommendation 

2.1)

• 10 CFR 50.54 (f) letters to licensees to review onsite shift minimum staffing levels for 

emergency response and performance of mitigating strategies in accordance with Order 

EA-12-049 (NTTF Recommendation 9.3)
18

• SECY 12-0095, Recommendation AR 5 "Expedited Transfer of Spent Fuel from Spent 

Fuel Pools to Dry Storage"

• an ongoing rulemaking related to security requirements for ISFSIs

• reevaluation of the role of defense-in-depth in regulatory decision-making

• reconsideration of the use of land contamination and economic consequences in the 

context of regulatory decision-making

• assessment of the effects of seismic events and accident conditions on neutron 

absorber materials used in SFPs

• performance of a site Level 3 PRA for Vogtle Units #1 and #2 (operating PWRs), 

including consideration of both wet and dry storage per SECY-11-0089.

Other aspects of SFP risk that have not been informed by this or past studies, may be 

addressed by future studies, such as the site Level 3 probabilistic risk assessment (PRA), as 

documented in SECY-11-0089, “Options for Proceeding with Future Level 3 Probabilistic Risk 

Assessment Activities,” dated July 7, 2011, and the associated staff requirements 

memorandum; or will be addressed through other inputs to the regulatory decision-making 

process, as needed.

1.9 Layout of Remainder of This Report

The remaining sections of this report provide the following information:

• major assumptions and limitations

• seismic hazard characterization

• structural analysis methods and results

• scenario delineation and probabilistic considerations

• accident progression analysis methods and results

• offsite consequence analysis methods and results

• human reliability analysis

• sensitivity studies to investigate selected assumptions

• comparison of results with past wet and dry storage consequence and risk studies

• summary of backfitting screen analysis

Finally, Appendix A provides details on the emergency response models, Appendix B provides a 

gap analysis related to the larger question of assessing the impacts of expedited fuel 

movement, Appendix C provides study-related correspondence with the US NRC’s Advisory 

Committee on Reactor Safeguards (ACRS) and Appendix D provides a regulatory analysis for 

expedited transfer at the reference plant.
19

2. MAJOR ASSUMPTIONS

2.1 Study Assumptions

Assumptions made during the conduct of this study are documented throughout this report. For 

reader convenience, major assumptions are catalogued in Table 3.

Table 3 Major Assumptions

Topical Area Major Assumption Comment

Overall 

Approach

A BWR Mark I with 

a non-shared SFP is 

studied.

This plant was chosen for a mix of reasons, including 

availability of computer models, and a perception of 

greater external stakeholder interest in elevated 

(relative to grade) SFPs combined with the fact that 

the nuclear reactors that felt the higher tsunami 

waves and stronger ground motions from the March 

11, 2011 Tohoku earthquake, which includes those at 

the Fukushima Daiichi nuclear power plant, were all 

BWR reactors. Its selection does not denote a belief 

that this type of design is more vulnerable.

The beyond design 

basis earthquake is 

assumed to occur. 

This is an unlikely 

event.

The earthquake studied has an estimated frequency 

of occurrence of one time in 60,000 years. The 

likelihood of the event is included in the reporting of 

frequency-weighted consequences.

The reliability of 

mitigation is not 

included in the 

likelihood estimates 

provided in Chapter 

5 through Chapter 7.

The accident progression and consequence analyses 

were originally conducted without the benefit of a 

human reliability analysis. The results were then used 

to frame a human reliability analysis, which is 

provided in Chapter 8.

Multi-unit / 

concurrent reactor 

accidents are not, 

in general, 

considered.

Specifically, the reactor (and its decay heat) is treated 

during the outage until the level in the reactor well / 

SFP drops to below the bottom of the fuel transfer 

canal. Beyond that point, and in all portions of the 

post-outage scenarios, the reactor is not considered 

as a source of steam, fission products or hydrogen. 

The human reliability analysis presented in Chapter 8

does consider multi-unit effects, and Sections 9.3 and 

2.2 further discuss this assumption.

This study 

represents a limitedscope 

consequence study

as opposed to a 

probabilistic risk 

assessment.

This approach focuses resources on a particular 

scenario of interest and places greater emphasis on 

modeling fidelity for that scenario, but also limits the 

potential end-uses of the study. See Section 1.5 for 

more information on this assumption.

Multi-unit aspects

are only considered 

for certain aspects of 

the study.

See Section 2.2 for more information on this 

assumption.
20

Topical Area Major Assumption Comment

Inadvertent 

criticality events 

are not considered.

See Section 2.3 for more information on this 

assumption.

Other 

considerations

associated with 

expedited fuel 

movement.

See APPENDIX B: for a qualitative consideration of 

fuel movement and APPENDIX D: for quantitative 

considerations.

Seismic Hazard 

Characterization

Vertical PGA equal 

to the horizontal 

PGA and vertical 

spectral 

accelerations equal 

to the horizontal 

spectral 

accelerations

A few studies (e.g., McGuire, Silva, and Costantino, 

2001; ASCE, 1999) indicate that for rock sites and 

frequencies near and above 10 Hz, and especially 

nearby seismic sources, vertical spectral 

accelerations may be as high as or exceed horizontal 

spectral accelerations. For this study, the frequencies 

of interest are, for the most part, frequencies near or 

above 10 Hz. Therefore, the assumption of equal 

vertical and horizontal spectral accelerations was 

deemed to be a reasonable starting assumption. This 

assumption is also supported by seismic hazard deaggregation with the USGS 2008 model which 

indicates that for the seismic bins of interest (high 

PGA, low likelihood hazard) the contributors to the 

hazard would be earthquakes with magnitudes less 

than 6 at about 20 km from the site.

Seismic hazard 

models - this study 

used the existing 

USGS 2008 model 

instead of the model 

in the ongoing 

program. 

A new probabilistic seismic hazard model is currently 

being developed and will consist of two parts: (1) a 

seismic source zone characterization and (2) a 

ground motion prediction equation (GMPE) model. 

Although part (1) is now complete (NRC, 2012b), it 

was not available at the start of this scoping study. In 

addition, the GMPE update is still in progress. 

Furthermore, the NRC is currently developing an 

independent probabilistic seismic hazard assessment 

(PSHA) computer code to incorporate part (1) and 

part (2) when complete. While the USGS (2008) 

hazard model is not sufficiently detailed for regulatory 

decisions, it is appropriate to use for this study 

because it was the most recent and readily available 

hazard model for the selected site at the start of the 

study.
21

Topical Area Major Assumption Comment

Structural and 

Related Initial 

Damage State 

Characterization

In-structure 

response spectra 

(ISRS) for the study 

are obtained by 

scaling ISRS 

developed for the 

seismic PRA for 

PBAPS for the 

NUREG-1150 study.

Given the differences in the ground motions for the 

NUREG-1150 PRA and for this study, the use of this 

scaling is likely to be at or somewhat past the limit of 

acceptability. The scaling was nevertheless used 

because it is consistent with the practice of expedited 

studies of risk or margin. In addition, the assumption 

(and the uncertainties that it introduces) were deemed 

to be consistent with the uncertainties in other 

approximations used in the structural and seismic 

assessments for the study. 

A static nonlinear 

pushover analysis

is used to estimate 

the overall response 

of the SFP, concrete 

strains and cracking, 

and related liner 

strains. This 

analysis used 

equivalent seismic 

forces, including 

hydrodynamic 

forces, based on 

elastic ISRS. 

As the structure cracks and behaves in a nonlinear 

manner, it becomes sensitive to frequencies less than 

its elastic frequencies and dissipates energy, 

especially if the structure can respond in a non-brittle 

failure mode. For the ground motions considered in 

this study, lower frequencies of vibration tends to 

correspond to lower loads on the structure because 

spectral accelerations also decrease as the 

frequencies of vibration decrease. The approach used 

is, therefore, thought to be conservative for the 

frequency content of the ground motion expected at 

the site. Some of this conservatism was accounted 

for, in part, by: (i) considering a higher damping ratio 

for the SFP/Reactor-building system (10-percent) 

than that used for the design basis loads, (ii) 

calculating equivalent loads using a reduced concrete 

stiffness, and (iii) by considering a small range of 

reduction for the calculated liner strains that also 

accounts for possible ground motion incoherency. A 

preferred approach to account for this conservatism 

and also ground motion incoherency effects would 

involve the sampling of representative ground motion 

time-histories, both vertical and horizontal, and their 

use in time-history analyses of the coupled reactor 

building and SFP structures. This approach, ideally 

also modeling the small embedment of the reactor 

building using soil-structure interaction analysis, 

would more rigorously account for the anticipated 

conservatisms referred to above but was deemed to 

be outside the scope of the study. 

Aftershocks are not 

likely to induce 

subsequent 

additional damage to 

the SFP

The main event would crack the SFP in this 

study, however it is expected that the SFP's structure 

would remain stable after the earthquake and 

resistant to additional loading cycles at this level.
22

Topical Area Major Assumption Comment

No significant debris

generated by the 

seismic event

enters the SFP.

Based on the expected structural response of the 

building, overhead crane, etc. there is no expectation 

that heavy debris that would damage the pool and 

fuel will be generated as a direct result of the seismic 

event itself.

The seals of the 

refueling gate do 

not fail.

Finite element analysis does not predict large 

deformations in this area that would suggest such an 

event is likely. Details of the gates provided by the 

licensee show that there are two gates with a gap in 

between and that each gate has mechanical seals to 

prevent leakage. These seals are kept under pressure 

by passive mechanical means (i.e., do not depend on 

air pressure, ac power, or dc power) that are unlikely 

to fail under the earthquake.

Failure of nearby 

dams is not explicitly 

addressed.

The Conowingo Dam is located approximately 9 miles 

downstream of the site. Failure of that dam could not 

flood the site. It could lead to additional complications 

for accident management strategies relying on the 

river as a water source. The Holtwood Dam is located 

approximately 6 miles upstream of the site. Failure of 

this dam, or partial failure in combination with the 

probable maximum flood, is considered in the plant’s 

Updated Final Safety Analysis Report (UFSAR), in 

Section 2.4.3.5. Based on the UFSAR, should a 

complete failure in the upstream dam take place, the 

rise in the Conowingo Pond level at the site is not 

expected to exceed grade level, since the pond is 

about 1 mi wide at the site and the water level would 

be relieved by the downstream dam.

Scenario 

Delineation and 

Probabilistic 

Treatment

Offloading of older 

fuel in to casks (as 

part of the normal 

fuel management 

practices as 

opposed to an 

expedited fuel 

movement program) 

is not explicitly 

treated.

This assumption is not expected to have a significant 

effect on the results. See Section 5.2 for more 

information.
23

Topical Area Major Assumption Comment

A full core offload

is not treated 

(except as 

discussed to the 

right) as either part 

of the routine 

refueling or in the 

context of an 

emergent need to 

defuel the reactor 

later in the operating 

cycle (e.g., due to a 

forced shutdown that 

requires accessing 

the lower internals of 

the reactor vessel).

In reality, the full core’s decay heat is considered 

during the outage, in that the reactor and SFP are 

hydraulically connected, and all fuel contributing to 

pool heatup is considered (along with the larger 

volume of water) until the point that water level drops 

below the fuel transfer canal (and the reactor well and 

SFP become hydraulically disconnected). That being 

said, radiological release from the fuel remaining in 

the reactor is not considered, since the simulation 

focuses only on the SFP once the reactor well and 

SFP have become hydraulically disconnected. The 

rationale for choosing a “core shuffle” rather than a 

full core offload is because the former is the typical 

case for BWRs. Emergent core offloads later in the 

operating cycle are not typical, and thus are not 

treated.

New fuel temporarily 

stored in the spent 

fuel pool is not 

treated.

This fuel would be placed in the spent fuel pool just 

prior to the outage (the subject plant does not use a 

separate new fuel vault). Thus, the fuel would only be 

present for a very short portion of the operating cycle. 

During the time that the new fuel is in the SFP, it 

would affect the amount of zirconium present to 

participate in a propagating zirconium fire, but would 

have a negligible effect on the source term. See 

Section 5.2 for additional information.

Use of a 1x4 

pattern, rather than 

the 1x8 pattern 

currently in use at 

PBAPS.

The 1x8 pattern currently in use at PBAPS is believed 

to be atypical and is not required by regulation. The 

timing of obtaining the actual pool configuration, along 

with modeling conveniences associated with how the 

MELCOR SFP model is currently designed, also 

played a role in the decision to use the 1x4 pattern. 

In cases where the use of a 1x8 pattern might affect 

study conclusions, this is identified, and Section 9.2

investigates this assumption.

For the low-density 

loading situation, the 

high-density 

racking will be used 

as opposed to lowdensity racking. 

Re-racking the pool would represent a significant 

expense, along with additional worker dose, and was 

not felt to be the likely regulatory approach taken 

based on consultation with the Office of Nuclear 

Reactor Regulation. Much of the benefit of lowdensity racking is achieved by the implementation of a 

favorable fuel pattern (1x4). Additionally, to get the full 

benefit of low-density racking, BWR fuel would likely 

need to have the channel boxes removed.

Effects on results if a 

contiguous storage 

pattern were used 

during the outage.

See Section 9.3 for more information on this 

assumption.
24

Topical Area Major Assumption Comment

An assembly in the

lifted position (i.e., 

in the process of 

being moved) at the 

time of the seismic 

event is not treated. 

The current tools do not allow for explicit treatment of 

this situation. Such a situation could lead to 

accessibility issues (which are already treated via the 

scenarios without 50.54(hh)(2) equipment), but could 

also lead to a small earlier release for some 

situations. Note that Section 5.4 does include 

information about dose rates on the refuel floor 

associated with uncovering a single assembly in the 

lifted position.

50.54(hh)(2) 

mitigation 

capacities (i.e., 500 

gpm makeup 

delivered or 200 

gpm spray 

delivered) are based 

on the generic NRCendorsed capacities 

in NEI-06-12, 

Revision 2.

For PBAPS, the capacities of the available equipment 

are somewhat higher. The use of 500 and 200 gpm 

here attempts to account for uncertainties in the 

speed at which the pumps would actually be run, as 

well as spray that goes outside the boundary of the 

pool

Mode of mitigation 

deployment (i.e., 

use of makeup 

versus spray)

For OCP 1 and OCP 2 with the “moderate” leakage 

condition, makeup is deployed. Other, equally 

reasonable assumptions about mitigation deployment 

could result in the deployment of sprays instead

(which have a potential advantage in terms of 

mitigation for these conditions). This difference in 

mode of deployment shows the potential benefit of the 

additional instrumentation required by NRC Order EA12-051. A sensitivity study related to this assumption 

is presented in Section 9.3, for a uniform pattern.

Use of ac power 

fragility as a 

surrogate for loss of 

normal SFP 

cooling and 

makeup availability

This study used the ac power fragility from NUREG1150 of 0.84 as a surrogate for the conditional

probability of normal SFP cooling and makeup not 

being available. This simplifying assumption was 

made in light of the fact that the study is not a PRA 

(but rather a consequence analysis with probabilistic 

considerations) and that this value already 

approximates the upper bound of 1. In reality, the 

availability of normal SFP cooling and makeup would 

be a combination of the AC fragility, the fragility of the 

actual equipment and its support equipment, and 

operator actions to recover the equipment, which 

could result in a conditional probability higher than the 

value used here.
25

Topical Area Major Assumption Comment

Accident 

Progression 

Analysis

The study uses 

best-estimate 

ruthenium release 

rates calculated by 

the MELCOR code. 

These release rates 

are most similar to 

the low ruthenium 

release case from 

NUREG-1738.

This is the best estimate for actual releases based on 

the current state of knowledge in this area. Past 

studies for which this was a concern (namely 

NUREG-1738) used assumed source terms spanning 

a very large range of uncertainty rather than 

mechanistic and integrated modeling. Section 6.1.5

of this report provides additional information.

Radionuclide 

releases occur only 

if the fuel has 

become uncovered 

by 48 hours and the 

radiological release 

has commenced

before 72 hours. 

Otherwise, the study

assumes the 

scenario results in 

no offsite 

consequences.

In the event of a prolonged severe accident, radiation 

and other hazards could make any truncation of an 

ongoing SFP release challenging. On the other hand, 

many resources are available at the State, regional, 

and national level that could be available to mitigate 

an accident. Considering both viewpoints on this 

issue, the project staff judged 72 hours to be a 

reasonable time truncation. The use of a time 

truncation is a point of uncertainty that can 

significantly affect the results. See Section 5 of the 

report for additional discussion on mitigation 

assumptions in this study and Section 9.8 for time 

truncation sensitivity.

The study does not 

consider debris

entering the pool as 

a result of any 

modeled hydrogen 

combustion event.

Such debris could be generated and could fall into the 

pool. However, the occurrence of a hydrogen 

combustion event in this study denotes that the fuel in 

the SFP has already become uncovered and is 

undergoing a fission product release. Thus, debris 

would primarily serve to inhibit longer term recovery 

actions not considered in this study. The occurrence 

of a hydrogen combustion event from a concurrent 

reactor accident has the potential to generate debris 

which could impair SFP natural circulation air or 

steam cooling (should the fuel in the SFP become 

uncovered) for conditions in which the fuel might 

otherwise be cooled by means of these passive 

cooling modes. However, this latter situation is 

inherently tied to the study’s lack of a comprehensive 

treatment of multiunit aspects.

Aerosol 

resuspension

inside the reactor 

building, such as 

from hydrogen 

deflagration, will not 

be significant.

Hydrogen burns in the refueling bay are predicted to 

occur about the time of fuel gap release and well 

before significant amounts of radioactive aerosols 

may settle on the floor.
26

Topical Area Major Assumption Comment

The study does not 

consider the effects 

of molten coreconcrete 

interaction (MCCI).

The MELCOR code models heat transfer from the 

debris to the pool floor, as well as the fission product 

release from hot debris. In some cases, the debris 

temperature remains above typical concrete ablation 

temperatures (~1500 K). MCCI may occur in selected 

scenarios in which the fuel relocated to the bottom of 

the pool following the failure of the rack baseplate and 

its temperature exceeded the concrete ablation 

temperature. These cases involve large-scale debris 

relocation and large releases of volatile fission 

products. Even without MCCI, the fuel in debris form 

continues to release fission products resulting in very 

large releases of volatiles. Section 9.5 of this report 

presents a sensitivity calculation.

The effective time 

dependent

decontamination 

factor (DF) of the 

reactor building can 

be used to 

reasonably estimate 

a cumulative 

release.

The use of an effective DF is based on a new 

methodology (see Section 6.1.5 of the report) for 

SFPs in an effort to account for a spatial distribution 

of the inventory and to more accurately account for 

the magnitude of the release based on the 

radionuclide, and not just the chemical group, to allow 

the offsite consequence code to process the source 

term.

Criteria for release 

of radionuclides in 

the fuel cladding 

gap

MELCOR does not have a fuel cladding deformation 

and strain model. It uses a value of 900°C for 

widespread cladding failure. NUREG-1738 cites a 

temperature range of 700–850°C for rod ballooning 

and burst; however, the security assessment work 

mentioned in Section 1.7 showed that rod ballooning 

has a low impact on the timing to breakaway oxidation 

and the impact on the peak cladding temperature 

response was relatively insignificant. In addition, 

NUREG-1738 assumes 900°C as the temperature at 

which the onset of significant fission product release 

is expected. In general, there may be some fuel 

cladding failures at lower temperatures but MELCOR 

is mostly concerned with larger thermal release from 

the fuel. In this sense, the gap release temperature of 

900°C is taken as a surrogate for start of rapid fuel 

heat up associated with breakaway oxidation and 

initiation of zirconium fire and its propagation.

Offsite 

Consequence 

Analysis

Calculated results 

are from

atmospheric-type 

releases only

Atmospheric releases are the primary scope of the 

project.
27

Topical Area Major Assumption Comment

A straight-line 

Gaussian plume

segment dispersion 

model is used for the 

atmospheric 

transport

The current model is a straight-line transport of plume 

segments; therefore, it does not capture the effects 

from changes in wind direction after the plume 

segment has been released. Despite this, the 

atmospheric transport model in MACCS2 has 

compared favorably to Lagrangian particle tracking 

models [NRC 2004]. This is because the use of 

ensemble averages of many meteorological 

conditions, such as the consequences reported in this 

study, has been shown to make reliable weatheraveraged results. 

Distance truncation

(from point of 

release)

Health effect risk estimates (e.g. latent cancer fatality 

risk and early fatality risk) are with respect to 

distance. The reported latent cancer fatality risk 

includes all distances that have doses above the 

modeled dose limit for habitability, as determined by 

the Pennsylvania Code Title 25 § 219.51. Total 

health effect estimates are not a function of distance, 

and have no distance truncation. 

The effect of low 

dose radiation on 

latent cancer 

fatalities is 

uncertain, and 

therefore a range of 

dose truncations are

reported.

See Section 7.1.3 for more information on this 

assumption.

The public will 

behave in an orderly

fashion during a 

severe accident, and 

can be represented 

by cohorts.

See Section 7.1.4 more information on this 

assumption.

The seismic event 

has a limited effect 

on emergency 

response.

The study assumed that the seismic event would not 

significantly affect emergency response. This is

based on an assessment in NUREG-1935 of the 

same site and seismic event that assumed the 

damage to local infrastructure is limited to 12 bridges, 

partly due to the few large structures in the area. 

Also, the extended loss of ac power is assumed to be 

limited to the EPZ (~10 miles) due to the assumption 

that the strength of the seismic event is from the 

proximity of the seismic event to the site, rather than 

being a wider impact from a larger magnitude. See 

section 7.1.4 for more details.
28

Topical Area Major Assumption Comment

Decontamination

will occur only if it 

will eventually allow 

for the return of land 

to habitability, and if 

it is economic to do 

so.

A long-term cleanup policy for severe accidents does 

not currently exist, although guidance is currently 

being drafted. In addition, guidance could 

recommend the development of localized cleanup 

goals after an accident, to account for sociopolitical, 

technical, and economic considerations.

Given that a policy for long-term cleanup does not 

currently exist (and because a developed policy may 

not contain explicit cleanup goals), the project instead 

uses dose levels associated with habitability to decide 

what land is to be decontaminated. This is consistent 

with previous studies. See Section 7.1.5 for more 

information on this assumption.

A single value for 

habitability is used 

for all affected 

areas. 

See Section 7.1.5 for more information on decisions 

regarding land interdiction and associated relocation 

of the public.

2.2 Multi-Unit Considerations

Observations Regarding a Concurrent Reactor Event:

There are four broad interplays that can be defined between the SFP and the reactor:

(1) an initiating event that directly affects both the reactor and the SFP

(2) a reactor accident that prevents accessibility to the SFP for a prolonged period of time

(e.g., due to high radiation fields), leading to a SFP accident

(3) a reactor accident that includes ex-containment energetic events (e.g., a hydrogen 

combustion event) or other ex-containment interplays (e.g., steaming through the drywell 

head that affects refuel floor combustible gas mixtures) and creates a hazard to the SFP 

(e.g., by causing debris to fall in to the pool) or otherwise changes the SFP event 

progression5

(4) an SFP accident that prevents accessibility to key reactor systems and components for a 

prolonged period of time or which creates a hazard for equipment used to cool the 

reactor (e.g., the flooding of low elevations of the reactor building due to a leak in the 

 5 For instance, a hydrogen combustion event caused by a reactor accident that affects the refuel 

floor superstructure can lead to additional oxidation (for an otherwise oxygen-limited situation), 

which in turn may result in higher releases from the SFP. Note that this can also include positive 

effects, in the sense that steam leaking through the drywell head can serve to steam inert the 

refueling floor.
29

pool or excessive condensation from continuous boiling of SFP water), leading to a 

reactor accident

For each of these interplays, large seismic events and severe weather SBO events are logically 

the most relevant initiators, as they are the type of initiators that are most likely to initiate an 

accident at the reactor and SFP, while simultaneously hampering further accessibility to key 

areas, key systems and components, and key resources. To the extent practicable, this study 

has attempted to qualitatively account for some of these effects. For example, when the reactor 

and SFP are hydraulically connected (during refueling), the decay heat and water volumes from 

both sources are considered. The study also explores these effects on mitigation (Section 8), 

and addresses some aspects of the uncertainty associated with this treatment (Section 9). 

However, explicitly modeling multiunit effects was not a focus of this study, because of the 

existing limitations with the available computational tools. An ongoing project described in 

SECY-11-0089 will attempt to more rigorously address these effects in the framework of a 

multiunit Level 3 PRA for Vogtle Electric Generating Plant Units 1 and 2.

Observations Regarding a Multiunit Event:

Along with the possibility of a concurrent SFP and reactor accident, there is the possibility for a 

concurrent accident at the SFP of one unit with an accident at the SFP or reactor of the other 

unit. Again, a large seismic event or a severe weather SBO are the events that are most likely 

to lead to a multiunit event. In general, if accidents at both SFPs proceed in similar manners 

and similar timeframes, and both pools have similar inventories of spent fuel, then the resulting 

source term from a dual-unit event would be roughly twice the single-unit source term. In 

reality, this type of perfect symmetry is unlikely because the two (or more) SFPs are very 

unlikely to have the same total pool heat load or peak assembly heat load. (Recall that for 

multiunit sites, the reactors did not usually start operation at the same time and outages are 

intentionally staggered.) Even if this symmetry did exist, the offsite consequences would not 

follow a linear scaling because of a number of nonlinearities associated with that portion of the 

analysis. Again, capturing these effects was not a focus of this study, and future work (the 

SECY-11-0089 Level 3 PRA) will attempt to more rigorously treat these effects.

2.3 Inadvertent Criticality

Inadvertent criticality events (ICEs) may be possible for specific combinations of conditions 

(e.g., during reflood of a drained pool for a region of the pool storing higher reactivity fuel 

assemblies where the boron poison in the rack panels has been significantly displaced as a 

result of the earthquake). If such an event affected a region of the pool (as opposed to only a 

portion of a particular assembly), and if it occurred at a point in the accident where the fuel was 

only partially covered, the event could have an important impact on onsite dose rates. Further, 

if an ICE were severe enough to produce significant heat, the fuel will be harder to cool and 

short-lived radionuclides will be produced. Design requirements and safety analyses ensure 

that the spent fuel stored in the pool, under normal conditions, will not result in a critical 

configuration. For the reference plant (and other U.S. SFPs), this is ensured through a 

combination of assembly spacing and neutron poison material (e.g., Boraflex). If a seismic 

event did cause reconfiguration of the fissionable material by means of either (1) direct 

movement of the fuel, (2) direct degradation of the poison material, or (3) indirect effect on 

either the fuel or poison material because of high temperatures associated with an induced 
30

accident, there are several “advantageous” considerations, including the following:

• The reactivity of fresh BWR fuel is suppressed by the high content of burnable 

absorbers.

• The majority of the fuel in the SFP has low net reactivity since it has gone through more 

than one operating cycle in the reactor.

• The fuel with the highest net reactivity will likely be the once-burned assemblies which 

will stay in the reactor during a “shuffling” refueling outage (but would not stay in the 

reactor for a full core offload).

• Critical configurations of low-enriched uranium fuel require the presence of a neutron 

moderator (in this case water or steam) such that an ICE would not happen in the 

presence of air.

• BWR SFPs do not use borated water so the fact that the SFP may be refilled with 

unborated water is not a deviation from the norm.

• Low-enriched uranium fuel assemblies (which are used in all U.S. reactors) are generally 

geometrically designed to maximize reactivity (moderator/fuel geometry) in the reactor 

and so any significant alteration of the geometry of a given assembly will likely be in the 

direction of a less criticality-prone configuration.

There are also a few counter considerations:

• The poison material in the rack panels contribute significantly to the net reactivity of the 

SFP configuration (i.e., they are a key component to ensuring subcriticality for high 

reactivity assemblies).

• The effects of large seismic events on already degraded SFP rack poison material are 

not easy to quantify.

• The rack panels and poison material have a lower melting temperature than the cladding 

and fuel.

• Termination of a SFP ICE during an event that required deployment of mitigation 

equipment could be difficult.

• The possibility of a criticality event cannot be summarily dismissed.

Finally, the offsite consequences of a criticality event (especially if it occurs when overlying 

water is present) are believed to be less severe from a public health and safety standpoint than 

the offsite consequences from a potential large release of radioactive material associated with a 

prolonged uncovering of the fuel in the SFP resulting from not attempting to reflood. In 

consideration of all of the above, common accident management practices in the United States

call for the use of any available water in responding to fuel uncovery in either the reactor or 

SFP. This study shows the precedent, while recommending that future work be done to better 

understand the specific combinations of conditions that could lead to ICEs during a large 

seismic event.
31

3. SEISMIC HAZARD CHARACTERIZATION

3.1 Basis for Probabilistic Estimates

The primary sources of information for seismic hazard estimates at nuclear power plant sites 

include (1) the NRC/ LLNL (Bernreuter et al., 1989; Sobel, 1994) model (hereafter referred to as 

the LLNL model); (2) the EPRI model (Toro et al., 1989); and (3) the USGS model developed in 

the mid-2000s (Peterson et al., 2008) (hereafter referred to as the USGS 2008 model. The 

implementation of the individual plant evaluation for external events (IPEEE) program utilized 

both the LLNL and EPRI models (NRC, 2002b). The National Seismic Hazard Mapping Project 

utilized the USGS 2008 model. The NRC also utilized the USGS 2008 model for the seismic 

hazards estimates used in screening level assessments for Generic Issue 199 (GI-199) (NRC, 

2012a).

The seismic hazard assessment in this study is the US Geological Survey (USGS, 2008) hazard

model. A new probabilistic seismic hazard model is currently being developed and will consist 

of two parts: (1) a seismic source zone characterization and (2) a ground motion prediction 

equation (GMPE) model. Although part (1) is now complete (NRC, 2012b), it was not available 

at the start of this scoping study. In addition, the GMPE update is still in progress. Furthermore, 

the NRC is currently developing an independent probabilistic seismic hazard assessment 

(PSHA) computer code to incorporate part (1) and part (2) when complete. While the USGS 

(2008) hazard model is not sufficiently detailed for regulatory decisions, it is appropriate to use 

for this study because it was the most recent and readily available hazard model for the 

selected site at the start of the study. 

Figure 2 (PGA) and Figure 3 (1-, 5-, and 10 Hz spectral acceleration) graphically show 

comparisons of hazard estimates for the reference site (a rock site) with the three information 

sources listed above. These comparisons are provided to compare the model used in this study 

to well-known and extensively documented information sources (LLNL model and EPRI model) 

that were used in past SFP risk studies. The comparisons support the following observations:

• For the PGA, the USGS 2008 model predicts higher annual probability of occurrence for 

high-level, low-probability events, specifically for events with PGAs greater than about 

0.35g.

• For moderate PGAs, from about 0.1g to 0.35g, the LLNL model is higher than the USGS 

2008 model. For events above about 0.35g, which are lower probability events, the 

USGS 2008 model is higher than the LLNL model until both hazard estimates differ by 

factors of about 2.5 at 0.75g and about 3 at 1.0g.

• The EPRI model hazard estimates are lower than those from the USGS 2008 model for 

all PGAs. Specifically, hazard estimates based on the USGS 2008 model are about 2 

times greater at 0.2g with the difference increasing to about 10 times at 1.0g.

• Thus, in terms of PGA, the seismic hazard estimates used for this study are about 2.5 

times greater than LLNL model estimates and about 6 times greater than EPRI model 

estimates at 0.75g.
32

• Curves for the USGS 2008 model and the LLNL model are comparable for each 

representation, with the USGS 2008 model sometimes being higher (higher annual 

probability of occurrence) and the LLNL model sometimes being higher.

• Generally, the 10-Hz curve is the highest, followed by the 5-Hz curve, followed by the 

PGA curve, followed by the 1-Hz curve. The notable exception to this is the fact that the 

5-Hz LLNL model curve, which is higher than the 10-Hz curve.

Figure 2 Comparison of PGA exceedance frequencies at the reference plant

Figure 3 Comparison of spectral exceedance frequencies at the reference plant (rock 

hazard curves)

1.E-07

1.E-06

1.E-05

1.E-04

1.E-03

0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

Frequency of Exceedance (/yr)

Acceleration (g)

USGS 2008 PGA

IPEEE LLNL PGA

IPEEE EPRI PGA

1.E-07

1.E-06

1.E-05

1.E-04

1.E-03

0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

Frequency of Exceedance (/yr)

Acceleration (g)

USGS 2008 @ 10 Hz

IPEEE LLNL @ 10 Hz

IPEEE EPRI @ 10 Hz

USGS 2008 @ 5 Hz

IPEEE LLNL @ 5 Hz

IPEEE EPRI @ 5 Hz

USGS 2008 @ 1 Hz

IPEEE LLNL @ 1 Hz

IPEEE EPRI @ 1 Hz
33

A comparison of the annual frequency of exceeding a given PGA for all Mark I sites (Figure 4) 

shows that the reference plant falls close to the upper end of the group in terms of hazard 

estimates. When comparing the annual frequency of exceeding a given 1-Hz spectral 

acceleration (Figure 5), the reference site is in the upper half of the group.

Figure 4 Comparison of annual PGA exceedance frequencies for U.S. Mark I reactors

(USGS 2008 model) (rock hazard curves)

Figure 5 Comparison of annual exceedance frequencies for 1 Hz spectral accelerations 

for U.S. Mark I reactors (USGS 2008 model) (rock hazard curves)
34

3.2 Characterization of the Event Likelihood

Hazard exceedance frequencies can be translated into initiating event frequencies by 

partitioning the PGA range into a number of discrete categories (bins) defined in terms of PGA 

intervals. These bins define a discrete number of seismic event scenarios with increasing 

intensity (PGA). Revision 1.01 of the NRC handbook entitled, “Risk Assessment of Operational 

Events, Volume 2—External Events,” issued January 2008 (NRC, 2008b), recommends the use 

of at least three bins unless plant-specific considerations require more bins. The SFPS used 

four bins.

Table 4 shows the resulting bins, along with the tabulated frequencies for various spectral and 

peak accelerations. Note that for bin 4, the representative bin PGA has been set to 1.2g by 

convention, whereas for the other bins, it is the geometric mean of the interval endpoints. 

Figure 6 shows these results graphically.

Table 4 Seismic Bins and Initiating Event Frequencies

Bin # Bin Range 

(g)

Bin PGA (g) Approximate Initiating Event 

Frequency (USGS 2008 model) (/yr)

1 0.05 - 0.3 0.12 5.2x10-4

2 0.3 - 0.5 0.4 2.7x10-5

3 0.5 - 1.0 0.7 1.7x10-5

4 > 1.0 1.21 4.9x10-6

1

 Assumed based on PRA modeling convention

Figure 6 Comparison of seismic initiating event frequencies

Based on this information, and on a review of the results of past studies which indicate that 

damage to the SFP and other relevant structures, systems, and components (SSCs) is not 

credible for events in bins 1 and 2, this study focused on bin 3. The project staff concluded that 

seismic bin 3 provides the best compromise between events with higher occurrence frequencies 

that would lead to little or no damage versus higher consequence events with very low 

frequencies. Review of past studies (e.g., NUREG-1738 (NRC, 2001)) indicates that events in 

1.E-07

1.E-06

1.E-05

1.E-04

1.E-03

1 2 3 4

Initiating Event Frequency (/yr)

Seismic Bin Number

Bin 1: 0.05 - 0.3g Bin 2: 0.3 - 0.5g

Bin 3: 0.5 - 1g Bin 4: > 1g

PGA
35

bin 3, with initiating annual frequencies on the order of 1×10-5 to 2×10-5

, could challenge the 

integrity of the SFP (i.e., of causing a leak) at the reference plant. Thus, this is the initiating 

earthquake chosen for this study. It is an event that is no more severe than events considered 

in past reactor and SFP PRAs and consequence studies.

This study therefore considers a challenging, but very low probability earthquake as the initiating 

event, selected based on the considerations indicated above. This decision translates into a 

seismic event with a PGA several times greater than that associated with the design-basis 

earthquake, currently called the safe-shutdown earthquake or SSE. The PGA for the reference 

plant SSE is 0.12g. (This is about a magnitude 5.3 earthquake at about 25 kilometers (km).) 

While the probability of occurrence of this earthquake was not used in its selection, the annual 

probability of occurrence for this PGA is about 1.8×10-4 (approximately one event in 5,500 

years) when calculated using the EPRI and USGS 2008 models and about 3.2×10-4

(approximately one event in 3,200 years) when calculated using the LLNL model. An initial 

determination, largely based on the results of past studies (NRC, 2001; Prassinos et al., 1989) 

and engineering judgment, was that the ground motions associated with the SSE (bin 1) would 

not be large enough to damage the SFP at the reference plant.

The information above coupled with the review of previous studies (NRC, 2001) suggests that 

the frequency of a seismic event that could challenge the integrity of the SFP at the reference 

plant is on the order of 1.7×10-5 per year (i.e., approximately one event in 60,000 years) or less. 

Table 5 contrasts this frequency against other sources of information. The Mineral, VA, 

earthquake of August 23, 2011, which occurred near the North Anna nuclear power plant, can 

serve as a point of reference. In that case, the NRC staff concluded, using data from USGS 

instruments, that the PGA at the North Anna site was about 0.26g (NRC, 2011b). Using the 

USGS 2008 information for North Anna, the hazard frequency for an event with this PGA is 

about 1.2×10-4 per year (one event every 8,300 years). This frequency places the Mineral, VA, 

event in bin 1.

Table 5 Comparison of Seismic Frequencies from Various Sources

Source Estimated initiating event frequency 

of a large seismic event1

Notes

USGS 2008 1.7x10-5

/yr

(one event in 60,000 years)

Frequency of seismic bin 3 of 

4 (0.5 to 1g)

The reference plant’s 

SPAR-EE Model 

(v3.21, Rev. 1)

1.3x10-5/yr

(one event in 77,000 years)

Frequency of seismic bin 3 of 

3 (> 0.5g)

NUREG-17381 1.1x10-5

/yr

(one event in 90,000 years)

Frequency of seismic hazard 

between 0.51g to 1.02g 1 Initiating event frequencies reported are those based on the LLNL models (Sobel, 1994).

In addition to the PGA, ground motions at a site are also characterized by their frequency 

content expressed in terms of response spectra. Section 3.3 describes the procedure used to 

develop the horizontal and vertical acceleration response spectra for the input ground motion for 

this study. 

Other response spectra of interest for this study are (1) the plant’s SSE response spectra and 

(2) the free-field response spectra used in the seismic PRA for the NUREG-1150 study. These 

spectra are of interest for comparison purposes. The spectra in the NUREG-1150 study are 

also of interest because in-structure response spectra calculated for those ground motions were 
36

scaled (see Section 4), in approximation, to estimate in-structure response spectra for the input 

free-field ground motion used in this study. Volume 1, Part 3, of NUREG/CR-4550, “Analysis of 

Core Damage Frequency: Peach Bottom, Unit 2 External Events,” issued December 1990 

(Lambright et al., 1990), provides the horizontal and vertical free-field response spectrum used 

in the NUREG-1150 seismic PRA for Peach Bottom in terms of the median spectral ordinates 

for various values of the PGA. As shown in Section 3.3, the spectral shape for this study differs 

from the SSE response spectrum, as well as from the median response spectra considered in 

the NUREG-1150 seismic PRA. Frequency content for the SSE and the NUREG-1150 PRA 

spectra generally resemble each other.

3.3 Characterization of the Ground Motion Response Spectra

Spectral shapes developed for the safety/risk assessment results for the GI-199, which utilized 

the USGS (2008) model, were used to develop the free-field acceleration response spectra for 

this study. The free-field acceleration response spectrum developed for the GI-199 for this site 

has a zero-period spectral acceleration (PGA) of about 0.34g. The acceleration response 

spectra for the free-field ground motion for the initiating seismic event considered for this study 

(bin 3 in Table 4 and a PGA of 0.7 g) were derived from the GI-199 spectra shape as follows: 

• Horizontal shaking: horizontal response spectrum is the GI-199 spectral shape scaled to 

the bin 3 PGA (zero-period acceleration) of about 0.7g (specifically 0.71g). While it is 

recognized that the frequency content of ground motions may change somewhat with 

increasing PGA levels, scaling of the spectral shape for the 0.34g PGA to the bin 3 PGA 

is considered a reasonable approximation for low probability hazard for this rock site and 

for the purposes of this study. Figure 7 compares the horizontal input acceleration 

response spectra for this study to the horizontal response spectrum for the plant’s Safe 

Shutdown Earthquake (SSE) (PGA of 0.12g) for 5-percent damping.

Figure 7 Input acceleration response spectrum and SSE (Horizontal Ground Motion)

• Vertical shaking: vertical spectral accelerations and the vertical PGA (0.7 g) are 

assumed to be the same as the horizontal spectral accelerations and PGA. A few 

studies (e.g., McGuire, Silva, and Costantino, 2001; ASCE, 1999) indicate that for rock 

sites and frequencies near and above 10 Hz, and especially nearby seismic sources, 

0

0.2

0.4

0.6

0.8

1

1.2

1.4

1.6

1.8

2

0.1 1 10 100

Acceleration (g)

Frequency (Hertz)

Ground Motion Acceleration Response Spectra - 5% Damping

SSE

SFP Study
37

vertical spectral accelerations may be as high as or exceed horizontal spectral 

accelerations. For this study, the frequencies of interest are, for the most part, 

frequencies near or above 10 Hz. Therefore, the assumption of equal vertical and 

horizontal spectral accelerations was deemed to be a reasonable starting assumption. 

This assumption is also supported by seismic hazard de-aggregation with the USGS 

2008 model which indicates that for the seismic bins of interest (high PGA, low likelihood 

events) the contributors to the hazard would be earthquakes with magnitudes less than 6 

at about 20 km from the site. 

Other response spectra of interest for this study are the free-field response spectra used in the 

seismic PRA for the NUREG-1150 study (Lambright et al., 1990). These spectra are of interest 

because in-structure response spectra calculated for that ground motion were scaled, in 

approximation, to estimate in-structure response spectra for the free-field ground motion 

considered for this study. Figure 8 compares the frequency content of the horizontal response 

spectra (5-percent damping) for the SSE, the median response spectrum used in the NUREG1150 study, and the spectral shape used in Spent Fuel Pool Study. For this comparison, all 

spectra are scaled to a PGA of 1.0g. When the three response spectra under consideration are 

scaled to the same PGA, the information in Figure 8 supports the following observations:

• For frequencies between about 10 Hz and 45 Hz, the spectral shape used in this study

has higher spectral accelerations than the ground shaking considered for the SSE and 

for the NUREG-1150 study.

• For frequencies between about 0.5 Hz and 10 Hz, which is generally the frequency 

range most damaging for nuclear power plant structures, the spectral shape used in this 

study has lower spectral accelerations than the ground shaking considered for the SSE 

and the NUREG-1150 study. 

Figure 8 Response spectrum for 5-percent damping scaled to 1.0 g PGA: SSE, 

NUREG/CR-4550 (NUREG-1150 PRA), and this study (GI-199)

As noted above, the input horizontal acceleration response spectrum for the event considered in 

this study is the spectral shape derived for the GI-199 study using the USGS 2008 model (PGA 

of about 0.34g) scaled to a PGA of about 0.7g. Figure 9 shows the horizontal response spectra 

(5-percent damping) for the event considered in this study, for the SSE (0.12g PGA), and for the 

0.00

0.50

1.00

1.50

2.00

2.50

3.00

0.1 1 10 100

Acceleration (g)

Frequency (Hertz)

Ground Motion Acceleration Response Spectra - Scaled to 1.0 g PGA - 5% Damping

SSE

SFP Study

NUREG-1150
38

response spectrum used in the NUREG-1150 PRA. The NUREG-1150 response spectra 

shown in the figure is scaled from a PGA of 0.3g to the PGA of the event for this study. Figure 

10 illustrates how the ground motions considered in this study are considerably more 

challenging than those for the SSE. However, these ground motions are also significantly less 

likely as indicated in Table 4. They are also richer at the high frequencies (greater than 10 Hz), 

which generally tend to be less challenging to nuclear power plant structures.

Figure 9 Horizontal response spectrum (5-percent damping): SSE, SFP Study and 

NUREG-1150 PRA (for 0.71g PGA)

0

0.2

0.4

0.6

0.8

1

1.2

1.4

1.6

1.8

2

0.1 1 10 100

Acceleration (g)

Frequency (Hertz)

Ground Motion Acceleration Response Spectra - 5% Damping

SSE

SFP Study

NUREG-1150 (0.71 g PGA)
39

4. STRUCTURAL ANALYSIS AND RELATED INITIAL DAMAGE 

CHARACTERIZATION

This section documents the structural analysis performed to estimate the initial damage states 

for the accident progression analysis. It provides:

• the objective,

• the approach including assumptions, 

• the structural modeling and analyses, and 

• the potential damage states and their relative likelihoods for the seismic event 

considered. 

The objective of the structural assessments was to provide damage states that might result from 

the seismic event described in the previous section and that would constitute the initial 

conditions for the accident progression analysis. Structural and related damage states have 

been divided into the following two major categories:

(1) structural damage to the spent fuel pool structure with potential locations of leakage from 

concrete cracking and related liner tearing

(2) other damage states including:

- amount of water, if any, displaced by sloshing of the water out of the SFP

- damage to the refuel gate, support systems and penetrations, as well as

qualitative assessment of damage to spent fuel racks and spent fuel assemblies

- damage to the reactor building and other relevant structures.

Most of the analytical effort focused on assessing potential structural damage to the spent fuel

pool structure, namely concrete distortions, concrete cracking, and metallic liner strains. 

Section 4.3 provides a review of the performance of SFPs at four nuclear power plant sites with 

a total of 20 reactors under two major recent earthquakes in Japan. This review compares 

relevant aspects of the seismic scenario and estimated damage states for this study with known 

aspects of the seismic scenario and performance of SFPs affected by those earthquakes. The 

review summarizes known or presumed reductions in water levels of the SFPs affected by those 

earthquakes associated with either water leakage from structural damage, if any, or water loss 

from sloshing. Although these reviews and comparisons use information available at the time of 

the execution of this study, they assist in the interpretation of the results obtained for the seismic 

scenario and SFP considered.

4.1 Damage States for the Spent Fuel Pool Structure

4.1.1 Approach and Seismic Loads

The general approach for the estimate of the damage states follows the approach reported in 

NUREG/CR-5176, “Seismic Failure and Cask Drop Analyses of the Spent Fuel Pools at Two 

Representative Power Plants,” issued January 1989 (Prassinos et al., 1989) modified to 
40

address specific needs of this study. (The general approach is fully described in the following 

section.) The analyses reported in NUREG/CR-5176 were conducted in conjunction with 

research activities related to Generic Issue 82 (GI-82) (NRC, 2012c). Appendix 2 to NUREG1738 (NRC, 2001), a technical study of spent fuel pool accident risk at decommissioning nuclear 

power plants, also addresses the seismic fragility of spent fuel pools and refers to the results 

and approach used in NUREG/CR-5176. The seismic evaluations in NUREG/CR-5176 

considered ground motions with frequency content that differs from those considered in this 

study. Specifically, NUREG/CR-5176 considered ground motions with maximum response 

spectra amplitudes for frequencies below 10 Hz while the ground motions considered for this 

study have maximum response spectra amplitudes for frequencies greater than 10 Hz. This 

difference in the characteristics of the ground motions tends to induce conservatism in the 

approach when applied to this study as indicated below. 

Approach

The overall approach used to assess damage to the SFP structure, namely concrete cracking, 

concrete distortions, liner strains and liner tearing, for the earthquake event considered, consists 

of the following nine steps:

(1) Obtain free-field acceleration response spectra (horizontal and vertical) for the site 

considered (a rock site and a reactor building with small embedment) as indicated in 

Section 3.3 and shown in Figure 7.

(2) From reliable and well-documented past studies, obtain in-structure response spectra 

(ISRS) (also called floor response spectra) for the vertical and horizontal directions at 

the elevation of the base of the SFP (Elevation 195 ft). (For reference, the elevation at 

the top of the refueling floor is Elevation 234 ft and the elevation at the top of the 

foundation slab is Elevation 92 ft 6 in.) The SFP Study used the median-centered ISRS 

calculated for Peach Bottom for the seismic PRA performed for the NUREG-1150 study 

(NRC, 1990) and reported in Volume 4, Part 1, Revision 1 of NUREG/CR-4550 

(Lambright et al., 1990). 

(3) Estimate ISRS for the ground motions of interest for this study at the elevations of 

interest (Step 2) by scaling the ISRS from previous studies (Step 2). The scaling 

accounts for differences in the response spectra for the NUREG-1150 seismic PRA and 

for this study. Given the differences in the ground motions for the NUREG-1150 PRA 

and for this study, the use of this scaling is likely to be near the limit of acceptability. Use 

of this scaling is justified on the basis that the approximations and uncertainties 

introduced are consistent with the uncertainties in other approximations used in the 

structural and seismic assessments for the study. It is expected that efforts by the NRC 

and industry related to Requests for Information in SECY-12-0025 (NRC, 2012f) 

associated with the Near Term Task Force (NTTF) Recommendation 2.1 (NRC, 2011a) 

will result in updated staff guidance on ISRS scaling.

(4) Use the scaled ISRS from Step 3 to estimate equivalent static forces to be applied, in 

conjunction with dead loads, to the floor and walls of the SFP as input for a static 

nonlinear pushover analysis. These equivalent static forces account for (1) peak vertical 

and horizontal accelerations of the floor and walls of the SFP structure (seismic 

coefficients), (2) peak vertical and horizontal hydrodynamic impulsive pressures on the 

floor and walls of the SFP from the water in the pool and (3) vertical dynamic forces on 

the SFP floor from the dynamic response of the racks and spent fuel assemblies. At this 
41

stage of the analysis also estimate vertical displacement of the water surface from 

sloshing.

(a) Use a simplified three-dimensional (3D) finite element model of the SFP structure 

to estimate or verify these loads. Specifically, use elastic solid elements and 

special fluid elements to model the water to estimate natural frequencies and 

mode shapes for the SFP structure. Use this model to calculate the spatial 

distribution of peak vertical and horizontal accelerations of the structural 

components using the ISRS from Step 3 as input.

(b) Calculate hydrodynamic impulsive pressures and peak vertical and horizontal 

pressures on the basis of simplified methods (Housner, 1963; AEC, 1963; 

Malhotra et al., 2000). Use the 3D finite element model in Step 4a together with 

the ISRS from Step 3 as input to estimate peak hydrodynamic pressures. This 

provides for verification and adjustment of the hydrodynamic pressures 

calculated using simplified methods. 

(c) Use the 3D finite element model in Step 4a together with the ISRS from Step 3 

as input to estimate vertical displacements of the water surface from sloshing. 

The estimated displacements were small when compared to the depth of water in 

the SFP as indicated below. 

(5) Perform a 3D static nonlinear pushover analysis of the SFP structure using a detailed 3D 

finite element model of the SFP structure that includes nonlinear modeling of concrete 

including cracking as well as modeling of the steel reinforcement, embedded steel floor 

beams and the SFP liner. Such analysis provides the load deformation behavior of the 

SFP for the loading pattern and intensity considered. Perform the static nonlinear 

pushover analysis for adequate combinations of the vertical and horizontal ground 

motions to account for the fact that maximum vertical and horizontal accelerations do not 

occur simultaneously (NRC, 2006a). Accordingly, perform the nonlinear static pushover 

analysis as follows:

(a) Incrementally apply the dead loads to the SFP structure to calculate initial 

stresses and strains. Dead loads considered for this study consist of: the weight 

of the pool structural components, weight of water, weight of the spent fuel 

assemblies and weight of the spent fuel racks.

(b) Follow Step 4a with an incremental application of adequate combinations of the 

horizontal and vertical equivalent static forces estimated in Step 4. The 

incremental application is needed to track development and effects of concrete 

cracking, concrete strains, steel yielding and liner strains. 

(c) Based on guidance for combining effects from three spatial components of an 

earthquake in Regulatory Guide 1.92 (NRC, 2006a), peak vertical seismic loads 

were combined with 40-percent of the peak horizontal loads. A combination of 

peak horizontal loads on both directions with 40-percent of the vertical loads was 

also considered. Preliminary analyses indicated that the load combination 

involving peak vertical loads and 40-percent of the horizontal loads would likely 

be the most severe combination for the SFP structure analyzed. Accordingly, this 

was the combination studied in more detail in the remainder of the study. 

(d) Use best-estimate median material properties for all materials (e.g., concrete, 

reinforcement, steel beams and liner) based on best available information for 

similar materials used in nuclear plants and other guidance for the assessment of 

beyond-design-basis events and for seismic fragility assessments. Also take into 
42

account the effect of aging on the concrete strength as recommended for the 

assessment of beyond-design-basis events (NEI, 2011; Prassinos et al., 1989). 

(6) Review and process the calculated structural distortions (as measured by the 

displacement of nearby nodes), structural deformations, concrete strains and liner 

strains for the following purposes:

(a) Assess the possible development of cracks through the floor or walls (the 

analyses indicated that critical concrete cracking such as this would only develop 

at the base of the walls along the intersection of the SFP walls with the SFP 

floor) and then estimate crack lengths and average crack width.

(b) Assess liner strains at the intersection of the base of the walls and floor slab in 

order to assess the potential for liner tearing. Take into consideration details of 

the attachment of the liner, in discrete locations, to the concrete floor and walls.

(7) Define three initial states for the subsequent accident progression analysis as follows:

(a) A state with no leakage, and no loss of coolant, from the bottom of the SFP. This 

state corresponds to concrete cracking at the base of the walls (estimated to be 

through-wall cracking for the event considered as shown in subsequent 

subsections) but without tearing of the liner.

(b) A state with moderate leakage rate from the bottom of the SFP, corresponding to 

through-wall concrete cracking at the bottom of the walls with tearing of the liner 

that propagates to an extent such that water leakage is controlled by the size of 

the cracks in the concrete. 

(c) A state with small leakage rate from the bottom of the SFP, corresponding to 

through-wall concrete cracking at the bottom of the walls and tearing of the liner 

that remains localized such that water leakage is controlled by the size of the 

tearing in the liner. 

(8) For the two damage states with leakage, estimate the leakage rate at the base of the 

SFP walls. When the rate is controlled by the cracking in the concrete (moderate 

leakage rate) use recent large scale test data for the flow of water through thick concrete 

slabs together with the estimated crack width and length to estimate the leakage rate. 

When the rate is controlled by localized liner tearing, use empirical date from leakage 

through cracks in steel pipes to estimate the leakage rates.

(9) Use data for ultimate strains in the types of steel used for SFP liners, together with 

uncertainties in the calculated liner strains as well as uncertainties in the estimation of 

the in-structure loads and concrete properties to estimate the relative likelihoods for the 

three initial damage states listed in Step 8 - no leakage, moderate leakage rate and 

small leakage rate. 

As noted above, this approach parallels part of the approach used in conjunction with the 

resolution of GI-82 (Prassinos et al., 1989). It augments the earlier approach in that it uses 

modern finite element methods in Steps 4 and 5. The use of finite element analyses in Step 4 is 

done to obtain a more accurate assessment of the natural frequencies of the SFP structure 

itself, to estimate the spatial distribution of seismic coefficients and to verify and adjust 

hydrodynamic impulsive loads on the floor and walls of the SFP. The use of finite element 

analyses in Step 5 is done to track the development of cracking and liner strains and then relate 

those to damage states, leakage rates and their relative likelihoods. 
43

The approach described above has potentially conservative aspects that may overestimate the 

damage to the SFP structure. These conservative aspects are as follows:

• As the high-frequency structure of the SFP (fundamental frequency on the order of 15 to 

25 Hz) cracks under the applied seismic loads, its natural frequencies decrease and are 

no longer resonant with the high frequency components of the ground motion (i.e., the 

frequencies corresponding to the higher spectral accelerations). Since the spectral 

accelerations decrease as the frequencies of the SFP structure decrease after cracking, 

the use of seismic loads calculated assuming elastic frequencies can introduce 

conservatism in the analysis for the seismic event considered. This would be the case if 

the SFP structure were to remain stable with only minor distortions after cracking as in 

the case of the SFP studied. This aspect was partially accounted for in this study 

through a small reduction in the spectral accelerations and by the use of a small 

reduction of the concrete stiffness in the calculation of the natural frequencies of the SFP 

structure. Assessment of the conservatism introduced by the approach used, which was 

outside the scope of this study, would involve the sampling of representative 

acceleration time-histories, both vertical and horizontal, and their use in time-history 

analyses of the SFP response to the seismic loads considered. 

• Generally ISRS accelerations do not increase proportionally (linearly) from low PGA 

events to an event with a PGA as high as that considered in this study. As the load 

increases, both the structure of the reactor building and of the SFP may crack and 

dissipate energy thus dampening the response of the building. This effect is taken into 

account, in part, by reducing ISRS spectral accelerations by the ratio of spectral 

amplification factors for 10-percent and 5-percent damping (Newmark and Hall, 1978).

The reduction of spectral accelerations implied by the use of a higher damping ratio is 

further justified by the decrease in the fundamental frequency of the structure related to 

cracking which, for the input free-field ground motion for this study (see Section 3.3 and 

Figure 9), would decrease the spectral accelerations. 

• Another potentially conservative aspect of the analysis is that the scaling of the ISRS 

does not take into account reductions on the high-frequency (greater than 10 Hz) 

spectral accelerations that may result, under some circumstances, from ground motion 

incoherency, wave scattering, soil-structure interaction effects and wave passage 

effects. This is accounted for, in part, in the calculation of the relative likelihood of the 

various damage states by considering a small range of reduction in the response and 

associated uncertainties, as discussed below.

Other approximations of note include (1) the scaling of ISRS calculated from a ground motion 

with response spectra markedly different from the ground motion spectra considered for this 

study (addressed in item 5(c)), (2) the decoupling of the response of the SFP from the response 

of the reactor building, and (3) neglecting the small embedment of the reactor building, as also 

done in previous studies, which may affect the calculation of horizontal ISRS. Follow-on 

subsections address these and the above approximations. More detailed approaches involving

the use of sampled time-histories, including sampling of incoherent ground motions, used in 

conjunction with 3D models of the entire reactor building and soil-structure interaction analysis, 

to calculate loads on the SFP would provide a better assessment of these possible 

conservatisms. However, these were outside the scoping nature of the study. 

The weight of other SFP equipment and appurtenances on the dead loads, and thermal 

stresses are not accounted for explicitly in the estimation of the initial stresses in the SFP 

components (Step 5a). The weight of those equipment and appurtenances is expected to be 
44

small in comparison to the other dead loads in the pool and accounted for by the 

approximations in the estimation of those dead loads. Thermal stresses are not accounted for 

under the assumption that concrete cracking will relieve the thermal stresses. Moreover, 

increase in the temperature of the water, if it were to occur, would not happen until several 

hours after the termination of the ground shaking.

Seismic Loads

Chapter 3 of this report discusses the bases for the free-field ground motion response spectra 

for the seismic event considered for the SFP Study. As noted in Chapter 3 other free-field 

response spectra of interest in this study are those documented in NUREG/CR-4550 (Lambright

et al., 1990) and used in the probabilistic risk assessment (PRA) for NUREG-1150. That report 

provides median-centered ISRS (for 5-percent damping) for the Peach Bottom reactor 

structures calculated using time-history analysis and an ensemble of free-field ground motion 

time-histories. Section 3 provides a comparison of the median-centered free-field ground motion 

response spectra for that ensemble of time-histories to the ground motion response spectrum 

for the seismic event considered in this study.

The free-field response spectra and ISRS reported in Lambright et al. (1990) form a set of 

reliable and well-documented response spectra for the Peach Bottom reactor buildings. 

Specifically, that report provides ISRS at various elevations of interest in the reactor building, 

namely at the bottom elevation of the SFP (Elevation 195 ft) and at the refueling floor (Elevation 

234 ft). In addition, the report also provides estimates of natural frequencies of vibration for the 

reactor building, which are listed in Table 6. These frequencies help understand the shape of 

the ISRS for the Peach Bottom reactor building. It is noted that the dominant, elastic 

(uncracked) frequencies of vibration of the SFP structure, considering hydrodynamic effects of 

the water and the mass of the spent fuel, range from about 17 Hz (vertical response of the floor 

slab) to 28 Hz (horizontal deformations of the walls). These frequencies are remote (detuned) 

from the frequencies for the horizontal mode of vibration for the reactor building but are close to 

its vertical frequency.

Table 6 Estimated Natural Frequencies of Vibration for the Peach Bottom Reactor 

Building (Lambright et al., 1990)

Direction Frequency (Hz) % Mass

Horizontal (NS) 7.1 68

Horizontal (EW) 7.6 71

Vertical 18.5 72

Using simplified scaling procedures, the ISRS in Lambright et al. (1990) were scaled to estimate 

floor vertical and horizontal ISRS at the elevation at the bottom of the SFP (Elevation 195 ft) as 

well as horizontal ISRS at the midheight of the SFP walls (by averaging scaled spectra at 

Elevation 195 ft and Elevation 234 ft). The scaling was done by estimating the ground motion 

amplification factors from the ground motion to the ISRS and then applying those factors to the 

response spectra for the SFP study. This scaling was done using the reported median-centered 

ISRS for 5-percent damping, the vertical and horizontal (EW) components for the ISRS 

(examination of the charts indicates that the horizontal (EW) component tends to have the 

higher spectral accelerations). The SFP Study considered identical horizontal ISRS for both

directions. Note that for the SFP studied the horizontal components of the ground motion are 

not those with the greatest damage potential. Justification for not considering reductions in the 

high frequency spectral accelerations is provided at the end of this subsection.
45

Figure 10 shows a comparison of the vertical ISRS for the elevation at the bottom of the SFP, 

calculated as indicated above to the corresponding ISRS (smoothed) for the NUREG-1150 

seismic PRA. Likewise, Figure 11 provides a similar comparison for the horizontal ISRS at the 

midheight of the SFP structure (average of the ISRS at Elevation 195 ft and Elevation 234 ft). 

Figure 10 Vertical ISRS for 5-percent damping at Elevation 195 ft (bottom of the SFP)

Figure 11 Horizontal ISRS for 5-percent damping midway between Elevation 195 ft and 

Elevation 234 ft (midheight of the SFP)

The spectra shown in Figure 10 and Figure 11 are for 5-percent damping (reactor building and 

equipment). Calculation of seismic load coefficients for the SFP floors and walls as well as of 

hydrodynamic impulsive pressures considered a reduction of these spectral accelerations. 

Specifically, seismic coefficients and hydrodynamic pressures calculated using the 5-percent 

damping ISRS were reduced by the ratio of scaling factors for 10-percent and 5-percent 

damping reported in NUREG/CR-0098 (Newmark and Hall, 1978). As noted above, this is done 

to account for, in part, the energy dissipation (damping) from cracking of the SFP and minor 

0.00

0.50

1.00

1.50

2.00

2.50

0.1 1 10 100

Acceleration (g)

Frequency (Hertz)

In-structure Spectra - Vertical - Scaled to 0.71 g PGA (free-field) - 5% Damping 

From NUREG/CR-4550

This study (based on 2008-USGS)

0.00

0.50

1.00

1.50

2.00

2.50

3.00

3.50

4.00

0.1 1 10 100

Acceleration (g)

Frequency (Hertz)

In-structure Spectra - Horizontal - Scaled to 0.71 g PGA (free-field) - 5% Damping 

From NUREG/CR-4550

This study
46

cracking of the reactor building. This is further justified by the reduction in the natural frequency 

of the SFP structure from cracking that would lead to reduced spectral accelerations for the 

input free-field ground motion (see Section 3.3 and Figure 9). An assumption is, for example, 

that for the intense ground motion of the event considered, the reactor building will undergo 

more cracking than that estimated for the design basis motion (SSE). This will absorb and 

dissipate energy and damp the response. ISRS obtained by reducing the 5-percent damping 

ISRS in this manner, herein called 10-percent damping ISRS, are shown in Figure 12 and 

Figure 13.

Figure 12 Vertical ISRS for 5-percent and 10-percent damping at Elevation 195 ft (bottom 

of the SFP)

Figure 13 Horizontal ISRS for 5-percent and 10-percent damping midway between 

Elevation 195 ft and Elevation 234 ft (midheight of the SFP)

The scaling used to obtain the 5-percent damping ISRS does not take into account reductions 

on spectral accelerations for frequencies greater than 10 Hz that would result, under some 

circumstances, from ground motion incoherency, wave scattering, soil-structure interaction 

effects and wave passage effects. The plant dimensions of the reactor building are about 150 ft 

by 120 ft above Elevation 135 ft (ground elevation) and about 150 ft by 150 ft below Elevation 

135 ft. The building foundation consists of a 4 ft 4 in. reinforced concrete (RC) slab lying on top 

0.00

0.50

1.00

1.50

2.00

2.50

0.1 1 10 100

Acceleration (g)

Frequency (Hertz)

In-structure Spectra - Vertical - Scaled to 0.71 g PGA (free-field) - 5% Damping 

From NUREG/CR-4550

This study

This study - 10% damping

0.00

0.50

1.00

1.50

2.00

2.50

3.00

3.50

4.00

0.1 1 10 100

Acceleration (g)

Frequency (Hertz)

In-structure Spectra - Horizontal - Scaled to 0.71 g PGA (free-field) - 5% Damping 

From NUREG/CR-4550

This study

This study - 10% damping
47

of sound rock with an elevated rock pedestal about 64 ft in diameter near the center for the 

drywell foundation (see Figure 14). The foundation slab above this rock pedestal is still an RC 

slab about 4 ft thick. The main structure of the reactor building extends from the top of the 

foundation at Elevation 92 ft 6 in. to the refueling floor at Elevation 234 ft, which is topped by a 

structural steel crane bay (rated at 120 tons). For this relatively complex and relatively flexible 

foundation, justification for large reductions on high frequency ISRS spectral accelerations is 

arguable. The distance between the supports of the SFP structure, which provide direct 

pathways from the vertical ground motions of the rock to the SFP, is on the order of about 65 ft. 

This distance is less than the distance that has been considered appropriate for justifying large 

reductions of high frequency ISRS spectral accelerations (ASCE 1999). 

The above notwithstanding, results of past studies justify consideration of some reduction of the 

high-frequency ISRS spectral accelerations even without further analysis. Possible reduction of 

high-frequency ground motions is accounted for, in part, in the subsequent calculation of the 

relative likelihood of the various damage states. This is done by considering a narrow range of 

reduction in the response and associated uncertainties, as discussed in Section 4.1.5.

Figure 14 Schematic diagram of the reactor building foundation near the drywell

4.1.2 Description of the Spent Fuel Pool Structure

This section provides a brief description of the SFP structure and its relation to the main reactor 

building. The description identifies the main structural components and other aspects relevant

for this study. 

The final safety analysis report (FSAR) for PBAPS describes the SFP and the dryer-separator 

storage pool as a large channel-shaped beam (approximately 40 ft wide at the SFP structure). 

This channel beam is supported at the center by the biological concrete shield structure around 

the drywell and at the ends by RC exterior walls on opposite sides of the reactor building. Figure 

15 is a 3D representation of the SFP structure and dryer-separator storage pool. Figure 16

shows cutouts of 3D models of the reactor building that show the location of the SFP in relation 

to the remainder of the building. The 3D model on the left-hand-side of that figure ends at the 

elevation of the refueling floor (Elevation 234 ft) while the model on the right shows the crane 

bay located above the refueling floor (but not the crane itself). 15’-10”
48

The detailed 3D finite element model of the SFP structure itself (see Figure 17) serves to 

identify the walls of the pool for further reference in this study. The east (E) and west (W) walls 

extend from the biological concrete shield to the outer wall of the reactor building. These walls, 

which are about 40 ft deep (above the top of the SFP floor) and about 6 ft thick in their lower 

half, support the entire weight of the SFP, which includes their own weight, the weight of the 

floor, water, spent fuel assemblies, spent fuel racks, and the partition wall (south, S, wall). The E 

and W walls are supported by the thick RC biological shield building on the north (N) side and 

by the outer wall of the building (on the south side). A cavity exists between the SFP itself and 

the outer wall of the reactor building. 

Figure 15 SFP details in cutout of 3D CAD model

Spent Fuel 

Pool

Dryer-separator 

storage pool

Biological 

concrete shield
49

Figure 16 Cutouts of 3D CAD models of the reactor building and SFP

Figure 17 Finite element model of the SFP structure with labels for the floor and walls
50

Of interest for the study is assessment of damage and cracking to the walls identified in Figure 

17 as well as to the floor of the pool from the low probability, seismic event considered in this 

study. The walls are RC walls with vertical and horizontal layers of #11 (1.41 in. diameter) 

reinforcing steel bars near each face as well as near the mid surface of the walls. 

The SFP floor consists of an RC slab 6 ft 3 in. thick, with embedded heavy steel W-Shapes (I 

beams) as shown in Figure 15. This floor framing was used during construction and designed to 

carry the weight of the wet concrete but the beams and decking were left embedded in the 

concrete floor to the depth of the lower flange of the shapes. The beams that extend from the 

biological concrete shield to the outer wall are W-36x300 (3 ft deep beams weighing about 300 

pounds per foot) and those extending from one wall to the other are W-36x230 (3 ft deep and 

weighing about 230 pounds per foot). The floor is reinforced with steel rebar layers in two 

directions at the top of the floor and with a complex reinforcing pattern in between the steel 

girders within the clear span of the floor as well as in the portion of the floor under the side walls 

of the SFP. Vertical reinforcement near each face of the wall extends vertically into the floor slab 

and some of those bars bend and then extend horizontally into the upper half of the pool floor. 

This is done to provide adequate embedment to the reinforcing bars. 

The floor and walls of the SFP are covered with a 1/4-in. thick stainless steel liner which is 

designed to preclude inadvertent loss of water and that is attached to the concrete using steel 

anchors, and welds to steel plates and shapes embedded in the concrete floor and side walls.

Figure 18 , which is an outline of the 3D finite element model of a portion of the liner and its 

attachments to the concrete floor and walls (E and W walls), is used to identify some of these 

attachments. Interconnected drainage paths are provided behind the liner for drainage of small 

amounts of water that might leak through small cracks to a sump drain. 

Figure 18 Outline of detailed finite element model of the SFP liner representing 

attachments to the SFP floor and walls (E and W walls)

According to the FSAR, there are no connections to the SFP that would allow water to drain 

below the refueling gate or below 10 ft above the top of active fuel. The FSAR further states that 

lines below the levels in the previous sentence are equipped with siphon breaker holes to 

prevent inadvertent drainage. In addition, the systems for maintaining water quality and quantity 

SFP floor

SFP wall

Liner 

attachments Liner attached 

to embedded 

shapes

Backup 

plates
51

are designed so that failure or inappropriate operation of these systems does not cause 

uncovering of the fuel. 

The refueling gate opening (in red in Figure 15) is covered with concrete blocks and closed by 

two steel gates, in which one steel gate backs the other to provide redundancy in the case of 

malfunction of a single gate. Each gate consists of steel plates with steel stiffeners. Each gate 

has polymeric seals around its perimeter that are kept under pressure by the mechanical locking 

system for the gates. Pressurization of the seals is not a pneumatic system that requires 

pressurization by electric power systems.

4.1.3 Finite Element Model Description

Step 5 of the approach described in section 4.1.1, the nonlinear pseudodynamic analysis of the 

SFP under the combined dead loads and seismic loads, requires a detailed finite element model 

of the entire SFP structure in order to estimate concrete cracking and liner strains for the 

estimation of leakage areas. The LS-DYNA finite element software was used for the analysis

(LSTC, 2007). Figure 17 shows the overall detailed finite element model. The model has about 

600,000 elements and uses 16 elements through the thickness of the E and W walls and 

equally refined detail for the SFP floor.

The finite element model included all major reinforcing bars for the floor and walls of the SFP 

structure as well as the outer walls and biological concrete shielding. This model also

considered all steel shapes embedded in the floor of the SFP which were modeled using LSDYNA shell elements. In addition, the finite element model also includes the steel liner on the 

inside surface of the SFP. Figure 19 shows some of the components included in the finite 

element model.

Figure 19 Cutouts of 3D finite element model showing components included in the 

model

Given the complexity of the structure, rather than using node-to-node modeling for the 

embedded shell elements modeling the steel beams, the model used the “Constrained 

Liner

Reactor 

Shield 

Building

Gate

SFP 

Floor 

Beams

Ext. 

Wall
52

Lagrange in Solid” option available in LS-DYNA to represent the coupling between the 

embedded elements and the concrete. For the steel liner, two levels of modeling detail were 

used. In the calculation of the overall response of the SFP to the combined loads, liner shell 

elements of the size of the underlying concrete elements were used and the liner was assumed 

to be bonded to the concrete (node-to-node connections). A more detailed model of sections of 

the liner (see Figure 18 above) with elements as small as 3.7 millimeters (mm) (0.15 in.) wide 

was subsequently used as an embedded gage to assess strain concentrations in the liner plates 

at the intersection of the floor and walls as discussed in the following section.

Boundary conditions for the nonlinear finite element analysis are as follows: (i) vertical and 

horizontal displacements fixed at the bottom of the exterior wall and of the radiological concrete 

shield building; (ii) horizontal displacements fixed at the edges of the exterior wall and at the 

edges of the radiological concrete shield building; and (iii) horizontal displacements in the 

direction perpendicular to the E and W walls fixed at the top of the E and W walls from the 

exterior wall to the radiological concrete shield building. Fixing the horizontal displacements at 

the top of the E and W wall in the direction perpendicular to the walls is justified on the basis of 

the 1 ft 7 in. thick composite floors with a reinforced concrete deck continuous with the SFP 

walls on each side of the SFP at the top of the E and W walls (Elevation 234 ft) and that extend 

to the exterior walls. 

The finite elements in the model for the nonlinear analysis are as follows:

• reinforcing bars—LS-DYNA beam elements with the truss option.

• concrete—Constant stress LS-DYNA solid elements (reduced integration)

• shell elements—Belytschko-Tsay shell elements.

Two material models were used as follows:

• Concrete—LS-DYNA material model 159 known as the Continuous Surface Cap Model 

(CSCM) (FHWA, 2007). The analysis used the option of specifying a minimum number of 

material properties, namely the unconfined compressive strength and aggregate diameter 

and allowing the model to calculate the other material properties of interest.

• Steel—LS-DYNA material model 3, called plastic kinematic, which was used for all steels 

but with different material properties.

Table 7 provides a summary of the material properties used in the nonlinear finite element 

analyses. The properties for the concrete and steel reinforcement, assumed to be the materials 

that would most influence the overall response of the SFP, were taken to be best estimates of 

the median material properties. In the case of concrete, the unconfined compressive strength of 

the concrete was estimated based on recommendations used for the analysis of extreme 

events, namely aircraft impact assessment (NEI, 2011) and a nominal concrete strength of 

4,000 pounds per square inch (psi) (27.5 MPa). For the other materials, the table primarily lists 

nominal properties. In the case of the liner, nominal material properties were assumed for its 

yield strength and Young’s modulus. These properties and the liner itself are not expected to 

have a significant effect in the overall response of the SFP structure. However, liner strains and 

failure strains for the liner are critical in assessing the leakage potential for the SFP. An 

approach to assess failure of steel liners in reinforced concrete containments is used together 

with simple probabilistic models to estimate the relative likelihood of the damage states as 

described in Section 4.1.5.
53

Table 7 Material Properties for the Nonlinear Finite Element Analyses

Material Properties

Concrete Unconfined compressive strength

Aggregate diameter

Unit weight (and density)

Young’s Modulus (for reference)

6,400 psi (44.6 MPa)

1.5 in. (38 mm)

145 lb/ft3

 (2.33 g/cm3)

4.5x106 psi (31,000 MPa)

Rebars Yield strength (Grade 40)

Yield strength (Grade 60)

Young’s modulus

Tangent modulus

Unit weight (and density)

Failure strain

47,850 psi (330 MPa)

69,000 psi (475 MPa)

31x106 psi (2.15x105 MPa)

15x104 psi (1000 MPa)

479 lb/ft3

 (7.7 g/cm3)

0.10

Liner

and steel plate 

anchorages

Yield strength (Grade 40)

Young’s modulus

Tangent modulus

Unit weight (and density)

Failure strain

36,000 psi (250 MPa)

30x106 psi (2.07x105 MPa)

15x104 psi (1,000 MPa)

479 lb/ft3

 (7.7 g/cm3)

Treated as variable

Beams Yield strength

Young’s modulus

Tangent modulus

Unit weight (and density)

Failure strain

36,000 psi (250 MPa)

30x106 psi (2.07x105 MPa)

25x104 psi (1,700 MPa)

479 lb/ft3

 (7.7 g/cm3)

0.10

Anchor studs Yield strength

Young’s modulus

Tangent modulus

Unit weight (and density)

Failure strain

36,000 psi (250 MPa)

30x106 psi (2.07x105 MPa)

25x104 psi (1,700 MPa)

479 lb/ft3

 (7.7 g/cm3)

0.10

This study used a simpler version of the model used for the nonlinear analysis. This model was 

used to estimate frequencies of vibration for the SFP structure, to estimate seismic load 

coefficients and to verify hydrodynamic impulsive pressures with the ANSYS (version 13) finite 

element software (ANSYS, 2011). The simplified finite element was used with linear analyses 

appropriate for its intended use, had fewer elements through the thickness of the walls and 

floor, and it had a simpler representation of the concrete biological shielding. 

This finite element model used solid, elastic finite elements to represent the structure of the SFP 

(concrete only) and fluid elements to represent the water. Specifically, it used the ANSYS 

SOLID185 element, a 3D structural solid element, and the ANSYS FLUID80 element for the 

modeling of the water. Material properties considered with this model are as follows:

• Concrete: (1) Young’s modulus of 3.15x106

psi (reduced to 70-percent of the Young’s 

modulus of reference to account partially for cracking effects on stiffness) (21,700 MPa), 

(2) unit weight of 145 lb/ft3 (2.33 g/cm3

), and (3) a Poisson ratio of 0.15.

• Water: (1) bulk modulus of 3.16x105 psi (2,180 MPa), (2) unit weight of 62.4 lb/ft3 (1 

g/cm3

), and (3) a viscosity of 1.64x10-7 psi-s (1.13x10-9 MPa-s).
54

The simplified finite element model was used in conjunction with the following analyses:

• Estimation of frequencies and modes of vibration for the SFP including the effects of 

water using Householder reduced methods for the low frequency modes and the Block 

Lanczos method for the high frequency modes.

• Related deterministic spectrum analysis using single-point spectral accelerations at the 

supports together with the complete quadratic combination (CQC) rule for the 

combination of modal responses. These analyses were done to estimate seismic load 

coefficients for structural components and to verify the magnitude of the hydrodynamic 

pressures on the SFP walls.

Summary of Dead and Seismic Loads for the Finite Element Analysis

As indicated in Section 4.1.1, the dead loads considered for the nonlinear seismic analysis are 

the weight of structural materials (concrete, reinforcement, steel beams, liner and other steel 

plates), the vertical and horizontal hydrostatic pressures of the water, and the weight of the 

spent fuel assemblies and racks. The weight of the structural elements was applied as gravity 

loads on the finite element analysis. Hydrostatic pressures were applied as vertical and 

horizontal pressures on the inside surfaces of the floor and walls of the SFP. Vertical loads on 

the SFP floor from the weight of the spent fuel assemblies and racks were also applied as 

pressures on the SFP floor. Table 8 lists approximate values of the dead loads on the SFP floor 

in terms of an equivalent vertical pressure on the SFP floor for the purpose of comparing the 

magnitude of these loads with those imposed by the earthquake. Table 9 has approximate 

values of peak equivalent seismic static loads (vertical) expressed in terms of an equivalent 

vertical pressure on the SFP floor. Horizontal hydrodynamic loads (not shown in Table 9) 

considered hydrodynamic pressures from the horizontal ground motions as well as pressures on 

the wall from the vertical ground motions. 

Table 8 Approximate Dead Loads on the SFP Floor in Terms of an Equivalent Vertical 

Floor Pressure

Load Approximate equivalent floor pressure 

in lb/ft2 (in kPa in the parentheses)

Weight of the floor 900 (43)

Vertical hydrostatic pressure 2,300 (110)

Weight of spent fuel assemblies and racks 1,700 (80)

Total 4,900 (230)

 

Table 9 Approximate Peak Equivalent Seismic Loads in Terms of an Equivalent Static 

Vertical Floor Pressure

Load Approximate equivalent floor pressure 

in lb/ft2 (in kPa in the parentheses)

Floor slab acceleration 1,400 (67)

Hydrodynamic impulsive vertical pressure 4,840 (230)

Dynamic forces from spent fuel assemblies 

and racks

1,750 (85)

Total 7,990 (385)
55

The results shown in Table 8 and Table 9 indicate that the seismic loads (in terms of equivalent 

vertical pressures on the SFP floor) are approximately twice as large as the dead loads and that 

the hydrodynamic impulsive pressures on the SFP floor are the largest of all forces considered. 

Finite element analyses with the simplified finite element model described above were used to 

estimate and verify the seismic forces listed in Table 9 using deterministic response spectrum 

analysis. The seismic input for this analysis was a single point spectral acceleration at the 

supports using the 5-percent vertical and horizontal ISRS described in Section 4.1.2. It is noted 

that the (lower) natural frequencies of the SFP, considering a reduction of the concrete Young’s 

modulus to about 70-percent of its original value, the water, and the mass of the spent fuel 

assemblies and racks, range from about 14 Hz (vertical motion of the floor) to 24 Hz (horizontal 

motion of the walls). These are frequencies of interest for the estimation of both hydrodynamic 

impulsive pressures (vertical and horizontal) as well as peak accelerations of the floor (vertical) 

and walls (horizontal). Comparison of these natural frequencies with the free-field response 

spectra for this study shown in Chapter 3 of this report indicates that these frequencies are 

similar to those for which the ground motions for this study have spectral accelerations higher 

than those from the SSE when scaled to the same PGA.

Figure 20 shows contours of the peak vertical accelerations of the SFP floor obtained using the 

deterministic response spectrum analysis described in the previous paragraph with the vertical 

ISRS as a single point spectral acceleration input at the supports. The results shown are for a 

free-field PGA of 1.0g and 5-percent damping ISRS. They were multiplied by 0.71 and by the 

ratio of spectral amplitudes for 10-percent and 5-percent damping to estimate the peak 

accelerations (seismic coefficients) to be used as input for the nonlinear finite element analysis. 

To obtain corresponding forces for the nonlinear analysis, the area of the SFP floor was divided 

into a 4-ftx4-ft grid and the peak vertical accelerations were sampled at the center of each 

element of this grid. These sampled peak accelerations were then used to calculate equivalent 

nodal forces for the nodes of the detailed LS-DYNA finite element model for the nonlinear 

analysis. Estimation of equivalent nodal forces for the walls, both horizontal and vertical used a 

procedure analogous to that described for the vertical forces on the SFP floor. 

Vertical hydrodynamic forces, which are proportional to the vertical spectral accelerations at the 

base of the SFP, are the largest seismic forces in Table 9. Given the significance of these 

pressures, deterministic response spectrum analysis with the simplified ANSYS finite element 

model of the SFP was used in their calculation. Figure 21 shows peak hydrodynamic vertical 

pressures calculated in this manner for the vertical ISRS at the supports of the SFP (taken to be 

the same at each support). The pressures shown in Figure 21 are for a free-field PGA of 1.0g 

and the 5-percent damping ISRS. They were multiplied by 0.71 for the PGA of interest and by 

the ratio of spectral amplitudes for 10-percent and 5-percent damping to obtain the values 

shown in Table 9. Note that water pressures from the vertical accelerations also apply 

hydrodynamic pressures to the walls, which decrease with height above the floor. The analysis 

accounted for these pressures.
56

Figure 20 Estimated peak vertical accelerations (in/sec2

) of the SFP floor from response 

spectrum analysis and vertical ISRS as input (1.0g PGA and 5-percent damping)

Figure 21 Estimated peak hydrodynamic pressures (psi) on the SFP floor from response 

spectrum analysis and vertical ISRS as input (1.0g PGA and 5-percent damping)

4.1.4 Finite Element Analysis Results for the Spent Fuel Pool

This section presents a summary of the results obtained with the nonlinear finite element model 

described in the previous section for the loads described in Step 5 of the approach and 

estimated in Section 4.1.3. The principal objective of the analysis was to track the deformation 

of the SFP structure, concrete cracking and liner strains to estimate potential leakage rates.
57

The analysis used the LS-DYNA software which is an explicit dynamic finite element code. 

Since this is an equivalent static analysis, the analysis used mass scaling (with only minor 

changes in total mass of the model) together with slow ramping of the loads in order to minimize 

spurious dynamic effects. Specifically, the analysis slowly (with respect to the periods of 

vibration of the SFP structure) and proportionally incremented all dead loads until they reached 

their full values. Subsequently, the analysis slowly and proportionally applied all the equivalent 

seismic static loads until they reached their full values. Full values of the peak seismic loads 

were kept constant for some time in order to verify the stability of the response.

Figure 22 shows vertical displacement contours for the load combination consisting of the dead 

loads, 100-percent of the vertical seismic loads and 40-percent of all horizontal seismic loads. 

The maximum displacements are near the center of the SFP floor and are small on the order of 

0.6 in. (15 mm) or about 0.6/(40x12) = 1/800 of the clear span. Small displacements are a result 

of the high stiffness of the SFP structure which consists of thick RC slabs and walls (on the 

order of 6 ft) and comparatively short spans (from about 35 ft in the N-S direction and about 40 

ft in the E-W direction). 

Figure 22 Contours of vertical displacements (mm) of the SFP floor and walls

Figure 23 shows vertical displacement along the outside face of the W wall. Of special interest 

in Figure 23 are the discontinuities of vertical displacement at the bottom of the SFP wall at the 

top of the SFP floor, which are identified by the transition between the blue and green contours 

near the center of span at the bottom of the wall. Discontinuities of vertical displacements in this 

region are of interest because this is the region of possible strain concentrations in the SFP liner 

as shown in Figure 24. Finally, Figure 25 shows (with the red contour) the region of the SFP, at 

the bottom of the SFP walls and at the top of the SFP floor where the tensile strain of the 

concrete is exceeded and a crack could likely develop. The crack would start as a flexure crack 

and develop into a mostly tension-flexure crack though the thickness of the wall accompanied 

by shear friction at the bottom of the wall. 
58

Figure 23 Contours of vertical displacement (mm) of the SFP walls

Figure 24 Liner strains (overall response not fully accounting for strain concentrations)
59

Figure 25 Region of concrete cracking initiation at the floor-wall junction

The higher liner strains in Figure 24 are, as expected, at the intersection of the SFP wall with 

the SFP floor, which is the region of strain concentrations. Although this is a region of strain 

concentrations, the liner strains shown are small, of the order of 5x10-4 to 1.9x10-3

. For 

comparison, the nominal liner yield strain is 1.2x10-3

. The mesh size for the liner for this overall 

finite element analysis is not sufficiently small to fully capture strain concentrations in the liner. 

The main objective of this analysis was to obtain the overall deformation of the structure and the 

development of concrete cracking which is not expected to depend significantly on the details of 

the liner modeling.

To assess strain concentrations in the liner, a detailed finite element model of the liner which 

includes the main details of its attachments to the floor and wall concrete was developed and is

shown in Figure 18. The fine mesh of this liner inset uses elements as small as 0.15 in. (about 

3.7 mm) at the transition from the floor to the wall. The analysis used this detailed liner insert to 

estimate the liner strains. Specifically, the detailed insert was embedded into the original 

nonlinear finite element model of the structure. The SFP structure was then analyzed with the 

embedded detailed model of the liner (using the “Constrained_Lagrange_in_Solid” option in LSDYNA and appropriate contact definitions) to assess strain concentrations in the liner. Note that 

the actual liner and the liner in the model are attached to the concrete only at a few discrete 

locations. Elsewhere, the liner is only in contact with the concrete. Specifically, at the junction 

with the wall, the liner is attached to concrete only near the backup plates between the floor and 

wall (see Figure 18) and is in contact with the concrete elsewhere along the floor-wall junction. 

For this reason, high strain concentrations are expected to develop only near the backup plates. 

Figure 26 shows results of the analysis of the SFP with the embedded liner in a portion of the 

wall near the region where strain concentrations are expected to be the largest. The results 

show that the presence of the embedded liner as a gage does not affect the overall response of 

the SFP in a significant manner. However, it permits an estimation of the strain concentrations 

in the liner. 
60

Figure 26 SFP displacements (mm) with detailed liner insert

Figure 27 shows the strain concentrations in the liner calculated using the detailed liner insert as 

indicated above. As expected strain concentrations are localized to the region of the liner near 

the backup plates, i.e., where the liner is attached to the shapes embedded in the SFP floor. 

Elsewhere the liner strains remain small as indicated by the overall analysis with the coarser 

model. The maximum membrane effective strain in Figure 27 is about 3.7 percent (0.037). The 

following section uses these strains as well as estimates of the width and extent of the concrete 

cracking (see Figure 25), to assess liner tearing likelihoods for the scenario considered.

Figure 27 Strain concentrations in the SFP liner

SFP Study
61

4.1.5 Damage States

This section documents the results for Steps 6 to 9 of the approach defined in Section 4.1.1, 

which uses results from the nonlinear finite element analysis described in Section 4.1.4 to 

estimate leakage rates. These leakage rates are then used in the accident progression analysis 

to define the rate of loss of water from leakage at the bottom of the SFP. The section starts with 

the approach used to estimate the likelihood for each damage state for the initiating seismic 

event considered. Then, the section provides the estimation of the leakage rates for the 

damage states with leakage.

Damage States and Relative Likelihoods

Step 6 of the approach (Section 4.1.1) defined three initial damage states as follows: 

a. No leakage: A state with no leakage from the bottom of the pool. This state 

corresponds to concrete cracking at the base of the walls (estimated to be throughwall cracking for the event considered as shown in previous subsections) but without 

tearing of the liner.

b. Moderate leakage rate: A state with leakage from the bottom of the SFP, 

corresponding to through-wall concrete cracking at the bottom of the walls and tearing 

of the liner that propagates to an extent such that water leakage is controlled by the 

size of the cracks in the concrete. 

c. Small leakage rate: A state with leakage from the bottom of the SFP, corresponding 

to through-wall concrete cracking at the bottom of the walls and tearing of the liner 

that remains localized to the where the floor liner is attached to the SFP floor near the 

walls. 

This study uses an approach and strain criteria, including uncertainties, for tearing of steel liners 

in reinforced concrete containments (Cherry, 2001 and 1996) together with uncertainties in the 

calculated liner strains to estimate the relative likelihoods for the three initial damage states 

listed. Uncertainties in the calculated liner strains account for (1) uncertainties in the ISRS 

spectral accelerations (of the order of 25 percent), (2) uncertainties in liner strains from 

uncertainties in concrete properties (namely concrete strength) and (3) an additional reduction 

in spectral accelerations to account for both ground motion incoherency and nonlinear effects.

The analysis used information in Cherry (1996) to estimate upper and lower bounds for the 

limiting failure strain, which were then used with a triangular probability density function to 

estimate their mean and coefficient of variation (Ang, 1984). This is expected to be a 

conservative assumption in that the SFP liner is of stainless steel which is likely to have larger 

limiting failure strains. This approach, adjusts the failure strain from coupon tests using 

reduction factors that account for the multi-dimensional state of stress (triaxiality effects), 

uncertainties in material properties, and the level of detail in the analysis used to estimate strain 

concentrations. Bounds in the liner limiting failure strain use a failure strain from coupon tests of 

21-percent (0.21) together with a triaxiality factor of 1.75 (typical of a cylindrical state of stress). 

Estimation of these bounds considers a high level of detail in the model for the calculation of

strain concentrations which used elements as small as 0.15 in (3.7 mm) wide. Accordingly, this 

study considered a range of reduction factors for the analysis detail that range from 0.4 to 0.9. 

Reduction factors for material properties were those reported in Cherry (1996). On these 

bases, the bounds on the failure strain for the purposes of estimating its mean and coefficient of 
62

variation came to be 0.045 and 0.14, and the resulting mean and coefficient of variation came 

out to be 0.09 and 0.20, respectively. 

Maximum effective tensile strains in the liner were calculated assuming reduced material 

properties and were found to be sensitive to the concrete strength. Effective strains calculated 

for the median concrete strength and a reduced concrete strength were used to assess the 

derivative of this strain to the concrete strength. This derivative was then multiplied by the 

standard deviation of the concrete strength which was calculated using an estimated coefficient 

of variation for the concrete strength of 0.15 (Lambright et al., 1990) to estimate the standard 

deviation and coefficient of variation of the liner strain associated with uncertainties in the 

concrete strength. This coefficient of variation was estimated to be about 0.65.

Maximum effective concrete strains calculated for the 0.7 g PGA and for reduced seismic loads 

(about 70-percent of the initial loads) were used to estimate the sensitivity of the effective strain 

to the estimated spectral amplitudes. Nonlinear analyses were done to estimate maximum 

effective strains for spectral accelerations equal to about 80-percent of the original to account 

for effects of ground motion incoherency and further reductions from nonlinear effects. 

Additional uncertainty measures for the calculated strain were then estimated using the 

calculated strains for the base case and the case with reduced spectral amplitudes in 

conjunction with an asymmetric triangular distribution for the calculated strains. The assumed 

triangular distribution used the strain for the reduced value as the least likely value and that for 

the base case as the most likely value. This procedure resulted in an adjustment of the median 

strain (reduction factor equal to 0.93) and an additional coefficient of variation (0.09) for the liner 

strain. An additional coefficient of variation for the ultimate strain of about 0.25 was used to 

account for uncertainties in the estimate of floor response spectra ordinates (Lambright et al., 

1989).

Uncertainties calculated in this manner were then used to estimate medians and coefficients of 

variation for the limiting failure strain (capacity) and for the induced strain (demand). Using 

these quantities and assuming lognormal distributions, the probability of liner tearing conditional 

on the occurrence of the seismic event was estimated to be less than 10-percent (Ang, 2006; 

Ang, 1984). This estimate indicates that the state with no leakage (no tearing of the liner) is the 

most likely with a relative likelihood in excess of 90-percent. The relative likelihood of the two 

states with leakage from the bottom of the SFP is estimated at less than 10-percent. Assigning 

relative likelihoods to the two damage states with leakage is subject to considerable 

uncertainties at this time. Accordingly, the assumption is made that both states are equally 

unlikely.

Concrete Cracking and Moderate Leakage Rate

Postprocessing of the displacements at the top of bottom nodes of the horizontal layer of 

concrete finite elements at the top of the SFP floor provides an estimate of the width and length 

of the cracking at the bottom of the SFP walls. The first step of this processing is the sampling 

of vertical displacements at the top and bottom nodes of this layer of concrete elements at 

various locations along the perimeter and through the depth of the wall. This is achieved by 

dividing the length of the base of the wall into segments and sampling those quantities at 

locations across the wall thickness near the center of each segment. The next step consists of 

subtracting the displacements of the top and bottom nodes for a first estimate of the crack width 

at the sampled locations. This estimate is then corrected by subtracting the vertical 

displacement of those nodes implied by the tensile strain of the concrete at cracking, which is 

comparatively small. A main assumption in this process is that a major single concrete crack 
63

(flexure-tension crack for this SFP) develops at the floor-wall junction rather than a set of closely 

spaced minor cracks. The next step averages the sampled crack widths though the thickness of 

the walls for each sampled segment at the base of the walls. Finally, the processing combines 

the crack areas estimated in this manner to estimate an average crack width of about 3.6 mm 

(about 0.14 in.) and an average crack length of about 33,000 mm (about 108 ft), with a nonsmooth and non-uniform surface. An average crack width is used because the overall change 

in the crack width is not expected to be large along the perimeter of the floor.

Estimation of the flow through this crack used recent experimental data for the flow of water 

through thick cracked concrete sections for hydraulic pressures similar to those in the SFP 

(Kanitkar et al., 2011). Crack widths and water pressures for those tests envelope the average 

crack width estimated for the SFP and the water pressures in the SFP. The thickness of the 

concrete slabs is about half of the thickness of the SFP walls, meaning that these are large 

scale tests. Main results of that testing are (1) an equation to estimate the leakage flow rate 

through concrete cracks that involves a friction factor and (2) quantification of that friction factor 

based on the experimental data. Specifically, the study recommends the use of the following 

equation derived from the Navier-Stokes equations for incompressible flow of a Newtonian fluid:

w

d

g

v f

g

v

g

P

2 2 2

2 2

= +

ρ

where P is the pressure, ρ is the fluid density, g is the acceleration of gravity, v is the flow 

velocity, d is the crack depth (concrete thickness), and f is a friction factor. The results reported 

indicate that a friction factor of 0.8 is adequate for the average crack width estimated above.

Using the equation above for the leakage flow, and a friction factor of 0.8, assuming no initial 

loss of water and using the crack width and length estimated above, the leakage flow was 

calculated as shown in Figure 28 in terms of the change of the water height in the SFP with 

time. The flow rate in that figure represents a moderate flow rate condition. The average flow 

rate for this condition to a height of about 16 ft above the SFP floor is about 1,500 gallons per 

minute. 

For this condition to occur it is necessary that the liner strains exceed failure strains for the liner 

material at the region of strain concentrations (near the backup plates), that these tears become 

unstable and that the liner tearing spreads to an extent such that the leakage rate through the 

liner is greater than the leakage rate through the concrete cracks. In this case, concrete 

cracking controls the leakage rate from the SFP. This is further discussed below in conjunction 

with the liner strains and liner failure criteria as well as the estimation of the relative likelihoods 

for the three damage states considered. 
64

Figure 28 Moderate leakage flow rate (through concrete cracks)

Liner Strains and Small Leakage Rates

Maximum effective membrane liner strains from strain concentrations at the floor-walls junction 

are on the order of 0.037 (3.7 percent). These strains are localized at the backup plates, which 

are spaced 24 in (609.6 mm) apart along the length of the E and W walls. Attachment details 

along the S wall are different, imposing less compliance of the liner to the concrete 

deformations, and are not expected to lead to strain concentrations as large as those at the 

base of the E and W walls. In addition, liner strains near the biological concrete shielding are 

smaller. Moreover, liner tearing or through wall (or floor) concrete cracking are not expected 

near the biological concrete shielding. Accordingly, tearing of the liner, if it were to occur, would 

be only along the base of the E and W walls.

An approach and failure criteria for steel liners used in reinforced concrete containments is used 

here to assess tearing of the SFP liner (Cherry, 2001 and 1996). Failure criteria for liners 

without corrosion damage reported by Cherry (1996) are used in this study to estimate limiting 

failure strains for the stainless steel SFP liner. The approach estimates the crack width by 

multiplying the liner strain at failure by the width of the finite element with the maximum induced 

effective strain, which is approximately equal to 0.15 in (3.7 mm) as indicated above. 

Since both the induced strains (demands) and failure strains (capacity) are treated as random 

variables, the strain at which the liner would tear, that is the condition at which the induced 

strain exceeds the limiting failure strain, is also random. An approach for a point estimate of 

that strain would be to calculate the most likely failure strain, which would be a strain greater 

than the estimated median induced strain (demand) of 0.37 but likely less than the median 

limiting failure strain (capacity) of about 0.10. Such an approach would involve a more detailed 

uncertainty analysis and probabilistic modeling than that used in this study, which does not 

seem justified given the approximations used as well as the uncertainties involved in the 

assessment of the flow rates through tears in the liner. This study assumed a failure strain of 

0.10 (10 percent) for the liner strain at failure, which is approximately equal to the assumed 

median failure strain.

0.0

5.0

10.0

15.0

20.0

25.0

30.0

35.0

40.0

0 3600 7200 10800 14400 18000 21600 25200 28800

Water Height (ft)

Time (s)

Water Height

Water Height (accounting

for volume of spent fuel)
65

The resulting crack width for a liner tear localized at the location of the backup bar is then 

estimated at 0.15x0.10 = 0.015 in (0.37 mm). The crack length at each location is taken to be 

equal to the width of a backup bar which is equal to 4.0 in (101.6 mm). Given that the spacing 

of the backup bars is 24 in (609.6 mm), a total of 40 backup bars (20 on each wall) are used to 

estimate the summed length of all localized cracks as 40x4 = 160 in (4,064 mm). The estimated 

width for each crack, if it were to occur, is then 0.015 in (0.37 mm) and the depth of the crack is 

the depth of the liner which is equal to 0.25 in (6.35 mm).

 

Given the estimated width, length and depth for each localized liner tear and their number, it is 

still necessary to estimate the leakage rate through these tears. Estimation of this flow rate 

uses the following assumptions (1) the flow rate can be estimated using an equation similar to 

that used for flow through the concrete cracks and (2) the friction factor for that equation can be 

calculated on the basis of test results for leakage rates through cracks in pipes. These 

assumptions are not validated at this time. Therefore, considerable uncertainty exists for the 

resulting leakage rate estimate. The following paragraph addresses the process used to 

estimate the flow rate through these liner tears as well as sources of uncertainty for this 

estimation. These uncertainties may result in flow rate estimates that can vary by more than 

100%. This damage state (small leakage rate) already is a result of binning the uncertain liner 

tearing into two discrete tearing conditions to cover a range of uncertainty for liner damage and 

associated flow rates. Assigning equal likelihood to the two highly distinct damage states 

acknowledges these uncertainties.

Estimation of a friction factor was made using data in Paul et al. (1994) for leakage through 

cracks in steel pipes. Back calculation of friction factors from data presented in this reference 

shows a large variability in the calculated friction factor. In particular, the friction factor appears 

to depend heavily on the smoothness of the crack surface. Also, the fluid in the pipe is at high 

temperatures and the driving pressures are much higher than those applicable to the SFP. 

Review of other flow models reported in Paul et al. (1994) indicates that for relatively smooth 

cracks friction will be low. Assuming relatively smooth cracks, the equation for the flow through 

concrete cracks was applied for flow through steel tears together with a small friction factor 

(0.11) in order to estimate the leakage flow rates. For this friction factor, the estimated leakage 

flow through the steel cracks (small leakage flow) is as shown in Figure 29. Considerable 

uncertainty continues to exist in the estimation of leakage flow rates for these localized liner 

tears. Given the assumption that the crack surface is relatively smooth, it is estimated that the 

flow rates in Figure 29 would be greater than the actual flow rates. The average flow rate for this

condition to a height of about 16 ft (488 mm) above the SFP floor is about 200 gallons per 

minute (757 liters per minute).
66

Figure 29 Small leakage flow rate (through localized steel tears)

4.2 Other Damage States

Assessment of other damage stages is primarily based on (1) finite element deterministic 

response spectra analysis to estimate maximum vertical displacements of the water surface 

(sloshing), (2) seismic fragilities used in conjunction with the NUREG-1150 seismic PRA study 

(Lambright et al., 1990), (3) the examination of design details for certain appurtenances such as 

the refueling gate, and (4) maximum displacements (vertical and horizontal) of the SFP floors 

and walls under the applied loads.

Loss of Water from Sloshing

Vertical displacements of the water surface (sloshing) that may lead to the ejection of some 

water from the SFP depend on the low frequency components of the motions at the base of the 

SFP. Finite element analysis using the ANSYS finite element model described above, show that 

the natural frequencies of the sloshing modes in the two horizontal directions parallel to the 

walls of the SFP are about 0.27 Hz and 0.29 Hz, corresponding to periods of vibration on the 

order of about 3.8 to 3.5 seconds. These results resemble those obtained using analytical 

methods (e.g., AEC, 1963; Malhotra et al., 2000). 

The free-field ground motion specified for the study does not have high spectral velocities and 

accelerations at the sloshing frequencies. Consequently, sloshing amplitudes are expected to 

be small. Deterministic response spectrum analyses with the simplified ANSYS finite element 

model of the SFP using the horizontal ISRS at midheight of the SFP (for the frequencies of 

interest to sloshing) as input and considering the low damping of the sloshing mode, show that 

the sloshing amplitude will not exceed about 20 in. Given that the water at the pool is about 1 ft 

below the top of the SFP, sloshing is not expected to cause more than 1 ft of water loss. 

Accordingly, an initial 1.5 ft decrease in the height of the water is considered at the end of the 

earthquake event for the subsequent accident progression analysis.

0.0

5.0

10.0

15.0

20.0

25.0

30.0

35.0

40.0

0 50000 100000 150000 200000

Water Height (ft)

Time (s)

Water Height

Water Height (accounting for 

volume of the spent fuel)
67

Damage to Refuel Gate, SFP Penetrations, Spent Fuel Assemblies and Racks

Refuel gate: A site visit and examination of the refueling gate structural drawings revealed the 

following:

• The steel gate next to the water is backed by a similar gate.

• Each of these gates consists of a steel-plated decking with steel stiffeners.

• Each gate has a polymeric seal around its perimeter that is pressed against the concrete 

by passive mechanical means that are not expected to be lost during the seismic event. 

Since these are passive mechanical means the effectiveness of the seals does not 

depend on the availability of ac or dc power.

• Tolerances around the seals are sufficient to accommodate the already small distortions 

of the biological concrete shielding in the refueling area from the seismic event.

Based on the above, the study assumes that the refueling gate will not fail for the seismic event 

considered and will continue to maintain its intended function during the accident progression. 

SFP penetrations: According to the FSAR, there are no connections to the SFP that would allow 

water to drain below the refueling gate or below 10 ft above the top of active fuel. The FSAR 

further states that lines below the levels in the previous paragraph are equipped with siphon 

breaker holes to prevent inadvertent drainage. In addition, the systems for maintaining water 

quality and quantity are designed so that failure or inappropriate operation of these systems 

does not cause uncovering of the fuel. Results of the nonlinear finite element analysis also 

indicate that overall distortions of the pool walls are small (on the order of a few millimeters). 

These distortions are not expected to lead to seismically induced damage of the penetrations 

that would lead to potential leakage.

Spent fuel racks and assemblies: Damage to the spent fuel assemblies and racks was not 

calculated as part of this study. The study assumes that under the applied seismic loads a 

coolable configuration would be maintained. This assumption is consistent with the seismic 

assessments made in conjunction with the resolution of GI-82 and reported in NUREG/CR-5176 

(Prassinos et al., 1989). As in the case considered in GI-82, the spent fuel racks for the site 

considered are allowed to slide, which tends to reduce the magnitude of the seismic 

accelerations on the racks and partially decouple their dynamic response from the response of 

the SFP. In addition, the high-frequency components (greater than 10 Hz) of the motion would 

not be expected to induce large sliding or rocking motions.

Damage to the Reactor Building and Other Relevant SSCs

According to the fragility analysis for the NUREG-1150 seismic PRA (Lambright et al., 1990), 

the median fragility for the reactor building is about 1.6g. The response of the reactor building 

structure is expected to be more sensitive to the horizontal ground motions than to the vertical 

ground motions. Natural frequencies of vibration for horizontal modes of vibration of the reactor 

building are about 7 Hz (i.e., frequencies at which the spectral accelerations of the ground 

motion for the scenario considered are less than those for the ground motions with the same 

PGA considered in earlier evaluations of the median fragility). On these bases, seismicallyinduced failure or severe damage to the reactor building would not be expected for the seismic 

scenario considered.
68

Examination of structural drawings for the Peach Bottom reactor buildings together with a 

simple kinematic analysis indicates that if the crane bridge were to lose support at one of its

ends as a consequence of the ground shaking, that end of the crane bridge would not fall inside 

the SFP. Depending on the end of the crane bridge losing support, the crane could fall only a 

few feet from the SFP, but not inside the SFP.

A LOOP is expected for the seismic scenario considered. Median fragilities for loss of offsite 

power, in terms of PGA, are less than half the PGA for the seismic motion considered in this 

study. Review of the fragilities estimated for NUREG-1150 study (Lambright et al., 1990)

indicates a high probability of loss of onsite ac power (about 0.84). This estimate is based on 

either direct failure of the onsite emergency diesel generators (assumed to be sensitive to 

spectral accelerations around the 20 Hz frequency) or failure of either the emergency service 

water or the emergency cooling water systems that provide cooling water for the diesel 

generators. The probability of losing dc power based on the fragility of the inverters alone is 

estimated to be close to but less than 50-percent for the seismic event considered in this study.

4.3 Review of Spent Fuel Pool Performance under Recent Major 

Earthquakes in Japan

Five Japanese nuclear power plant sites with a combined total of 20 reactors and 20 SFPs were 

subjected to severe ground motions from two major earthquakes in the past 5 years (NERH, 

2011a; NERH, 2011b; Kawamura, 2008; Sato, 2010):

• March 11, 2011, Tohoku earthquake (with moment magnitude Mw= 9.0)

- Fukushima Daiichi (5 BWR Mark I and 1 BWR Mark II SFPs)

- Onagawa (3 BWR SFPs)

- Fukushima Daiini (4 BWR SFPs)

- Tokai (1 BWR SFP)

• July 16, 2007, Niigataken Chuetsu-Oki earthquake (Mw= 6.6)

- Kashiwazaki-Kariwa (7 BWR SFPs)

This review addresses reductions in water levels for the SFPs affected by those events that 

might have resulted from either water leakage from structural damage or water loss from 

sloshing, if any. 

No leakage of water near the bottom of the SFPs has been reported for any of the 20 SFPs in 

those five nuclear power plants for these two major earthquakes. For the Kashiwazaki-Kariwa 

site, the only report of water loss (leakage or sloshing) for the seven SFPs at the site was a loss 

of about 320 gallons (about 1.2 cubic meters) from sloshing of the water in the SFP of Unit 6

(Kawamura, 2008). 

Loss of water other than from sloshing was not reported for the SFPs of the power plants 

affected by the March 11, 2011 Tohoku earthquake (NERH, 2011b). According to the NERH 

(2011b) report, minor leaks of radioactive material (all contained inside buildings) at the 

Onagawa plant were attributed to sloshing of SFP water, and SFP sloshing overflow lead to a 8 

in (20 cm) decrease of the water level in the SFP at Tokai. Actual decreases in SFP water levels 

from sloshing at the Fukushima Daiichi units are not known, but decreases in water level from 

sloshing have been assumed in evaluations of SFP performance (NERH, 2011b). Specifically, a 
69

water level reduction of about 1.6 ft (0.5 m) was assumed for Unit 2 as a result of sloshing 

induced by the ground motion while reductions of about 5 ft (1.5 m) were assumed for Units 1, 3 

and 4 from sloshing associated with ground motions and explosions.

This review also provides a comparison of ground motion indices and ISRS spectral 

accelerations considered for this study and observed at the various units of those nuclear power 

plants. Although this review and comparison use information available at the time of the 

execution of this study, they assist in the interpretation of the results obtained for the seismic 

scenario and SFP considered in this study. 

It is noted that the seismic design loads for the various reactors considered in this comparison 

differ, for the most part, from the design basis loads for the site considered in the SFP Study. A 

possible exception to this would be Unit 1 at Fukushima Daiichi, which initially considered 

comparable seismic design-basis loads. However, seismic design basis loads for this reactor 

were subsequently revised upwards (those are the design loads reported in this comparison). 

Differences in the seismic design-basis loads and uncertainties regarding the construction 

details (e.g., out of plane shear reinforcement if any) for the various SFPs listed above add to 

the overall level of uncertainty in the comparisons. However, this section provides a 

comparison of the structure of the SFP considered in this study and the structure for the SFP of 

Fukushima Daiichi Unit 4, for which some structural information was available at the time of the 

writing of this report. 

Another source of uncertainty for this comparison is that the recorded ground motions and 

related PGAs at the various sites are not, for the most part, free-field ground motions and, 

therefore, are not directly comparable to the free-field PGA considered in the study. However, 

the free-field ground motion for this study is also taken to be the foundation ground motion 

because the reactor building is considered to be a fixed-base structure. Additional sources of 

uncertainty are the type of reactor (several of the plants have Mark II containments instead of 

Mark I containments), site conditions (soil versus rock sites), reactor building foundation (slab 

thickness and uniformity) and reactor building embedment. Generally, the foundation slabs for 

the reactors listed above are thicker and more uniform than that for the reactor considered in the 

study. Also, the site for the study is a rock site and stiffer than the sites for Fukushima Daiichi 

and Kashiwazaki-Kariwa. 

An additional source of uncertainty for the comparison is that ISRS reported for some plants 

may be affected by localized structural details such as the vertical response of a floor slab. Such 

ISRS would not be representative of the seismic loads on the SFP in the same sense as the 

ISRS used in this study. Precise determination of the location of the accelerometers used for the 

observed ISRS was not done for these review and comparison. Table 10 to Table 14 show 

horizontal and vertical PGAs observed at the foundation slab of the various units for each of the 

nuclear power plants. Those tables also list the design PGAs for each of the reactors. For 

comparison, the vertical and horizontal PGAs for the free-field ground motion considered in this 

study are about 0.7g. On the basis of the values reported on those tables, the following 

observations are possible:

• Horizontal PGAs at the foundation slabs of all reactors are less than those considered in 

the study with the exception of that for Kashiwazaki-Kariwa Unit 1. 

• Vertical PGAs at the foundation slabs of all reactors are for the most part less than 

horizontal PGAs with the exception of Fukushima Daiini Unit 1 and Kashiwazaki-Kariwa

Units 6 and 7.
70

• Vertical PGAs at the foundation slabs of all reactors are less than those considered in the 

study. 

• The difference between the recorded PGAs and the PGA for the study is greater for the 

vertical accelerations than for the horizontal accelerations. 

- The study assumes that the vertical PGA is approximately equal to the horizontal 

PGA (see Section 3.3).

Table 10 Fukushima Daiichi, Measured and Design (DBGM SS) PGAs at Foundation Slab

(Tohoku, 2011 Earthquake)

Unit Containment

Measured (cm/s2) Design Values (cm/s2)

Horizontal Vertical Horizontal Vertical

NS EW NS EW

1 Mark I 460 447 258 487 489 412

2 Mark I 348 550 302 441 438 420

3 Mark I 322 507 231 449 441 429

4 Mark I 281 319 200 447 445 422

5 Mark I 311 548 256 452 452 427

6 Mark II 298 444 244 445 448 415

Table 11 Onagawa, Measured and Design (DBGM SS) PGAs at Foundation Slab

(Tohoku, 2011 Earthquake)

Unit Reactor

Measured (cm/s2) Design Values (cm/s2)

Horizontal Vertical Horizontal Vertical

NS EW NS EW

1 BWR 540 587 439 532 529 451

2 BWR 607 461 389 594 572 490

3 BWR 573 458 321 512 497 476

Table 12 Fukushima Daiini, Measured and Design (DBGM SS) PGAs at Foundation Slab 

(Tohoku, 2011 Earthquake)

Unit Reactor

Measured (cm/s2) Design Values (cm/s2)

Horizontal Vertical Horizontal Vertical

NS EW NS EW

1 Mark II 254 230 305 434 434 512

2 Mark II 243 196 232 428 429 504

3 Mark II 277 216 208 428 430 504

4 Mark II 210 205 288 415 415 504
71

Table 13 Tokai, Measured and Design (DBGM SS) PGAs at Foundation Slab (Tohoku, 

2011 Earthquake)

Unit Reactor

Measured (cm/s2) Design Values (cm/s2)

Horizontal Vertical Horizontal Vertical

NS EW NS EW

1 Mark II 214 215 189 393 400 456

Table 14 Kashiwazaki-Kariwa, Measured and Design PGAs at Foundation Slab

(Chuetsu-Oki, 2007 Earthquake)

Unit Reactor

Measured (cm/s2) Design Values (cm/s2)

Horizontal Vertical Horizontal Vertical

NS EW NS EW

1 Mark II 311 680 408 274 273 235

2 Mark II 304 606 282 167 167 235

3 Mark II 308 384 311 192 193 235

4 Mark II 310 492 337 193 194 235

5 Mark II 277 442 205 249 254 235

6 ABWR 271 322 488 263 263 235

7 ABWR 267 356 355 263 263 235

Another aspect of interest for this comparison is the frequency content of the ground motions as 

characterized by response spectra. The site chosen for the study is a rock site and dominant 

seismic event for this scenario would be an earthquake in the CEUS at a distance of about 15 

km or less. Accordingly, the ground motion response spectra for the seismic scenario 

considered has maximum spectral accelerations for frequencies greater than about 10 Hz and 

at frequencies near the lower fundamental frequencies of the spent fuel pool structures. 

Figure 30 includes vertical response spectra for 5-percent damping at the foundation slab of 

Unit 1 (the case with a horizontal PGA of about 0.7g) and Unit 4 of Kashiwazaki-Kariwa together 

with the corresponding response spectrum for the vertical ground motion considered for the 

study. This comparison indicates that the ground motion for this study has higher vertical 

spectral accelerations near the lower fundamental frequencies of vibration of the SFP structure. 

Spectral accelerations for the ground motion used in the study remain higher than those for Unit 

4 down to a frequency of about 5 Hz and those for Unit 1 down to frequencies of about 4 Hz. 

The results shown are typical of those for the other reactors at Kashiwazaki-Kariwa (with the 

possible exception of Unit 6 which has significantly higher spectral accelerations between 6 and 

2.5 Hz). The reactors at this plant are Mark II reactors, have reinforced concrete base slabs 

several times thicker than the reactor considered in this study, and are deeply embedded in the 

ground.
72

Figure 30 Vertical response spectra: Kashiwazaki-Kariwa Units 1 and 4 (foundation level) 

and SFP study (free-field)

With the exception of Unit 4 at Fukushima Daiichi, vertical response spectra for the reactors 

affected by the March 11, 2011 Tohoku earthquake were not available at the time of the study,

so the comparison of foundation response spectra are, for the most part, made using horizontal 

spectra. Figure 31 shows horizontal response spectra for 5-percent damping at the foundation 

slab of Unit 1 and Unit 4 of Fukushima-Daiichi and the corresponding response spectrum for the 

horizontal ground motion used in the study. Comparison of those spectra indicates that the 

ground motion for this study (rock site) has higher horizontal spectral accelerations at the lower 

natural frequencies of the SFP structure (about 0.05 seconds). Spectral accelerations for the 

ground motion used in the study remain higher than those for Unit 1 for frequencies down to 

about 4 Hz and those for Unit 4 for frequencies down to about 3 Hz. The results shown are, in 

general, typical of those for the other reactors at Fukushima Daiichi.

0

500

1000

1500

2000

0 1 10 100

Acceleration, cm/s

2

Frequency (Hz)

Vertical Response Spectra - 5% damping

KK - Unit 1

KK - Unit 4

SFP Study
73

Figure 31 Horizontal response spectra: Fukushima Daiichi Units 1 and 4 (foundation) 

and SFP study (free-field)

Figure 32 shows vertical response spectra for 5-percent damping at the foundation slab of Unit 

4 of Fukushima-Daiichi and the corresponding response spectrum for the vertical ground motion 

used in the study. Comparison of those spectra indicates that the ground motion for this study 

(rock site) has higher horizontal spectral accelerations at the lower natural frequencies of the 

SFP structure (about 0.05 seconds). Spectral accelerations for the ground motion used in the 

study remain higher than those for Unit 4 for frequencies down to about 3.5 Hz.

Figure 32 Vertical response spectra: Fukushima Daiichi Unit 4 (foundation) and SFP 

study (free-field)

0

500

1000

1500

2000

0 1 10 100

Acceleration (cm/s

2)

Frequency (Hertz)

Horizontal Response Spectra - 5% damping

Fukushima 1 - Unit 1

Fukushima 1 - Unit 2

Fukushima 1 - Unit 3

Fukushima 1 - Unit 4

SFP Study

0

500

1000

1500

2000

0 1 10 100

Acceleration (cm/s

2)

Frequency (Hertz)

Vertical Response Spectra - 5% damping

Fukushima 1 - Unit 4

SFP Study
74

Figure 33 shows vertical ISRS at an elevation at about midheight of the SFP for Unit 1 and Unit 

4 of Kashiwazaki-Kariwa together with the vertical ISRS for the study. ISRS for the study are 

shown for 5-percent and 10-percent damping. In both cases, the ISRS for the study is higher 

than the observed ISRS for frequencies close to the lower natural frequencies of the SFP 

considered in the study. The 5-percent ISRS for this study remains above that for Unit 4 down 

to frequencies of about 12 Hz and approximately equal to it for frequencies down to about 7 Hz. 

The 5-percent ISRS for this study remains higher than that for Unit 1 for frequencies down to 

about 4 Hz.

The 10-percent damping ISRS for the reactor building approaches that of Unit 4 at frequencies 

equal to about 17 Hz. The 10-percent ISRS for this study is higher than that for Unit 1 for 

frequencies down to about 6 Hz and is close to it at about 12 Hz. The ISRS for Unit 1 is typical 

of those for the other units with the exception of Unit 3, which approaches the 5-percent 

damping ISRS for the study at about 11 Hz.

Figure 33 Vertical ISRS for Kashiwazaki-Kariwa Units 1 and 4 and for the SFP Study

The comparisons, especially the comparison of the vertical response spectra at the foundation 

of Unit 4 of Fukushima Daiichi and at the base of the SFP for the study, indicate that the 

vibratory loads for this study, especially the vertical loads, are likely to be more challenging to 

the SFP than those from the actual events. 

Structural and Construction Details 

The seismic design loads for the various reactors considered in this comparison differ, for the 

most part, from the design basis loads for the SFP considered in this study. A possible 

exception to this would be Unit 1 at Fukushima Daiichi, which initially considered comparable 

seismic design-basis loads. However, seismic design basis loads for Unit 1 were subsequently 

revised upwards (those are the design loads reported in this comparison).

The depth of the 20 SFPs affected by the recent earthquakes in Japan is similar to that for the 

SFP considered in the study. The horizontal dimensions for the SFPs in Fukushima Daiichi 

(EW and NS dimensions with reference to Figure 17) are also similar with the exception to the 

0

5

10

15

20

25

30

0 1 10 100

Acceleration, m/s

2

Frequency (Hz)

Vertical In-Structure Response Spectra

KK - Unit 1 - 5% Damping

KK - Unit 4 - 5% Damping

SFP study - 5% Damping

SFP Study - 10% Damping
75

SFP for Unit 1, which has a significantly smaller NS span that tends to make the SFP for Unit 1 

less vulnerable to seismic loads. The thickness of the floor slabs for the SFPs at Fukushima 

Daiichi are likely similar to those for the SFP considered in the study. The SFP for which more 

structural details were known at the time of the writing of this report is the SFP for Unit 4 at 

Fukushima Daiichi (hereafter referred to as Unit 4). The following provides a comparison of 

structural details for Unit 4 with those of the SFP considered in this study.

• For Unit 4, the thickness of the SFP floor is about 1.5 m (about 5 ft) which is less than the 

thickness of the floor of the SFP considered in this study which is about 1.83 m (about 6 

ft). 

• Available information indicates that the reinforcement of the wall of the SFP in Unit 4 is not 

significantly different from the reinforcement in the wall of the SFP considered in this 

study. 

• Although no reference is made to out-of-plane shear reinforcement for Unit 4 of 

Fukushima Daiichi, it is not known with certainty at the time of the writing of this report if 

out-of-plane shear reinforcement was provided at the edges of the floor slab of Unit 4 or at 

the intersection of this floor slab with the vertical walls. 

• For the SFP of Unit 4 no reference is made to a grid of steel beams analogous to that 

embedded in the floor and bottom of the walls of the SFP considered in this study (used to 

support the weight of wet concrete during construction). 

• Cross section drawings of the reactor building for Unit 4 indicate the possibility of a load 

bearing wall under the South wall (with reference to Figure 17) of the SFP of Unit 4, which 

does not exist for the SFP considered in this study. This difference, if confirmed, would 

result in a longer span for the entire structure of the SFP considered in this study.

Although there are differences between the structures of the Unit 4 SFP and the SFP 

considered in this study, these differences do not seem to be sufficiently significant to assert 

without further analysis that the Unit 4 SFP would be stronger for the same seismic demands 

than the SFP considered in this study. Differences in the vertical and horizontal response 

spectra at the foundation of Unit 4 and at the base of reactor building considered in this study (a 

fixed base structure) (see Figure 32 and Figure 31) indicate that the seismic forces for Unit 4 

would have been significantly less than those considered in this study. The difference between 

these seismic demands would have been the main factor affecting the relative performance of 

the Unit 4 SFP (under the March 11, 2011 earthquake) and the performance of the SFP 

considered in this study under the hypothetical beyond design basis earthquake. 

Major observations from these comparisons are: 

• For the challenging events that affected 20 reactors and SFPs, leakage from the bottom of 

the SFPs of the 20 BWR reactors was not reported. This is consistent with the highest 

relative likelihood estimate for this study being that for the state with no leakage. 

• Possible differences in the design and construction of the reactor buildings and SFPs, 

which considered higher design-basis seismic loads, and the SFP considered in this study,

introduces uncertainties in these observations.

• The ground motion used in this study may be more challenging for the spent fuel pool 

structure than those experienced at the Fukushima Daiichi nuclear power plant from the 

earthquake that occurred on March 11, 2011, off the coast of Japan, which did not cause 

spent fuel pool leaks at the bottom of the walls. 
76
77

5. SCENARIO DELINEATION AND PROBABILISTIC 

CONSIDERATIONS

5.1 Representative Operating Cycle Characterization

This section captures initial and boundary conditions related to the assumed operating cycle, as 

well as other related assumptions about the contents and layout of the SFP. Specifically, Table 

15 captures these boundary and initial conditions for the high-density loading configuration and 

the alternate low-density loading configuration. Information about the operating cycle length 

and outage length are based on averages of this information for the last five operating cycles at 

the reference plant.

Table 15 Remaining Boundary and Initial Conditions

Item High-Density Loading

Low-Density Loading (if 

different)

General:

Operating cycle duration 23 months —

Rack geometry:

Support leg height

Cell pitch

Open vs. closed cell

# of storage locations

18.41 cm (7.25 in.)1

15.95 cm (6.28 in.)

Closed cell2

3,819

—

—

—

Fuel loading

 Min. assem. during outage3

 Max assem. during outage

 # of assem. after outage

 Newer fuel (<5 years)

 Older fuel (>5 years)

 Pattern for “hot” fuel

 Coherent downcomer area7

3,819 – 764 – 284 = 2,771

3,819 – 764 = 3,0554

3,819 – 764 = 3,055

GE14/GNF25

Actual, based on 2003 info.

prearranged in 1x46

Yes

284 × 2 = 568

284 × 3 = 852

284 × 3 = 852

—

N/A

1x4 “with empties”

—

Outage specifications:

Shuffling vs. full core offload

Removal of weir gate

Start of defueling

Completion of defueling

Start of refueling

End of refueling

Replacement of weir gate

End of outage

Cycle length

Shuffling (roughly 1/3 core)8

2 days (after subcriticality)

2 days

8 days

14 days

20 days

20 days (modeled as 25 days)

25 days

700 days (23 months)9

—

—

—

—

—

—

—

—

— 1 Later in the conduct of the study the authors became aware that the distance between the pool floor liner 

and the bottom of the rack baseplate is actually (on average) closer to 26 centimeters (cm) (10.25 in.), 

depending on adjustments made to the leveling pad during installation. For the cases studied in this report, 

in which the leakage location is at the junction of the floor and side wall, side calculations have shown that 

the results are insensitive to this difference (i.e., even at 18 cm sufficient cross-sectional flow area exists to 

accommodate natural circulation flow). Nonetheless, any future analyses for this site (particularly if they 

involve leak locations on the pool side wall), should consider correcting this error. 2 This terminology refers to a rack design in which the sides of the rack cells have panels that inhibit or 

prevent cross-flow, while being relatively open at the top and bottom for axial flow. 3 It is assumed that a full core offload capability (an industry commitment as opposed to a regulatory 

requirement) is maintained. Further, it is assumed that 284 assemblies are offloaded each outage (roughly 
78

37 percent of the core) based primarily on the information in Exelon (2011), with a slight change from 270 to 

284 assemblies for MELCOR modeling convenience. 4 Sixty of these rack locations may be reserved for storing guide tubes. The study does not address this

situation, but it is expected to have a very minor effect on the results. By assumption, these 60 rack cells 

are filled with very low decay heat fuel, and represent less than 2 percent of the overall SFP inventory (and 

less than 2 percent of the radionuclide inventory available for release). 5 See Exelon (2011) for more information. 6 See Section 9.3 of this report for a discussion of how the use of contiguous (uniform) patterns would affect 

the results. 7 This term is used to describe whether an open area exists within the pool (such as an unracked region, a 

cask laydown area, or large gaps between the edge of the racks and the pool walls) that would facilitate 

downflow during conditions resulting in natural circulation air flow under the rack baseplate. 8 Note that the decay heat from the fuel left in the reactor is considered when the pool and reactor well are 

hydraulically connected. 9 After results were calculated based on a 700 day operating cycle, the authors realized that the correct 

operating cycle length should be 725 days (including the 25 day outage) rather than 700 days (which didn’t 

include the outage). This error is expected to have a small impact on the overall results.

The above table depicts a 1x4 storage pattern for the recently discharged fuel, based on the 

approach PBAPS has taken to meet the requirements associated with license condition 2.C.(11) 

and 10 CFR 50.54(hh)(2). The plant studied actually currently utilizes a 1x8 pattern. Because 

this pattern is believed to be atypical (relative to the fleet), it is not modeled as the base case in 

this study. Section 9.2 of this report provides additional analysis that shows the benefit of the 

1x8 pattern. Section 9.3 discusses of how the use of contiguous (uniform) patterns would affect 

the results. Figure 34 illustrates the different patterns. 

Figure 34 Illustration of SFP patterns

From left to right: Uniform/contiguous; 1x4 (used as base case);

1x8 (actually used by plant as of May 2012)

Red = a recently discharged assembly; Blue = an older, lower decay heat assembly, black outline shows repeating 

pattern

5.2 Operating Cycle Phase Specification

As described in Section 1.5, constant changes to the conditions in the SFP affect the 

consequences of a postulated accident (e.g., changes in the decay heat, changes in the 

inventory of fuel in the pool). Thus it is necessary to discretize this continuous behavior into a 

manageable set of discrete quasi-steady snapshots. Further, it must be recognized that the 

number of quasi-steady snapshots (or OCPs as they are termed throughout this report) has 

roughly a linear scaling effect on the number of MELCOR analyses that must be performed. As 

such, defining the OCPs becomes a minimization/optimization problem (i.e., the analysis needs 
79

to minimize the number of OCPs while optimizing the resulting OCPs’ accuracy in representing 

the above pool-reactor configurations/spent fuel loading configurations/decay heat levels).

Based on these considerations, timing associated with the movement of fuel and key changes in 

plant configuration were combined with the peak assembly and whole pool decay heat curves to 

arrive at a set of five OCPs, as outlined in Table 16.

Table 16 OCP Definition for the Modeled Operating Cycle

O

C

P 

#

OCP 

Description

Time 

(d)

% of 

opera

-ting 

cycle

Pool-reactor 

configuration

Modeled 

spent fuel 

config. 

for highdensity 

loading

Total decay 

power

Peak assembly 

power

1

Defueling of 

the reactor 

(~ 1/3 core)

2–8 0.9 Refueling 1x4

Existing1 + 

(27% of 

offloaded 

assemblies) @ 

4 days2

Highest powered 

offloaded 

assembly @ 4 

days2

2

Reactor 

T&M / 

inspection 

and 

refueling

8–25 2.4 Refueling 1x4

Existing1 + 

(offloaded 

assemblies) @ 

13 days2

Highest powered 

offloaded 

assembly @ 13 

days2

3

Highest 

decay power 

portion of 

nonoutage 

period

25–60 5 Unconnected 1x4

Existing1 + 

(offloaded 

assemblies) @ 

37 days2

Highest powered 

offloaded 

assembly @ 37 

days2

4

Next highest 

decay power 

portion of 

nonoutage 

period

60–

240 25.7 Unconnected 1x4

Existing1 + 

(offloaded 

assemblies) @ 

107 days2

Highest powered 

offloaded 

assembly @ 107 

days2

5

Remainder 

of operating 

cycle

240–

700;

0–2

66 Unconnected 1x4

Existing1 + 

(offloaded 

assemblies) @ 

383 days2

Highest powered 

offloaded 

assembly @ 383 

days2

1 The term “existing” refers to the fuel residing in the SFP at t = 0 (before offload). 2 These times are based on mean decay heat load (as opposed to mean time) during the specified phase 

(see text for additional discussions); time zero is set to the time of reactor shutdown

The following key assumptions in the above OCP definition warrant highlighting:

• The study does not explicitly treat the offloading of older fuel into casks (as part of the 

normal fuel management practices as opposed to an expedited fuel movement 

program). Rather, a stylized assumption is made that the 284 assemblies that would be 

loaded into dry casks during the operating cycle are instantaneously removed from the 

pool just before the outage.

• The study does not treat new fuel. This fuel would be placed into the SFP just before 

the outage (the subject plant does not use a separate new fuel vault). Thus, the fuel 

would only be present for a very short portion of the operating cycle. During the time 

that the new fuel is in the SFP, it would affect the amount of zirconium available to 
80

participate in a propagating zirconium fire, but would have a negligible effect on the 

source term.6

• The actual time at which the snapshots are evaluated is based on the mean decay heat 

during the OCP, as opposed to the mean time. Recall that the OCP snapshots are 

intended to represent a continuous function of possible consequences. While the 

likelihood of a seismic event occurring is constant in time within one of these OCPs, the 

consequences associated with the event are not. Furthermore, the exponential decay 

heat function better represents the change in the post-accident timeline within an OCP 

than does a linear function, and provides a better mean estimate of the OCP’s expected 

consequences. Therefore, the exponential functional form is used to determine the time 

within the OCP that is used for the quasi-steady evaluation. In the case of OCP1, a 

minor adjustment is made from 4.4 to 3.9 days for modeling convenience (the model is 

nodalized such that having 88 recently discharged assemblies can be more readily 

represented, and 3.9 days is the point at which this many assemblies would have been 

offloaded given the outage assumptions previously discussed).

5.3 Treatment of Mitigation

One of the objectives of this study is to provide insights into the effectiveness and benefits of 

mitigation measures currently employed at nuclear power plants. In addition to the redundant

and diverse physical systems designed to prevent severe accidents, NRC requires plant owners

to have preplanned emergency measures in the unlikely event an accident occurs. When they 

are successfully implemented, NRC expects these emergency measures will mitigate accident

consequences by preventing, delaying, or reducing a potential release of radioactive material 

from the SFP. These measures include a site-specific emergency plan, emergency operating

procedures, severe accident management guidelines, and 10 CFR 50.54(hh)(2) mitigation

measures put in place to respond to the loss of large areas of the plant due to fires or 

explosions. NRC requires its licensees to train and practice emergency measures to ensure 

that they have proper equipment, procedures, and training. NRC inspectors periodically 

observe these activities to help ensure that NRC regulations are met at each plant. The study 

assumes that the licensee’s emergency response organization would implement these 

measures in accordance with approved emergency plans, procedures, and guidelines.

Regarding onsite mitigative actions, the assumptions chosen by the project team to define the 

scenarios analyzed using MELCOR and MACCS2 are described here. Two cases are modeled for 

each scenario, a mitigated case and an unmitigated case. In the mitigated case, the model 

includes what would happen if the operators are fully successful in carrying out onsite mitigating

actions. However, NRC analyzes extreme events to gain insights on the safety margin provided 

by NRC’s regulatory framework. The uncertainties associated with the response to a beyonddesign-basis seismic event, and the resultant effects on the SFP, make consideration of 

unmitigated scenarios prudent from an informed decision-making standpoint. Thus, each 

scenario is also analyzed assuming that the operators are not successful in implementing onsite 

mitigating actions.

 6 The radioactive material that is of concern during an accident is the fission products generated while the fuel 

is in the reactor. The uranium dioxide (UO2) present in fresh fuel would not contribute noticeably to the 

source term, and in particular, not in a SFP accident in which the temperatures during a postulated accident 

are lower than those during a reactor accident.
81

In the unmitigated case, all onsite mitigative actions are unsuccessful for an extended period of 

time, meaning that there is no credit for repair or recovery of damaged systems (e.g., offsite 

power) and no credit for successful deployment of 10 CFR 50.54(hh)(2)7 equipment. The cases 

which assume lack of successful mitigation are presented to (1) acknowledge uncertainties in 

the effectiveness of these efforts during a beyond-design basis event and (2) demonstrate the 

effectiveness of successful mitigation. Section 5.3.2 of this report discusses further the 

rationale for developing results for this situation.

In the mitigated case, (1) mitigative actions associated with the regulatory requirements of 

10 CFR 50.54(hh)(2) are successfully deployed, (2) additional onsite capabilities are used to 

extend the use of this equipment, and (3) arrival of offsite resources allows this equipment to be 

utilized for an extended period of time (e.g., days) until onsite capabilities can be recovered. 

This study’s original scope did not include an attempt to quantify the likelihood of successful 

execution of different mitigative actions that might take place (e.g., makeup using a portable 

pump, recovery of ac power). Subsequent to completion of the MELCOR (Chapter 6) and 

MACCS2 (Chapter 7) analyses described in the following chapters, the project staff performed a 

human reliability analysis for the purpose of providing context regarding human response. The 

HRA results provide informative data to gain insights on the likelihood of mitigation being 

successfully implemented as well as possible regulatory enhancements for consideration. 

Chapter 8 describes the HRA. Since the HRA was performed after the bulk of the analysis was 

completed, some of the assumptions differ from those described in this Section.

In addition to onsite mitigation, offsite support is considered in the paragraphs below.

The reference plant is supported by an offsite emergency operating facility (EOF). The 

emergency response organization at the EOF has access to fleetwide emergency response 

personnel and equipment, including the 10 CFR 50.54(hh)(2) mitigation measures and 

equipment from the sister plants. Every licensee participates in full onsite and offsite exercises 

every 2 years where response to severe accidents and coordination with offsite response 

organizations is demonstrated and inspected by the NRC and the Federal Emergency 

Management Agency. In addition, the Institute for Nuclear Power Operations and the Nuclear 

Energy Institute would activate their emergency response centers to assist the site as needed.

Concurrent with the industry response, the U.S. National Response Framework (NRF) would 

establish a coordinated response of national assets. As described in the Nuclear/Radiological 

Incident Annex to the NRF, the NRC is typically the Coordinating Agency for incidents occurring 

at NRC-licensed facilities. As Coordinating Agency, the NRC has technical leadership for the 

Federal Government’s response to the incident. The NRF conducts periodic exercises and 

provides access to the full resources of the Federal Government. The NRC has an extensive, 

well-trained and exercised, emergency response capability and has onsite resident inspectors. 

The NRC would activate the incident response team at the NRC regional office and 

Headquarters. The focus of the NRC response is to ensure that public health and safety are 

protected and to assist the licensee with the response.

 7 This section of the regulations deals with the development and implementation of guidance and strategies 

intended to maintain or restore core cooling, containment, and SFP cooling capabilities under the 

circumstances associated with loss of large areas of the plant resulting from explosions or fire.
82

However, for the large beyond-design-basis seismic event under consideration in this study, it is 

possible that significant damage to local infrastructure could occur, requiring emergency 

resources to also be needed in other areas. Additionally, radiation and other hazards 

(discussed in Section 5.3.2 of this report) could hinder access to the SFP and key equipment, 

making prevention or truncation of an ongoing SFP release challenging.

Considering the uncertainties associated with this event as described above, project staff chose 

a 72-hour time truncation (assumed that the event would be terminated by some means by 72 

hours after initiation). The use of a time truncation is a point of uncertainty that can significantly 

affect the results and is further analyzed in Section 9.8 of this report. Note that like other 

aspects of this study, the incorporation of ongoing changes in regulatory commitments related to 

offsite response emanating from the Japan Lessons Learned initiatives is beyond this study’s 

scope.

Regarding offsite support for these situations, the accident progression analysis assumes the 

following for the purposes of this study:

• Within 24 hours, offsite support arrives.

• Within 48 hours, actions are planned and equipment is staged.

• At 48 hours, if the fuel is not uncovered and the pool can be refilled with an injection rate 

of 500 gpm (which is true for the cases with no leak or a small leak), the sequence is 

truncated.

• Otherwise, the sequence is run to 72 hours because of the additional complexities of 

(1) accessing the area of the pool when the fuel is uncovered and stopping an excontainment release in progress and (2) performing a large leak repair.

These assumptions are similar to the assumptions used in NUREG-1935, “State-of-the-Art 

Reactor Consequence Analyses Project” (NRC, 2012i).
83

Table 17 summarizes each situation.

Table 17 Summary of Mitigation Assumptions

Item

Situations with successful 

deployment of onsite 

mitigation

Situations without 

successful deployment of 

onsite mitigation

Installed accident mitigation 

equipment Damaged by the event; recovery/repair not credited

10 CFR 50.54(hh)(2) 

equipment

Successfully deployed 2 hours 

after diagnosis Not credited

Other onsite resources

Successfully deployed to 

extend operation of 

10 CFR 50.54(hh)(2) 

equipment

Not credited

Offsite resources Successfully deployed for terminating the accident at 48 or 72 

hours (see Section 9.8)

Emergency preparedness Effective (see APPENDIX A: of this report for more details)

Mitigation equipment being 

considered under NRC Order 

EA-12-049, dated March 12, 

2012

Not considered; may be substantively similar to 

10 CFR 50.54(hh)(2) capabilities within the context of this 

study

5.3.1 Approach Details and Assumptions

Scenarios that credit successful deployment of the 10 CFR 50.54(hh)(2) measures must include

assumptions about how that deployment is executed. In general, this study utilizes some of the 

limits associated with these capabilities that are contained in Nuclear Energy Institute (NEI) 06-

12, Revision 2, “B.5.b Phase 2 & 3 Submittal Guidance,” issued December 2006 (which the 

NRC has endorsed8

). For instance, the time at which the mitigative capability is assumed to 

commence (meaning that it has been deployed and is starting to operate) is 2 hours after 

diagnosis. The guidance in NEI 06-12, Revision 2, does include a provision that allows for a 

deployment time of 5 hours after diagnosis for spray, if the fuel has been favorably configured. 

This study does not invoke that provision because the site in question strives to deploy the 

equipment within 2 hours regardless of the fuel pattern and the existence of cases without 

successful deployment of mitigation envelops this effect.

The flow rates associated with the two modes of delivery considered (spray and makeup) are 

assumed to be the minimum amounts required (200 gallons delivered per minute for spray and 

500 gallons delivered per minute for makeup). For PBAPS, the capacities of the available 

equipment are somewhat higher. The use of the 500 and 200 gpm values in this study attempts 

to account for uncertainties in the speed at which the pumps would actually be run, as well as 

 8 The NRC originally endorsed this document for operating reactors by letter dated December 22, 2006 and 

this endorsement was carried forward in the Statement of Considerations for the associated rulemaking (see 

“Power Reactor Security Requirements, Final Rule,” published in the Federal Register on March 27, 2009). 

B.5.b refers to Section B.5.b of Order EA-02-026, dated February 25, 2002, and later made generically 

applicable in 10 CFR 50.54(hh)(2).
84

spray that goes outside the boundary of the pool.

9

 As a result, no additional “penalty” is given 

for inefficiencies associated with spray coverage (i.e., the spray flow rate is applied uniformly 

across the pool cross-sectional area without further reduction). In either spray or makeup 

mode, the licensee would utilize a portable diesel-driven pump to pump water from either the 

fire ring header, the intake canal, or the emergency cooling tower basin to the refueling floor via 

hoses that would be run up a reactor building stairwell.

The following set of criteria was established to model the time to diagnosis of the need to deploy 

10 CFR 50.54(hh)(2) mitigative strategies:

• no ac power

• SFP level decrease by 1.5 m (5 ft), keeping in mind that 0.5 m (1.5 ft) is lost because of

sloshing

• 30-minute delay associated with manual observation/decision-making

These criteria were developed with consideration of the plant-specific procedures for problems 

associated with the SFP, though these specific criteria do not exist in those procedures and they 

are not intended to represent a specific procedural pathway. It is also important to note that, for 

the plant studied, the various procedures related to loss of SFP cooling or loss of SFP inventory 

do refer plant personnel to the guidelines for use of the 10 CFR 50.54(hh)(2) equipment, even if 

the cause of the event is not a loss of large area of the plant. More specifically, if control room 

alarms are available, the loss of inventory would cause an alarm that would direct the operators 

to a local panel on the refuel floor. The alarm procedure would also start a procedural pathway 

that would explicitly lead to consideration of the use of the 50.54(hh)(2) equipment. If control 

room alarms are not available, the special event procedure related to an earthquake directs the 

operators to inspect the status of the SFP and its cooling systems. The special event procedure 

also triggers a procedural pathway that would explicitly lead to consideration of the use of the 

50.54(hh)(2) equipment. Note that the details of onsite response are covered more thoroughly 

in Chapter 8.

The above criteria could be conservative or nonconservative depending on the priorities of 

operators, and different criteria would clearly be more applicable to other scenarios, particularly 

those that did not include loss of offsite and onsite power at time zero. The assumption that 

pool elevation must drop 5 ft can lead to long diagnosis time periods for slowly progressing 

events, thus leading to a potentially conservative timeline for mitigative action. However, these 

same slowly developing scenarios are the ones that are least important for offsite 

consequences (i.e., are less severe and less likely to lead to a release). The use of a 2-hour 

deployment time, as opposed to a 5-hour deployment time allowed in some situations, has a 

compensating effect for some scenarios. Chapter 8 discusses the issue of diagnosis in greater 

detail.

 9 MELCOR does not model the details of the spray delivery from the nozzle(s) to the SFP. Rather, it assumes 

a uniform flux of water at the top of the SFP. A system flow rate of greater than 200 gpm is necessary to 

achieve this uniform 200 gpm-equivalent spray flux, to allow for water striking the pool deck or walls and not 

entering the pool. The regulatory implementation of 10 CFR 50.54(hh)(2) accounts for this inefficiency 

effect.
85

Regarding the implementation mode, for cases in which the water level in the pool is greater 

than 0.9 m (3 ft) above the top of the racks (a surrogate for high radiation levels on the refueling 

floor near the edge of the SFP (see Section 5.4 of this report)) at the earliest time the 

sprays/makeup are ready for initiation (i.e., 2 hours after diagnosis), makeup will be utilized. 

Otherwise, sprays will be utilized. This represents one possible approach to the decision point 

in Figure 2-1 of NEI 06-12, Revision 2 (NEI, 2006), regarding whether SFP leakage is 

excessive. In some respects it is a more complicated approach than might be used, but is 

arguably a more straightforward approach to enact in the absence of instrumentation. In 

practice, both approaches end up prompting the same implementation mode for most scenarios 

studied in this report. The exception is for the “moderate” hole for OCP1/2, in which, because of

the larger volume of water since the SFP is connected to the reactor well and separator/dryer 

pool, the water level has not reached the 3 ft mark (above the top of the racks) by the time 

mitigation is deployed. In these cases, makeup is deployed even though the leakage rate 

actually exceeds 500 gpm. Section 9.3 of this report investigates the effect of this assumption 

for a uniform pattern.

Whichever mode is initiated (spray versus makeup), it is assumed to be used for the duration of 

the event (i.e., no later switching to a different mode). For OCP 1/2 with the “moderate” leakage 

condition, makeup is deployed. Other, equally reasonable assumptions about mitigation 

deployment could result in the deployment of sprays instead (which have a potential advantage 

in terms of mitigation for these conditions). Section 9.3 presents a sensitivity study related to 

this assumption, for a uniform pattern.

Practically speaking, the above set of assumptions leads to the following process when 

establishing mitigation timeline boundary conditions in the MELCOR analyses (recall that this 

only applies for half of the studied sequences since each scenario has a calculation without 

successful deployment of mitigation):

• Start of calculation/earthquake occurs.

• When SFP level has decreased by 1.5 m (5 ft), and 30 (diagnosis delay) plus 120 (initial 

deployment delay) additional minutes have transpired, then the following applies:

- If the water level is greater than 0.9 m (3 ft) above the top of the fuel, then 

500 gpm of makeup into the top of the pool commences.

- If the water level is less than 0.9 m (3 ft) above the top of the fuel (thus indicating 

excessive leakage) then 200 gpm of spray at the top of the pool commences.

The above assumptions are characterized as optimistic relative to the unmitigated (pessimistic)

case. However, it is important to note that aspects of these assumptions assume failures where 

they may not occur. For instance, the above set of assumptions only credits a single successful 

spray/makeup strategy, whereas multiple strategies may be deployed. Along these lines, there 

are several other ways to recover makeup to the SFP, several of which have much higher 

capacities than the mode selected. Table 10.3.1 of the FSAR captures these alternatives, 

which range from capacities of 25 gpm to 18,000 gpm. For each of the modes capable of 

delivering more than 200-500 gpm (the mode selected in this study), these modes require either 

multiple manual alignments in the vicinity of the SFP and reactor, the availability of ac power for 

valve manipulations, or the use of equipment that might be involved in reactor recovery (most 

notably a residual heat removal pump), as well as ac power for pump operation. Finally, as 
86

mentioned previously, the selected set of assumptions does not allow for switching from one 

mode of makeup/spray to the other.

5.3.2 Rationale for Producing Unmitigated Results

NRC licensees that operate nuclear power plants are required to maintain the facility in a 

manner that makes the occurrence of a severe accident unlikely. This is achieved through a 

number of mechanisms involving facility design and operator training, and by applying the 

concept of defense-in-depth. Even so, uncertainties associated with the response to a beyonddesign-basis seismic event, and the resultant effects on the SFP, make consideration of 

unmitigated scenarios prudent from an informed decision-making standpoint. Some specific 

issues relevant to the situation considered in this report include the following:

• The regulatory requirements for 10 CFR 50.54(hh)(2) equipment currently focus on the 

use of this equipment for responding to a loss of a large area of the plant from explosion 

or fire. Ongoing regulatory activities related to the NRC’s response to the March 2011 

accident at the Japanese Fukushima-Daiichi site will alter this situation (e.g., see NRC 

Order EA-12-049, dated March 12, 2012). Note that some plants (including the 

reference plant) have already acquired some additional equipment in anticipation of this 

requirement, with full compliance scheduled for 2016.

• The large seismic event could damage onsite (and offsite) infrastructure designed to 

facilitate accident response, as well as cause general disruption at the site.

• If circumstances led to the uncovery of fuel in the SFP, radiation fields on the refueling 

floor might hamper mitigative actions. Section 5.4 of this report describes the shielding 

analyses that inform this aspect of the accident analysis. Chapter 8 further discusses 

accessibility issues in the context of human response. Note that, as part of the 

implementation of 10 CFR 50.54(hh)(2), the licensee has committed to an ability to carry 

out the required mitigative actions even in such situations (e.g., using portable shielding 

or implementing from a location other than the refueling floor itself).

10

• A concurrent reactor event (resulting from the loss of ac power or other damage), or an 

ongoing accident at the other unit’s SFP, could hamper mitigative actions by reducing 

accessibility because of radiation fields, impeding accessibility because of other hazards 

such as hydrogen accumulation, or diverting resources (both personnel and equipment). 

Chapter 8 discusses this issue further.

• An assembly being moved within the SFP (or from the reactor to the SFP or vice versa) 

at the time of the event, could lead to an earlier radiological hazard for responders, if this 

assembly were to become uncovered earlier in the event progression, because of its 

higher position in the SFP. Section 5.4 of this report provides refuel floor dose rate 

estimates for this situation.

 10 The industry’s FLEX proposal, developed in response to NRC Order EA-12-049, includes a specification for 

a means to connect makeup to the installed SFP cooling system to overcome the potential for lack of access 

to the SFP deck area. This is primarily to address the potential needs for makeup in a saturated condition 

caused by boil off for an uncooled pool.
87

• Accessibility could be reduced if an inadvertent criticality event in the SFP were to occur. 

See Section 2.3 of this report for more information about inadvertent criticality events.

For these reasons, this study presents results for cases in which accident mitigation efforts are 

unsuccessful for some period of time.

5.4 Refueling Floor Dose Rate Analysis Using SCALE

This study included analyses to predict the radiological conditions on the refuel floor for a range 

of conditions associated with loss of water in the SFP. Note that the analyses described in this 

section only account for the radiological conditions stemming from neutron and gamma “shine” 

from exposed radioactive material and do not account for the concern of radiological conditions 

associated with the release of that material following fuel heatup. It is expected that, if a 

radiological release of fission products from the SFP were to commence, radiation fields in the 

vicinity of the pool would be extremely high.

The analyses described, which Oak Ridge National Laboratory (ORNL) performed, looked at a 

range of conditions. This range included both the high-density and low-density loading 

conditions studied in this report, as well as the situation in which a single assembly is in the 

lifted position at the time of the event. The times following discharge that were considered are 

the same as those associated with the different OCPs. This portion of the analyses is plant

specific for the reference plant, and utilized 2011 vintage information for representing the fuel 

design and characteristics in the SFP. Calculations were performed using the ORIGEN and 

MAVRIC modules of the SCALE code suite. MAVRIC in turn used BONAMI, CENTRM, 

DENOVO, and Monaco routines, along with the FW-CADIS methodology. The analysis used 

the flux-to-dose conversion factors in American National Standards Institute/American Nuclear 

Society (ANSI/ANS) 6.1.1-1977. The 200 neutron group and 47 gamma group cross sections 

based on the ENDF/B-VII.0 cross-section library distributed with SCALE 6.1 were used.
88

Figure 35 Cutaway depiction of a lifted assembly with water level at the top of the racks

Results of the analyses for the high-density loading situation can be summarized as follows:

• For water depths of 3 m (10 ft) above the top of the racks, projected dose rates are very, 

very low (less than 0.1 millirem (mrem) per hour). This is consistent with Regulatory 

Guide 1.13, “Spent Fuel Storage Facility Design Basis,” which uses this water depth as a 

conservative measure of adequate shielding.

• Dose rates for the maximally exposed location on the refueling floor once the water level 

drops to the top of fuel hardware are very high (on the order of 450 to 600 rem per hour, 

depending on the OCP).

11

• At a water depth of 0.6 m (2ft) above the top of the fuel, the projected dose at the 

maximally exposed location on the refueling floor surpasses 25 rem in one hour. 25 rem 

is the value above which actions can be taken to save lives or protect large populations, 

on a voluntary basis, as defined in Table 2-2 of U.S. Environmental Protection Agency 

(EPA) 400-R-92-001, “Manual of Protective Action Guides and Protective Actions for 

Nuclear Incidents,” issued May 1992)

 11 This range shows that, for situations in which the water level is at the top of the fuel hardware, the dose 

rates are somewhat sensitive to the time during the operating cycle (a 33-percent decrease in this case). For 

instances in which water is covering the fuel hardware, this sensitivity decreases. For example, the 

analogous range of values for a water level 100 cm (3.3 ft) above the fuel hardware is 1.6 to 1.7 rem per 

hour.
89

• Dose rates elsewhere on the refueling floor are significantly lower than those at the 

maximally exposed location (e.g., see Figure 36).

Figure 36 Approximate dose rate of elevation contours, water at top of fuel hardware, 

around the time of defueling (rem per hour).

Relative to the high-density loading situation, the other situations can be compared as follows:

• For low-density loading situations, dose rates for the maximally exposed location on the 

refueling floor once the water level drops to the top of fuel hardware are lower than the 

high-density loading case, but still very high (on the order of 300 to 470 rem per hour, 

depending on the OCP).

• For a recently discharged assembly in the lifted position, dose rates for the maximallyexposed location are on the order of 3 rem per hour when the water level is at the top of

the lifted assembly and over 1,000 rem per hour when the water level is at the top of the 

racks (i.e., when the lifted assembly is completely exposed). These are dose rate 

Storage 

Spent Fuel 

Reactor

300-

100-300

30- 10-

1- 3-

0.3-
90

contributions solely from the lifted assembly (i.e., they are in addition to the dose rate 

contributions from the assemblies in the racks).

• For an older assembly (discharged more than a decade previously) in the lifted position, 

dose rates for the maximally exposed location are on the order of 0.2 rem per hour when 

the water level is at the top of the lifted assembly and 7 rem per hour when the water 

level is at the top of the racks. Again, these are dose rate contributions solely from the 

lifted assembly.

The high dose rates associated with the single lifted assembly (particularly those for the recently

discharged assembly) are sensitive to the assumed position of that assembly. This case 

assumes that the assembly is located somewhere near the middle of the pool (see 

Figure 35), which results in direct line-of-sight from the edge of the SFP. Placement near a wall 

would reduce the dose rate for locations near the edge of the pool that do not have a direct line 

of sight to the assembly.

5.5 Discussion of Repair and Recovery

This study makes no attempt to account for repair or recovery of onsite equipment or offsite 

power. This is a simplifying assumption, and is motivated in part by the lack of quantitative 

information available to support such a determination for the large seismic event being 

considered. Procedures would direct the operators to attempt to recover failed equipment and 

pursue alternate means of establishing ac power, such as the ability to obtain ac power from an

SBO cross-tie line to the Conowingo Dam. The study assumes that the damage sustained by 

the onsite and offsite electrical distribution systems from the earthquake is enough to 

significantly delay these recoveries until after the 48- or 72-hour truncation times.

That being said, and as covered previously in this section, the scenarios with successful 

deployment of mitigation do assume that onsite and offsite resources are able to extend 

operation of the 10 CFR 50.54(hh)(2) equipment indefinitely, which could represent a situation 

in which ac power is recovered at an intermediate point and ac-dependent means of SFP 

makeup are brought back online.

5.6 Scenario Development

5.6.1 Identification of Key Events

The scenario development included the following major assumptions based on the structural 

analysis documented in Section 4 of this report or other considerations:

• All offsite and onsite ac power is lost as a direct result of the seismic event (see Section

4.2 of this report).

• Direct current power may be lost. Because of the difference from the reactor situation 

(in which dc power to control turbine-driven systems is important in an SBO), the 

availability or unavailability of dc power has a much narrower effect. For the specific set 

of assumptions used in the MELCOR and MACCS2 analyses, there is no effect as 

analyzed. Chapter 8 further discusses this issue with respect to the HRA.
91

• The 10 CFR 50.54(hh)(2) equipment (when credited) is available for the duration of the 

event, following delays associated with diagnosis and deployment (see Section 5.3.1 of 

this report).

• Initial water loss from “sloshing” will be 0.5 m (1.5 ft) (see Section 4.2 of this report).

• Tearing of the SFP liner is not the most probable outcome, but is possible (see Section 

4.1 of this report).

• There is no failure of penetrations, including the refueling transfer canal gate (see 

Section 4.2 of this report).

• The overhead structures (building debris, crane) do not pose a threat to the SFP in 

terms of failure resulting from the initiating event (see Section 4.2 of this report).

• Inadvertent criticality, including seismic effects on the integrated poison rack material, is

not treated (see Section 2.3 of this report).

5.6.2 Scenario Calculation Matrices

The following table shows how the combinations described thus far translate to the scenarios 

considered for each OCP.

Table 18 Scenario Breakdown per OCP

Case #

Scenario Characteristics Radioactive Release Commences before 72 

Hours?

SFP Leakage 

Rate? Mitigation? High-Density Loading—1x4 Low-Density Loading

1 None Yes

See later sections of the report for results

2 No

3 Small Yes

4 No

5 Moderate Yes

6 No

5.6.3 Summary of Event Split Fractions

As described previously, the analysis considered the available seismic hazard information to 

obtain an initiating event frequency of approximately one event in 60,000 years for the reference 

plant.

Table 19 Refresher on the Seismic Hazard Estimates

Seismic 

Bin #

PGA 

Range 

(g)

Geometric 

Mean 

Accel. (g)

Likelihood 

based on 

PGA (yr)

Likelihood 

based on 

PGA (/yr)

Potential for damage to 

SFP liner?

1 0.1 to 

0.3

0.2 1 in 2,000 5.2×10-4 Damage not expected

2 0.3 to 

0.5

0.4 1 in 40,000 2.7×10-5 Damage not expected

3 0.5 to 0.7 1 in 60,000 1.7×10-5 Damage possible
92

1.0

4 > 1.0 > 1.0 1 in 200,000 4.9×10-6 Damage possible

Regarding the probability of losing ac power from this particular seismic event, the results 

described earlier in this report are summarized below.

Table 20 Refresher on ac Fragility

Item

Relative 

Likelihood Comments

Loss of normal 

SFP cooling 0.84

This study used the ac power fragility from NUREG-1150 of 

0.84 as a surrogate for the conditional probability of normal 

SFP cooling and makeup not being available. This 

simplifying assumption was made in light of the fact that the 

study is not a PRA (but rather a consequence analysis with 

probabilistic considerations) and that this value already 

approximates the upper bound of 1. In reality, the 

availability of normal SFP cooling and makeup would be a 

combination of the AC fragility, the fragility of the actual 

equipment and its support equipment, and operator actions 

to recover the equipment, which could result in a 

conditional probability higher than the value used here.

As described previously, the structural assessment led to the SFP leakage estimates stated 

below.

Table 21 Refresher on SFP Leakage Conditional Probabilities

Damage State

Relative 

Likelihood Comments

No leakage 0.9 Significant damage to concrete; no rupture of SFP liner

“Small” leakage 0.05 Small rupture of SFP liner; drains pool in tens of hours

“Moderate”

leakage 0.05 Tearing of SFP liner; damaged concrete limits outflow; drains pool within ones of hours

Finally, since a seismic event is equally likely to happen throughout the operating cycle, the 

conditional probability for its occurrence during a specific OCP is simply the duration of that 

OCP divided by the duration of the operating cycle. These weights range from 1 percent for 

OCP1 to 66 percent for OCP5 (recall that the OCPs were intentionally “front loaded” because 

the most change in SFP conditions occurs during the outage).

Table 22 Refresher on the OCP Fractional Contributions

OCP 

#

Time window (Time of 

evaluation) (in days)

Fraction of 

operating 

cycle

Pool-reactor 

configuration

Spent fuel 

configuration for 

high-density loading

1 2–8 (5) 0.01 Refueling Dispersed (except for 

2 8–25 (13) 0.02 Section 9.3)

3 25–60 (37) 0.05

4 60–240 (107) 0.26 Unconnected Dispersed

5 240–700 and 0–2 (383) 0.66
93

The above conditional probabilities are combined, algebraically, to provide likelihoods 

associated with each of the different sequences treated. At times, sequences are grouped 

(e.g., those that lead to a release versus those that do not), so as to assign scenario-specific 

release frequencies, scenario-specific individual risk of an LCF, or the like. It is important to 

keep in mind that all such frequencies only consider the particular large seismic event studied in 

this report.
94
95

6. ACCIDENT PROGRESSION ANALYSIS

6.1 Modeling Spent Fuel Pools with MELCOR

6.1.1 Overview and Experimental/Analytical Basis

The MELCOR computer code (Gauntt, 2005) represents the current state of the art in severe 

accident analysis. MELCOR has been developed through the NRC and international research 

performed since the accident at Three Mile Island in 1979. MELCOR is a fully integrated, 

engineering-level computer code and includes a broad spectrum of severe accident phenomena 

with capabilities to model core heatup and degradation, fission product release and transport 

within the primary system and containment, core relocation to the vessel lower head, and 

ex-vessel core concrete interaction.

The MELCOR code comprises an executive driver and a number of major modules, or 

packages, that together model the major systems of a reactor plant and their generally coupled 

interactions. The various code packages have been written using a carefully designed modular 

structure with well-defined interfaces between them. This allows the exchange of complete and 

consistent information among them so that all phenomena are explicitly coupled at every step. 

The structure also facilitates maintenance and upgrading of the code. Plant systems and their 

response to off-normal or accident conditions include the following:

• thermal-hydraulic response of the primary reactor coolant system, the reactor cavity, the 

containment, and the confinement buildings

• core uncovering (loss of coolant), fuel heatup, cladding oxidation, fuel degradation (loss of 

rod geometry), and core material melting and relocation

• heatup of reactor vessel lower head from relocated fuel materials and the thermal and 

mechanical loading and failure of the vessel lower head, as well as transfer of core materials 

to the reactor vessel cavity

• core-concrete attack and ensuing aerosol generation

• in-vessel and ex-vessel hydrogen production, transport, and combustion

• fission product release (aerosol and vapor), transport, and deposition

• behavior of radioactive aerosols in the reactor containment building, including scrubbing in 

water pools, and aerosol mechanics in the containment atmosphere such as particle 

agglomeration and gravitational settling

• impact of engineered safety features on thermal-hydraulic and radionuclide behavior

MELCOR modeling is general and flexible, making use of a “control volume” approach in 

describing the thermal-hydraulic response of the plant. No specific nodalization is provided, 

which allows a choice of the degree of detail appropriate to the task at hand. Reactor-specific 

geometry is imposed only in modeling the reactor core. The MELCOR code has been 

modernized (source code upgrade to Fortran95) to provide an efficient code structure for ease 

of maintenance, resulting in the release of MELCOR version 2.1. The new upgraded version of 

the code architecture supports advancements in computer hardware and software, and the code 

numerics improvements are underway to carry out more reasonable execution times. The input 

structure for MELCOR 2.1 differs completely from that of MELCOR 1.8.6. MELCOR is an ideal 

tool for this type of application because (1) its capabilities have been recently developed and 

validated for treating SFP accidents and (2) it is able to model the accident progression, 

radionuclide release, and in-building transport/retention. MELCOR 1.8.6 was used in the 
96

present study, and the SFP models in both versions of the code (1.8.6 and 2.1) are functionally

the same.

As part of NRC’s post-9/11 security assessments, the agency developed and applied SFP 

modeling using detailed thermal-hydraulic and severe accident progression models integrated 

into the MELCOR code to assess the realistic heatup of spent fuel under various pool draining 

conditions. SNL performed the analyses for a reference BWR, with additional supporting 

analyses for separate effects and fluid flow modeling using an earlier version of the code 

(MELCOR 1.8.5 Version RP) which is no longer maintained. Some of the modeling 

improvements in MELCOR 1.8.6 include revised modeling of the lower plenum to account for 

the curvature of the lower head (not relevant for an SFP) and formation and convection of 

stratified molten pools.

MELCOR 1.8.5 Version RP added two modeling enhancements applicable to BWR SFP 

modeling (also included in MELCOR 1.8.6 and 2.1): (1) a new rack component, which permits 

better modeling of an SFP rack and (2) a new oxidation kinetics model. The new BWR SFP

rack component permits proper radiative modeling of the SFP rack between groups of different 

assemblies. The new oxidation kinetics model predicts the transition to breakaway oxidation in 

air environments on a node-by-node basis. These new SFP features can be used to perform 

two types of SFP calculations: (1) a partial loss-of-coolant inventory accident and (2) a 

complete loss-of-coolant inventory accident. A complete loss-of-coolant inventory accident is 

characterized by the draining of the water to uncover the bottom of the racks leading to air 

circulation patterns inside the pool and associated air oxidation of the cladding (pre- and postbreakaway) and enhanced ruthenium release. A partial loss-of-coolant inventory or boiloff 

accident could involve no or late uncovery of the bottom of the racks. Boiloff of the coolant 

leads to steam generation and steam oxidation of the cladding and hydrogen generation that 

could lead to hydrogen combustion.

Breakaway Oxidation Model

Argonne National Laboratory (ANL) has performed oxidation kinetics testing on zirconium-based 

alloys, including Zircaloy-4 which is similar to the Zircaloy-2 alloy. The testing showed that air 

oxidation can be observed at temperatures as low as 600 K. In the tests, a specimen was held 

at constant temperature and the weight gain associated with oxidation as a function of time was 

measured. The reaction rates for air oxidation are described by parabolic kinetics similar to the 

ones used to describe steam oxidation. The general form of the equation is as follows:

 (1)

where, w is the reacted metal mass per unit surface area. The rate of oxidation was initially 

steady versus the square root of time at a particular temperature. However, the rate of oxidation 

increased after some time and persisted for the remainder of the test. The ANL pre- and postbreakaway Zircaloy-4 oxidation correlations are provided below.

The steam preoxidized, wide-temperature, prebreakaway Zircaloy-4 oxidation correlation 

(Natesan and Soppet, 2004) is as follows:

 (2)
97

The steam preoxidized, wide-temperature, postbreakaway Zircaloy-4 oxidation correlation 

(Natesan and Soppet, 2004) is as follows:

 (3)

The new oxidation model was implemented in MELCOR by adding a breakaway lifetime 

calculation. The model calculates an oxidation “lifetime” value for Zircaloy components in each 

cell using the local Zircaloy cladding temperature:

 (4)

 (5)

 (6)

where PLOX is the MELCOR fit of the timing for the transition from prebreakaway to 

postbreakaway oxidation reaction kinetics for Zirlo and Zircaloy-4 in the ANL experiments.

The air oxidation model was benchmarked against experimental data from the SNL SFP facility

as part of the security assessment work. The calculations with and without breakaway oxidation 

kinetics showed different heatup rates following breakaway. Both the data and the calculation 

with breakaway kinetics show a sharp increase in the heatup rate following breakaway. The 

new breakaway kinetics model provided a better prediction of the measured data, including a 

transition to accelerated postbreakaway oxidation kinetics.

Hydraulic Resistance Model

The MELCOR modeling approach for flowpaths connecting control volumes includes 

constitutive relationships to specify form losses (i.e., minor losses) and wall friction losses (i.e., 

major or viscous) along a flowpath as a hydraulic flow loss term to the momentum equation. 

The format of the user-specified input for MELCOR is defined from the sum of the local viscous 

and major pressure drops:

 (7)

where ρ is the fluid density, v is the fluid phase velocity, L is the inertial flow path length, D is a 

representative hydraulic diameter, and K is the form loss coefficient. The laminar friction factor 

(f) is given as:

 (8)

where SLAM is a user-specified MELCOR input parameter, Re is the Reynolds number (ρvD/μ), 

and μ is the fluid dynamic viscosity.

Hydraulic resistance measurements were performed on a Global Nuclear Fuel 9x9 BWR 

assembly at SNL (Durbin, 2005) to obtain the required frictional and form loss coefficients,
98

including the effects of grid spacer and partial rods. The present study used these 

measurements given a lack of data for a 10x10 BWR assembly.

6.1.2 Heat Transfer Modeling within Spent Fuel Pool and to Surrounding Walls

The MELCOR core models calculate the thermal response of the core.12 The core is nodalized 

into a number of axial levels and radial rings (each ring represents a collection of assemblies). 

All important heat transfer processes are modeled in each core cell, including thermal radiation 

within a cell and between cells in both the axial and radial directions, as well as radiation to 

boundary heat structures. Each core cell is hydraulically interfaced to a control volume to obtain 

the necessary boundary conditions (e.g., water level, flow velocity) and in turn supplies the 

calculated heat and mass transfer to the control volume. Each core cell may contain a number 

of components, including fuel, cladding, canister (BWRs), and other structures (e.g., control 

rods).

The new SFP rack component permits separate modeling of the SFP rack and radiative heat 

transfer between the rack and existing components in the core. The new air oxidation kinetics 

model predicts the transition to breakaway oxidation kinetics in air environments on a 

node-by-node basis. The SFP racks and the lower gap region below the SFP racks can be 

modeled using the existing core and lower plenum components. The MELCOR core model is 

designed in two-dimensional cylindrical geometry, and nodalization of the SFP must fit within 

this framework. Implicit in this framework is the assumed direction of heat and mass transfer 

between adjacent rings and adjacent elevations. For SFP models, the user can take advantage 

of this preexisting framework and arrange the fuel rack cells in a similar ring pattern. 

The heat transfer paths modeled within the core are appropriate for conventional commercial 

light-water reactors. The capability has been added to define arbitrary (“generalized”) additional 

heat transfer paths between core components to allow for more flexible intracell radiation or 

conduction, but the user is responsible for defining a single input parameter that captures the 

geometry of the heat transfer path. Figure 37 depicts the heat transfer paths within a ring and 

across a ring boundary. For radiation between different core rings, the user adjusts the view 

factors and the surface areas.

The core models radiative heat transfer from the outermost ring components (if present) to the 

core boundary specified as a heat structure. The SFP wall is modeled as a heat structure 

composed of a steel liner and concrete which can receive radiative energy from the core as well 

as convective heat transfer from the adjacent control volume.

 12 MELCOR core models were originally designed for the reactor core. Because of the code flexibility, the 

same modeling approach can be used for the spent fuel pool (with the addition of the rack as a separate 

component). Therefore, as far as code models are concerned (e.g., heat transfer between groups of 

assemblies and with the fluid, and radionuclide release, transport and deposition), there is no difference 

between reactor assemblies and spent fuel assemblies. It is up to the user to define the proper 

information in the input deck.
99

Figure 37 MELCOR modeling of heat transfer paths

6.1.3 Modeling of Mitigative Sprays

The MELCOR containment spray model was used to calculate the thermal response of the fuel

for the mitigated scenarios involving spray activation. The spray model mechanistically models 

the interaction of the spray droplets with the atmosphere and includes droplet heat and mass 

transfer and fission product removal capabilities. All calculations used a droplet size of 

1,250 microns. The spray was positioned at the top of the SFP (elevation of the refueling bay),

thus allowing the droplets to be directed into the assemblies and open spaces based on their 

respective cross-sectional areas. 

The interphase momentum model, which replicates the Wallis flooding curve, controls the 

penetration of the spray water into the assembly. Once the spray water enters the assembly, 

the spray is assumed to form a thin film on the fuel structures in the assembly, which drains 

downward. The MELCOR simplified flow regime model identifies the spray flow as a film in 

contact with the fuel rods (see Figure 38). Heat transfer takes place between the fuel rods and 

water in core cells where the flow regime model is active. Nucleate or film boiling heats the 

water film to saturation conditions as it drains down the assembly. Simultaneous heat transfer 

from the rods and surrounding gas causes the spray flow to boil. The spray film travels 

downward in contact with the fuel rods until the local control volume void fraction becomes 

greater than 99.8 percent (i.e., α > 0.998). Because of numerical considerations, the residual 

water is converted into a shallow pool where the liquid heat transfer area is apportioned by the 

depth of the pool in the control volume. Typically, the remaining water boils away in the first 

core cell after the flow regime model is disabled.

MELCOR thermal-hydraulic model interprets the liquid film as a small pool at the bottom of each 

control volume (see Figure 38). Because of the high void fraction, the phasic resistance of the 

steam or air flowing through the pool is relatively insignificant, which is the expected impact of a 

liquid film. Similarly, the depth of the spray water penetration is controlled by the heat transfer 

rate from the fuel rather than the momentum solution. Axial, stepwise heat transfer from the 

core cells limits how far the spray water penetrates into the assembly. A possible limitation of 

the thermal-hydraulic representation is the relatively small heat transfer area between the two 

phases (i.e., heat transfer through the pool and the surface versus a film). However, the rate of 

heat transfer from the gas to the water film is minor in comparison to the nucleate and film 

boiling heat transfer on the surface of the fuel rods. A detailed nodalization is used to track the 

water as it penetrates into the assembly which permits a better local representation of the fluid 
100

conditions and the location of the spray dryout. Parametric calculations are performed to show 

the impact of this modeling parameter (i.e., flow regime model active or inactive as discussed in 

6.3.1).

Figure 38 Spray model for SFP analysis

6.1.4 Modeling of Fuel Collapse and Baseplate Failure

Fuel collapse is based on user-defined cumulative fuel damage fraction logic, in which the fuel 

failure time is defined as a function of cladding temperature and only applied if the unoxidized 

Zircaloy cladding thickness is less than 0.0001 m. The failure logic calculates the fuel damage 

fraction for the current timestep, if the unoxidized cladding thickness criteria are met, and adds 

that fractional damage to any previously calculated damage. When the cumulative fuel damage 

fraction exceeds unity, the fuel is failed in the SFP MELCOR model. This lifetime damage 

model eliminates the threshold behavior present in the other fuel failure criteria and predicts 

accumulating damage if the fuel remains above the melting temperature of Zircaloy and below 

the absolute threshold collapse criteria of 2500 K.

All components other than fuel rods (fuel and cladding) will be immediately converted to 

particulate debris whenever the unoxidized metal thickness is reduced below a user-defined

minimum value. The minimum thickness criterion for the two MELCOR canister components is 

0.0001 m. The unoxidized metal thickness is reduced both by oxidation and by melting and 

candling of metal. Molten Zircaloy held up by an oxide shell is released from the fuel rods at 

2400 K and from the canister at 2100 K (i.e., just above the melting temperature of the Zircaloy). 

Particulate debris will be formed for canister components following the release of the molten 

Zircaloy or if the temperature of the component reaches the melting temperature of the 

associated oxide.

Baseplate failure is defined by the grid-supported or egg-crate plate model in MELCOR. In 

general, the beams that form the grid have sufficient strength that their failure is not an issue, 

and the interest is in failure of the web between them. Upon failure of the plate, the capability to 

support particulate debris or intact components is lost; however, the plate will remain in place 

until it melts. This model calculates baseplate failure based on the maximum stress in a plate of 
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user-defined thickness supported by beams of user-defined spacing with a total load on the 

area of the ring. In the SFP model, the thickness of the baseplate is defined as 0.0127 m with 

grid spacing of 0.07 m. The melting temperature of the plate is 1700 K.

6.1.5 Radionuclide Transport Modeling and Treatment of Hydrogen

In MELCOR, the RN package models the release and transport of fission product vapors and 

aerosols (referred to as radionuclides). Release of radionuclides can occur from the fuelcladding gap by exceeding a failure temperature criterion or losing intact geometry or from 

material in the SFP using various empirical release correlations based on fuel temperatures. 

After release to a control volume, masses may exist as aerosols or vapors, depending on the 

vapor pressure of the radionuclide class and the volume temperature.

Aerosol dynamic processes and the condensation and evaporation of fission product vapors 

after release from fuel are considered within each control volume. Aerosols can deposit directly 

on surfaces and water pools or can agglomerate and eventually fall out by gravitational settling. 

Aerosols deposited on surfaces can be vaporized (if volatile) but cannot currently be 

resuspended in MELCOR. All deposition mechanisms are mechanistically modeled. Aerosols 

and vapors are transported between control volumes by bulk fluid flow of the atmosphere and 

the pool.

For tracking purposes, the radionuclides are combined into material classes, which are groups 

of elements (and their isotopes) with similar chemical and transport behavior. Radionuclide 

masses include both the radioactive and nonradioactive mass to properly model the transport of 

fission products. The SFP MELCOR model includes 15 default material classes and two 

user-defined classes to model the behavior of cesium iodide (CsI) and cesium molybdate

(Cs2MoO4), as shown in Table 23.

The fuel release model is based on the CORSOR-Booth model that more accurately predicts 

the release rates from the Phebus and VERCORS experiments (Gauntt, 2010). The default 

MELCOR radionuclide package input was modified to accommodate new insights from the 

Phebus experimental program. The cesium, iodine, and molybdenum radionuclide classes 

were reconfigured as follows:

• Class 4—Characteristic released compound is iodine with the default inventory wholly 

transferred to Class 16.

• Class 7—Characteristic released compound is molybdenum with the default inventory 

reduced by the amount allocated to Class 17.

• Class 16—Characteristic released compound is CsI with the default inventory 

representing all of Class 4 and sufficient cesium from Class 2 to form CsI.

• Class 17—Characteristic released compound is Cs2MoO4 using the remainder of the 

cesium not in the gap (already included in Class 2) or not already combined with the 

iodine in Class 16. Sufficient molybdenum is included from Class 7 to Class 17 to form 

Cs2MoO4. The released vapor pressure and compound mass is consistent with 

Cs2MoO4.
102

Gauntt (2010) proposes an approach for the estimation of increased ruthenium release under 

air-oxidation conditions. Ruthenium (Class 6) has the lowest of vapor pressures in the default 

MELCOR model that prevents prediction of large releases.

13 There is evidence of higher 

volatility of ruthenium oxides (many orders of magnitude higher than the default MELCOR). It is 

assumed (Gauntt, 2010) that there is always air present leading to formation of a moderately 

hyperstoichiometric fuel (UO2.15) and release of ruthenium dioxide (RuO2). The default vapor 

pressure parameters in MELCOR are adjusted for the ruthenium class to match RuO2 vapor 

pressure at 2200 K.

14 The new ruthenium release model is applied only to scenarios involving 

rapid draindown (for moderate leak rates) of the SFP pool. These cases lead to relatively early 

clearing of the rack baseplate and flow of air (and possibly steam) through the assemblies. It 

should be noted that the model does not take into account the concentration of oxygen or steam 

during the oxidation process.

Table 23 MELCOR Radionuclide Class Composition

Class # Class Name Representative Member Elements

1 Noble Gases Xe He, Ne, Ar, Kr, Xe, Rn, H, N

2 Alkali Metals Cs Li, Na, K, Rb, Cs, Fr, Cu

3 Alkaline Metals Ba Be, Mg, Ca, Sr, Ba, Ra, Es, Fm

4 Halogens I F, Cl, Br, I, At

5 Chalcogens Te O, S, Se, Te, Po

6 Platinoids Ru Ru, Rh, Pd, Re, Os, Ir, Pt, Au, Ni

7 Early Transition 

Elements

Mo V, Cr, Fe, Co, Mn, Nb, Mo, Tc, Ta, W

8 Tetravalent Ce Ti, Zr, Hf, Ce, Th, Pa, Np, Pu, C

9 Trivalents La Al, Sc, Y, La, Ac, Pr, Nd, Pm, Sm, Eu, 

Gd, Tb, Dy, Ho, Er, Tm, Yb, Lu, Am, 

Cm, Bk, Cf

10 Uranium U U

11 More Volatile Main 

Group

Cd Cd, Hg, Zn, As, Sb, Pb, Tl, Bi

12 Less Volatile Main 

Group

Sn Ga, Ge, In, Sn, Ag

13 Boron B B, Si, P

14 Water H2O H2O

15 Concrete - -

16 Cesium Iodide CsI CsI

17 Cesium Molybdate Cs2MoO4 Cs2MoO4

The gap inventory specified in Table 24 is based on NUREG-1465, “Accident Source Terms for 

Light-Water Nuclear Power Plants,” issued February 1995 (NRC, 1995). However, NUREG1465 states that, for accidents in which long-term cooling is maintained (e.g., postulated spent 

fuel handling accident), the gap release could be as low as 3 percent, and, in the unmitigated 

scenarios in this study, the fuel experiences prolonged high temperatures (and even failure in 

some instances). Therefore, the present work assumes that 5 percent applies to all scenarios.

 13 There is a mass transfer limitation to the release from the fuel. 14 The rationale for an increased ruthenium class release is based only on increased vapor pressure and 

requires further experimental validation.
103

The decay heat calculation was based on security assessment analyses that used a decay heat 

program provided by the licensee. The decay heat power is calculated based on the discharge 

time and other parameters, such as the fuel burnup and power history. The utility provided the 

program and the appropriate input files for the SFP configuration after its last offload (i.e., 

September 2001) to perform decay heat calculations. Consequently, the decay heat power of 

every assembly was calculated as a function of time from reactor shutdown.

15 The decay heat 

and radionuclide package for MELCOR was conceived for reactor analysis. Therefore, all 

assemblies are assumed to have the same shutdown time. MELCOR calculates the initial 

fission product inventory from tables of inventories and specific decay power for 29 elemental 

groups. The elemental decay heat is normalized per unit of mass of the element and stored as 

a function of time after shutdown.

Table 24 Radionuclides Gap Inventories

Class # Gap inventory Class combination

1 5% —

2 100% Characteristic released compound is CsOH with the default 

inventory wholly representative of the cesium in the fuel gap 

except what is already included in Class 16. Required amount of 

cesium not in gap of Class 16 to yield a 5% total cesium gap 

inventory.

3 1% —

5 5% —

16 5% 5% of the Class 16 inventory to yield 5% of the total iodine 

inventory in the gap

Since SFP accident calculations involve fuel assemblies with multiple shutdown times, the 

following procedure was used to implement the batch-average decay heat results. First, the 

effective reactor operating power was estimated using SFP inventory burnup. The effective 

operating power was calculated as the total burnup of all assemblies in the SFP (gigawatt days 

per metric tons of uranium) divided by the average assembly metric tons of uranium and the 

total number of days of criticality. Based on the effective operating power, MELCOR calculates 

the specific time-dependent decay heat and mass inventory for each element. The aging time 

in the specific element decay heat tables is specified as the scenario time minus the shutdown 

time of the assemblies in the most recent offload. Next, the above results for element 

inventories (kilogram (kg)) times the specific element decay heat (watts per kilogram) at the 

scenario time are scaled to match the total SFP decay power. This scaling procedure 

addresses any limitations in the relatively long-term decay heat power in the MELCOR data 

base. Finally, inventory scaling coefficients are used to partition the decay heat amongst the 

various MELCOR rings. In summary, the batch-average decay heat is explicitly conserved but 

the fission product inventory is not properly scaled to account for differences in the various 

assembly discharge dates. A postprocessing routine is implemented that uses the MELCOR 

predicted release fractions along with actual inventories calculated for each batch.

 15 An interpolation scheme was used to calculate the individual assemblies decay power at different times 

relevant to this study (the error in interpolation is typically less than 1 percent). Since the number of old 

assemblies was increased by 60 (3,055 total in the pool), the decay heat for these assemblies was assumed 

to be an average of the older assemblies.
104

To accommodate consequence calculations using MACCS2, an extensive control system was 

written in the MELCOR input file that tracks the fission product releases from each ring16 and 

the subsequent release to the environment. Time-dependent, nondimensional environmental 

release fractions are calculated for each batch (i.e., MELCOR ring) that can be multiplied by the 

specific batch fission product activities to evaluate the environmental source term. The 

following procedure was used to map the releases from MELCOR to MACCS2. MELCOR 

activity release for each isotope (e.g., m = Cs-137, Cs-134, Cs-136 for Class 2) is given by the 

following:

 (9)

MACCS activity release is given by the following:

(10)

where RF� (t) is defined as:

(11)

(12)

Where

r = ring number (total 6 rings)

m = radionuclide {1:M} where M is the number of ORIGEN-S isotopes in each class

t = time since start of event

RF (t) = environmental release fraction (ring by ring from MELCOR)

RF� (t) = environmental release fraction (by radionuclide group for MACCS2) 

A = released activity (Becquerel (Bq))

Ao = initial inventory (Bq) from ORIGEN-S (69 isotopes for each MELCOR ring)

Radionuclide Inventories

The radiological inventories and decay heat for assemblies in the SFP were calculated using 

information provided by the utility for all assemblies discharged to the pool through Cycle 18 

(September 2011). The information included the assembly identification, design type, initial 

enrichment, discharge burnup, and discharge date. The analysis basis for the high-density SFP 

inventory was 3,055 assemblies, a number based on the pool capacity of 3,819 assemblies, 

reduced by 764 assemblies to accommodate a full core offload capability. Information on 

 16 A ring is a collection of assemblies in the MELCOR radial nodalization.
105

assemblies discharged before Cycle 7 is not considered since the target pool inventory was 

achieved with the assemblies from Cycles 7 to 18. 

Assembly depletion and decay calculations were performed using the ORIGEN code (Gauld et 

al., 2011), maintained within the SCALE nuclear safety analysis code system (Rearden et al., 

2012). The nuclear cross-section libraries used for the burnup analysis of the assemblies were 

those distributed in SCALE 6.1. These libraries are developed using ENDF/B-V cross-sections 

and include representative 7x7, 8x8, 9x9, and 10x10 General Electric assembly designs (Ilas et 

al., 2006) used in the reference plant reactor. ORIGEN calculations performed using these 

libraries have been validated against experimental destructive assay measurements, and 

calorimeter measurements of assembly decay heat have been demonstrated in previous 

validation studies (Ilas and Gauld, 2008) to be accurate within plus or minus 2 percent.

For the burnup analysis, the irradiation and decay history for each of the 3,055 assemblies in 

the pool was simulated using ORIGEN and assembly-specific design and operating history data 

provided by the utility. Each assembly was decayed to a reference date corresponding to the 

end of Cycle 18, and the assembly inventories combined into analysis groups. The groups were 

then further decayed to calculate spent fuel assembly activities and decay heat power for 

analysis cooling times of 3.6, 3.9, 5.0, 13.1, 37.0, 107.0 and 383.0 days after shutdown of the 

reactor. The assemblies were grouped according to the cycle they were discharged:

• Group 1 (268 assemblies from Cycle 18)

• Group 2 (272 assemblies from Cycle 17)

• Group 3 (272 assemblies from Cycle 16)

• Group 4 (276 assemblies from Cycle 15)

• Group 5 (284 assemblies from Cycle 14)

• Group 6 (1,683 assemblies from Cycles 7 to 13) 

This division of assemblies by group facilitated use of the data for an analysis of a low-density 

SFP configuration, whereby all assemblies with a cooling time greater than 5 years have been 

removed from the pool. For the present analysis, each offload was assumed to be 

284 assemblies for modeling convenience and to avoid modifying the MELCOR model 

nodalization.

17 Therefore, the actual inventories from batches were scaled appropriately to 

correspond to the rings in the MELCOR nodalization. For example, for the low-density case, the 

Cycle 18 inventories were increased by 284/268 and the sum of Cycles 16 and 17 were scaled 

as 568/(272 + 272), resulting in 852 assemblies as opposed to the actual 812. 

The SFP results were compiled for each assembly group and all decay times and included 

activities (Bq) for 69 radionuclides and decay heat.18

Results from the present analysis were compared with those generated previously for the

reference plant pool using assembly data provided by the utility through 2001 as part of the 

security assessment work. A limitation of the 2001 data was that the utility did not provide the 

actual discharged burnup distribution of assemblies from Cycles 12 and 13. Consequently, 

previous analyses assumed burnup distributions for these cycles based on data from Cycles 10 

and 11. Review of the actual burnup distributions included in the 2011 data indicates that the 

 17 The nodalization was based on the security assessment work. The additional data on later cycles were 

received after the MELCOR model had been developed and the calculations were started. 18 The decay heat in the present analysis is based on the past security assessment work.
106

average discharge burnup increased significantly after Cycle 12. The burnup values used in the 

present analysis are significantly higher and therefore more representative of modern SFP

inventories than earlier analyses. Previous analyses using the 2001 data are representative of 

discharged fuel up to about 1995.

Other differences are attributed to the specific power of the assemblies which influences the 

decay heat power and activities of short-lived fission products in the analysis time range. The 

utility did not provide information on the specific power. Notwithstanding power uprates for the 

reference plant reactor, the most recent occurring in 2002, the specific power used to calculate 

inventories for the assemblies in the present analysis was lower than that assumed using the 

earlier 2001 data. The present analysis normalized the average specific power of the 

discharged assemblies to the reactor specific power. Previous information provided by the 

utility in the 2001 data included the effective full-power days used to derive slightly higher 

specific power values compared to those used in the present study. 

The net impact of differences between the analyses performed using 2001 data and the present 

analysis is an increase in the inventories for cooling times longer than about 30 days, attributed 

to higher assembly burnup in the 2011 data. For shorter cooling times the previous analyses 

predicted decay heat rates about 5 percent larger than the current results, likely the result of

more conservative estimates of specific power used in the previous analyses. A comparison of 

the present decay heat results with values calculated by the utility in 2001 show agreement to 

better than 3 percent over all cooling times, with present results slightly larger than utility values, 

most likely because of the increase in discharge burnup since 2001.

Hydrogen Burn

A burn is initiated in a control volume if the mole fraction of the reactants (hydrogen and oxygen) 

satisfies the burn criteria. In addition, control volumes that are specified to contain igniters are 

tested against different criteria than control volumes without igniters. In an SFP calculation, 

ignition is assumed to occur in the reactor building when the hydrogen concentration exceeds 

10 percent by volume. In addition, MELCOR checks to determine whether there is sufficient 

oxygen. The minimum oxygen mole fraction for ignition is 5 percent. The maximum diluents 

mole fraction for ignition (mole fraction of steam plus mole fraction of carbon dioxide) is 

55 percent. If all of these conditions are satisfied, a burn is initiated. Some uncertainty may 

exist regarding the combustion of hydrogen, especially with regard to the timing of a 

spontaneous ignition. A hydrogen burn may occur at higher or lower concentrations of 

hydrogen, air, and steam that have both epistemic and aleatory uncertainties. Many SFP 

calculations resulted in conditions in which combustion was very likely or very unlikely. 

Consequently, the SFPS presents the results of cases with and without combustion. However, 

some cases have conditions in which the occurrence or timing of a combustion event has more 

uncertainty; these cases were assumed to ignite or not ignite according to the default 

spontaneous combustion criteria in MELCOR (see Section 9.1). Once a burn is initiated, it can 

propagate to other control volumes using the default hydrogen concentrations of 4 percent, 

6 percent, and 9 percent for upward, horizontal, and downward propagations, respectively.

6.2 Description of MELCOR Models

The SFP, 40 ft (12.2 m) wide by 35.3 ft (10.8 m) long by 38.75 ft (11.8 m) deep, is located on 

the refueling floor of the reactor building. The pool is constructed of reinforced concrete with a 

wall and floor lining of 1/4-in.- (0.63-cm-) thick stainless steel. The walls and the floor of the 
107

SFP are approximately 6 ft (1.83 m) thick. In the northeast corner of the SFP is a cask area that 

is 10 square feet (ft2

) (0.93 square meter (m2)).

The high-density SFP racks provide spent fuel storage at the bottom of the fuel pool. The fuel 

storage racks are normally covered with about 23 ft (7 m) of water for radiation shielding. The 

SFP racks are freestanding, full length, and top entry and are designed to maintain the spent 

fuel in a spaced geometry that precludes the possibility of criticality. The high-density SFP 

racks are of the “poison” type utilizing a neutron-absorbing material to maintain a subcritical fuel 

array. The racks are rectilinear in shape and are of nine different sizes. A total of 3,819 storage 

locations are provided in the pool. The racks are constructed of stainless steel materials, and 

each rack module is composed of cell assemblies, a baseplate, and base support assemblies. 

Each cell is composed of (1) a full-length enclosure constructed of 0.075-in.- (0.2-cm-) thick 

stainless steel, (2) sections of Bisco Boraflex, which is a neutron-absorbing material, and 

(3) wrapper plates constructed of 0.020-in.- (0.05-cm-) thick stainless steel. The inside square 

dimension of a cell enclosure is 6.07 in. (0.15 m). The cell pitch is 6.28 in. (0.16 m). The 

baseplate is made from 0.5-in.- (1.27-cm-) thick stainless steel with 3.8 in. (0.1 m) chamfered 

through-holes centered at each storage location, which provides a seating surface for the fuel 

assemblies. These holes also provide passage for coolant flow.

Each rack module has base support assemblies (i.e., “rack feet”) located at the center of the 

corner cells within the module and at interior locations to distribute the pool floor loading (see 

Figure 39). Each base assembly is composed of a level block assembly, a leveling screw, and 

a support pad. The top of the leveling block assembly is welded to the bottom of the base plate. 

SFP fuel cells are located above each rack foot. Four 1-in. holes are drilled into the side of the 

support pad. The interior of the support pad is hollow and permits flow to the opening in the 

base plate. The square tube cells are used to construct the rack cells, which results in an equal 

number of cells resulting from the square tube cell checkerboard layout. Figure 39 shows the 

layout of the rack cells. There is the potential for lateral cell-to-cell flow between connected rack 

cells.

Figure 39 Typical SFP rack cut away cross sections

Figure 40 shows the control volume nodalization of the SFP region of the whole pool model. 

The bottom of the pool was divided into eight regions. CV299 represents all open regions in the 

SFP around the racks, including the cask area. The racks are subdivided into the other seven
108

regions. Ring 7 (CV170 and CV171) represents the empty rack cells on the periphery of the 

SFP. All of the assemblies in the SFP are located in Rings 1 through 6. Each ring with 

assemblies is further subdivided into 19 control volumes—one control volume below the racks,

nine control volumes inside the canister, and nine control volumes in the bypass region between 

the rack and canister. For example, CV110, CV111 through CV119, and CV211 through CV219 

represent the region beneath the rack, the region within the canister, and the bypass region 

between the rack and canister, respectively (see Figure 41). Similarly, Rings 2 through 6 

contain similar canister and bypass region nodalizations. The region above the pool is divided 

into two control volumes. Typically, flow goes down CV301 and CV299 and rises through 

CV300. The flow enters the bottom of the racks through CV110 through CV170. For 

low-density configurations, the control volume nodalization does not contain a bypass region 

(between the channel box and rack) as shown on the right side of Figure 41.

Figure 40 MELCOR nodalization of the whole pool high density model
109

Figure 41 MELCOR nodalization of the assemblies (only two rings shown)

The hydraulic resistance was specified using the results from the SNL experimental test 

program (Durbin, 2005).19 For example, for the flowpath connecting CV113 and CV114 in the 

fully populated region, the MELCOR input values included a form loss coefficient of 3.8, and a 

friction factor (SLAM) of 31.3 (equal to 125/4 since MELCOR uses the fanning friction factor 

definition). The flow resistance under the racks was represented using typical contraction 

inertial loss coefficients and viscous losses consistent with a flow length to the center of the 

SFP. The BWR assembly canister is modeled with the MELCOR canister component. The rack 

walls are modeled with the new rack component with stainless steel and Boraflex materials. 

MELCOR does not include an option to model the two large water rods in the center of the 

assembly. Consequently, the water rod mass and surface area was included in the canister 

wall.

The axial channel and bypass wall blockage models were active and controlled the resistance in 

the respective flowpaths. The blockage model monitors the porosity of the materials in the 

channel and bypass regions. If a debris bed forms, the flow resistance is adjusted via an Ergun 

flow resistance model. The canister wall radial blockage model controls flowpaths between the 

bypass region and the assembly. Initially, the canister wall precludes flow. However, if the 

canister fails, a radial flowpath is activated that permits flow between the two regions. Similar to 

the axial blockage model, the flow resistance is adjusted based on the local debris porosity.

 19 In the present study, the assembly nodalization is based on the GE14C 10x10 configuration (NRC, 2012) to 

account for the latest offloads used in the low-density configuration. Both 9x9 and 10x10 configurations 

have partial fuel rods. The flow area for each assembly is reduced by about 4 percent compared to the 9x9 

design. The hydraulic resistance data are assumed to apply. The frictional loss coefficient for a 10x10 array 

could be somewhat different since it is a function of hydraulic diameter and grid spaces design.
110

A complete reactor building has been developed for the reference plant (NRC, 2012d). 

However, the bulk of the reactor building does not play a significant role in SFP accidents, given 

that the study does not explicitly model (1) the effect of the SFP accident on reactor systems or 

(2) specific obstacles to deploying mitigation (e.g., presence of steam on lower elevations). 

Consequently, the reactor building model was simplified to only model the refueling room (i.e., 

within the red dashed line in Figure 42). 

A single control volume models the refueling bay. An open hatch in the southeast quadrant 

connects (via a flowpath) the refueling room to a boundary condition volume representing the 

flow connection to the lower sections of the building. The nominal reactor building leakage is 

modeled at the center elevation of the refueling bay, and the leakage flow from elevations in the 

simplified model from the lower regions was tuned to match the leakage flow rate of a detailed 

reactor building model. 

The detailed reactor building model simulated many overpressure failure flowpaths within the 

reactor building. The simplified refueling floor model included the two most important 

flowpaths—(1) the blowout panels on the refueling room walls and (2) a pathway representing 

the structural failure of the reactor building roof. The refueling room blowout panels will fail if 

there is an overpressure greater than 1,720 pascal (Pa) (0.25 pounds per square inch gauge 

(psig)). If the reactor building pressure rises above 3,450 Pa (0.5 psig), failure of the roof

decking will occur.

MELCOR does not include models for stratification of hot gases. Each control volume is 

assumed to be well mixed and have a single temperature. Large-scale natural circulation flow 

patterns can be predicted when the bulk temperature differences between adjacent rooms 

create mixing flows. However, it would be awkward or perhaps impossible to predict complex 

plume behavior within regions typically modeled with a single control volume (e.g., the room 

above the SFP). Consequently, the MELCOR calculations are expected to overpredict the 

amount of thermal mixing within the building. Based on insights from the computational fluid 

dynamics calculations for the security assessment work, the MELCOR refueling room model 

nodalization included modeling features to minimize excessive mixing. The refueling room is 

modeled as a single control volume. However, the inlet flow into the SFP (i.e., CV301 in 

Figure 40) comes directly from the hatch region (see left side of Figure 42). In this manner, the 

cool gases leaving the lower regions of the building are not brought into thermal equilibrium with 

gases above the SFP. Cross-flow is simulated between CV300 and CV301 as observed in the 

computational fluid dynamics calculations.
111

Figure 42 MELCOR reactor building model

6.2.1 High-Density Loading during Outage

During an outage in which the SFP and reactor are hydraulically connected, a single control 

volume is used to represent both the reactor well and separator/dryer pool, as shown in Figure 

43. The total volume of pool in CV601 is about 1,900 m3 (neglecting the dead-end pool volume 

of 243 m3 below the separator/dryer gate elevation). CV601 is hydraulically connected to 

CV300 (see Figure 40) using two flowpaths until the water level reaches the SFP gate and no 

more water can flow into the SFP. The reactor power is applied as an external energy source 

until the pools become disconnected. The total additional volume of water above the SFP gate 

is about 1,400 m3

.
112

Figure 43 SFP and reactor connection model during outage

For both OCP1 (at 4 days) and OCP2 (at 13 days), CV601 is connected to the detailed model of 

the SFP (Figure 40). Figure 44 shows the assembly layout for OCP1 in a 1x4 pattern in which

the assemblies are grouped into six types or “rings” by decay heat power and time of discharge. 

The 88 assemblies from the most recent offload in Ring 1 are surrounded by 352 old 

assemblies in Ring 2.

20 Ring 3 is empty during the outage where the assemblies still reside in 

the reactor.

21 Ring 5 contains the last offload (284 assemblies) with an additional 

31 assemblies from previous offloads. Rings 2, 4, and 6 have a total of 2,456 assemblies with 

their total decay heat distributed in each ring scaled by the number of assemblies. Within each 

MELCOR ring, the assembly decay heat is uniform. Consequently, for any given scenario, the 

decay heat in each ring is adjusted to give the average assembly power. Finally, the 764 empty 

cells in Ring 7 were placed around the outside of the SFP, which promotes open air downflow 

into the SFP in the event of a complete loss-of-coolant inventory accident. The empty cells (764 

in Ring 7 and 196 in Ring 3) have no decay heat. For the empty cells in Ring 3, the axial 

nodalization is detailed (see Figure 41) without the bypass control volume. This will ensure a 

better representation of flow through the assemblies and modeling of heat transfer between 

components in various rings.

 20 All of the old assemblies are smeared in MELCOR Rings 2, 4, and 6 (i.e., decay power per assembly is the 

same). 21 The decay power for the Ring 3 assemblies is added to the CV601 external power. Therefore, OCP1 has 

less power in the SFP since the 196 Ring 3 assemblies have not been moved yet.
113

Figure 44 Layout of assemblies for OCP1 high density (1x4) model

Figure 45 shows the cell-wall radiation view factors between the various rings.22 The resultant 

view factor specifies the amount of coupling from each region to another. For example, the 

Ring 1 cells are completely surrounded by Ring 2 cells. Hence, the view factor from Ring 1 to 

Ring 2 is 1.0. Similarly, Rings 3 and 4 and Rings 5 and 6 are coupled in 1x4 patterns. Using 

the specific layout in Figure 44, the special MELCOR generalized radiative heat transfer 

coupling model was prescribed to represent the thermal coupling between Rings 2 and 4, 

Rings 4 and 6, Rings 6 and 7, and Ring 7 and the SFP wall. The radial coupling for these 

regions was specified as the product of the area (i.e., represented as the number of coupling 

panels) times the view factor.

23 In OCP2, the 196 assemblies have been moved to Ring 3, as 

shown in Figure 46, and the radial thermal coupling is preserved as in Figure 45.

 22 MELCOR models intracell radiation between concentric rings by default. To disable the radiation model for 

Rings 2 to 3 and 4 to 5, the radial view factor area is set to zero. 23 The view factor is assumed to be unity. It should be noted that there is a temperature gradient within each 

ring, and MELCOR attempts to model a multidimensional geometry with a simplified two-surface radiation 

model.
114

Figure 45 MELCOR radial radiative coupling scheme

Figure 46 Layout of assemblies for OCP2 high-density (1x4) model

The methodology described in Section 6.1.5 was used to estimate the decay heat power as a 

function of time for different OCPs. Table 25 shows the results of this analysis. The reactor 

power was based on the decay power for all assemblies residing in the reference plant reactor 

(NRC, 2012d) by subtracting the power associated with assemblies that have already been 
115

moved to the SFP. For example, for OCP1, the analysis assumed that 88 assemblies are 

already in the SFP.

Table 25 Distribution of Decay Heat in the Reactor and SFP for High Density Loading 

Reactor

(kW)

Spent Fuel Pool (kW)

Days Ring 1

(88)1

Ring 3

(0)

Ring 5

(315)

Ring 2

(352)

Ring 4

(784)

Ring 6

(1,320)

Total

(2,859)

OCP1

10,216 3.6 1,927 0 465 80 179 301 2,951

9,915 3.9 1,867 0 452 80 179 301 2,878

9,006 5.0 1,690 0 417 80 178 300 2,666

7,406 8.0 1,403 0 358 80 178 300 2,320

6,710 10.0 1,282 0 334 80 178 300 2,174

Days Ring 1

(88)

Ring 3

(196)

Ring 5

(315)

Ring 2

(352)

Ring 4

(784)

Ring 6

(1,320)

Total 

(3,055)

OCP2

4,395 13.1 1,144 1,533 332 80 178 300 3,567

4,117 15.0 1,077 1,444 330 80 178 299 3,409

3,530 20.0 957 1,294 318 79 176 296 3,120

Days Ring 1

(88)

Ring 3

(196)

Ring 5

(315)

Ring 2

(352)

Ring 4

(784)

Ring 6

(1,320)

Total 

(3,055)

OCP3 37 720 973 324 79 177 298 2,571

OCP4 107 422 602 301 78 173 292 1,868

OCP5 383 191 315 230 73 162 273 1,245

1. The numbers in parentheses are the number of assemblies.

6.2.2 High-Density Loading Postoutage

The layout for the postoutage high-density loading is similar to OCP2 (see Figure 46). In 

postoutage, the assemblies are assumed to be in a 1x4 pattern, which applies to OCP3, OCP4, 

and OCP5. The assembly layout remained constant for these OCPs. However, the decay heat 

decreased from OCP3 to OCP5 as the aging time since reactor shutdown increased. Table 25

summarizes the decay heat power in each ring.

6.2.3 Low-Density Loading during Outage

For the low-density loading configuration, only the latest and the previous two offloads are 

considered. Therefore, for OCP2, the total number of assemblies in the pool is 852 (equal to 

284 × 3). For OCP1, the 196 assemblies from the current offload are still in the reactor and only 

88 have been moved, resulting in only 656 assemblies in the pool. Figure 47 shows the layout 

of assemblies in the SFP for OCP1, and Figure 48 shows the layout for OCP2. For both 

configurations, all of the old fuel has been removed from the pool, and the current offload is in a 

1x4 pattern with empties. Because of space limitations, the last two offloads are placed in a 

checkerboard pattern.24 For the axial nodalization, Ring 1 contains both the channel (inside the 

 24 There is not enough room to place all of the fuel in a 1x4 pattern. The current offload eventually requires 

1,420 cells (284 for assemblies and 284 × 4 for empties surrounding them), which would leave only 

1,635 cells (excluding Ring 7). The 568 assemblies would require 2,840 cells for storage in a 1x4 pattern.
116

canister) and the bypass (outside between canister and rack) control volumes, while both 

volumes are combined for Ring 2 (see Figure 41). The basic radial thermal coupling from 

Figure 45 still applies, but the boundary area from Ring 6 to Ring 7 is 472 panels. For modeling 

convenience, Rings 2, 4, and 6 from the high-density layout are still present, but the cells 

contain only the rack component.

Table 26 provides the distribution of decay heat in the pool. A comparison with the high-density 

decay heat shows that the total decay heat in the pool for the low-density case is reduced by 

less than 20 percent. The total pool decay heat is dominated by the last offload, which is the 

same for the low- and high-density configurations. However, removing the old fuel also 

increases the available water volume (not occupied by the fuel and canister), while at the same 

time modifying the propagation characteristic of zirconium fire because of reduced mass in the 

empty assemblies.

Figure 47 Layout of assemblies for OCP1 low-density model
117

Figure 48 Layout of assemblies for OCP2 low-density model

6.2.4 Low-Density Loading Postoutage

The postoutage low-density layout for OCP3, OCP4, and OCP5 is identical to OCP2 (see Figure 

48), and the pool decay heat is provided in Table 26.
118

Table 26 Distribution of Decay Heat in the Reactor and SFP for Low Density Loading 

Reactor

(kW)

Spent Fuel Pool (kW)

Days Ring 1

(88)

Ring 3

(0)

Ring 5

(568)

Ring 2

(0)

Ring 4

(0)

Ring 6

(0)

Total

(656)

OCP1

10,216 3.6 1,927 0 599 0 0 0 2,526

9,915 3.9 1,867 0 587 0 0 0 2,454

9,006 5.0 1,690 0 551 0 0 0 2,241

7,406 8.0 1,403 0 492 0 0 0 1,895

6,710 10.0 1,282 0 468 0 0 0 1,750

Days Ring 1

(88)

Ring 3

(196)

Ring 5

(568)

Ring 2

(0)

Ring 4

(0)

Ring 6

(0)

Total 

(852)

OCP2

4,395 13.1 1,144 1,533 466 0 0 0 3,143

4,117 15.0 1,077 1,444 464 0 0 0 2,985

3,530 20.0 957 1,294 448 0 0 0 2,699

Days Ring 1

(88)

Ring 3

(196)

Ring 5

(568)

Ring 2

(0)

Ring 4

(0)

Ring 6

(0)

Total 

(852)

OCP3 37 720 973 455 0 0 0 2,149

OCP4 107 422 602 427 0 0 0 1,451

OCP5 383 191 315 339 0 0 0 845

6.3 MELCOR Analysis Results

6.3.1 Sequences That Do Not Lead to a Release

In general, the following four classes of scenarios do not result in a release from the fuel:

1. boiloff scenarios with no SFP leaks

2. mitigated scenarios for small leaks

3. unmitigated scenarios in late phases (OCP4, OCP5)

4. mitigated moderate leak scenarios in OCP2, OCP3, OCP4, and OCP5

Boiloff

For the boiloff scenarios, a simplified model was used to estimate the pool heatup and water 

level drop. Figure 43 shows this model in which all of the assemblies are combined in two rings 

representing the fuel and empty cells. Only the thermal-hydraulic models in MELCOR are 

active, and the power for both the reactor well pool and SFP are provided as external sources to 

the water pool. The results are considered conservative since the heat capacities of the 

assemblies are not taken into account. The time-dependent power is taken from Table 25 for 

high-density cases or Table 26 for low-density cases. The top of the pool is connected to the 

reactor building (see Figure 42) in the same manner as in the detailed model. This simplified 

model is used as a screening tool to determine whether more detailed analysis is needed. 

Figure 49 shows the water level as a function of time for both high- and low-density cases for 
119

OCP1, OCP2, OCP3, and OCP4.25 Figure 49 also identifies the time required to reach pool 

saturation. For cases in the same OCP, the high-density cases become saturated sooner since 

there is less water volume and more decay heat. In the late OCPs following refueling, the 

difference in the timing directly correlates to the decay heat power. While there are differences 

in postsaturation water level for OCP3 and OCP4, the water level for OCP1 and OCP2 is similar 

as a result of mixing assumed between the reactor well water and the SFP water (see Figure 

43). For the OCP4 low-density and OCP5 cases, the SFP never becomes saturated in 72 

hours. The slight water level increase during the sensible heating period results from the

change in pool density as the water heats up. The analysis shows that there is 4.6 m (15 ft) of 

water above the top of racks in OCP1 at 72 hours.

Figure 49 Water level for boiloff scenarios

Mitigated Scenarios (Small Leaks)

The small leak is modeled in MELCOR with a 4.4 cm (1.75-in.) diameter hole at the bottom of 

the pool based on the structural analysis and damage to the pool (effective size of cracks in the 

liner and the concrete). Figure 50 and Figure 51, respectively, show the water level and the 

injection and leak mass flow rates for the low-density OCP1 case. Once the water level reaches 

10 m at about 7 hours, the leak is detected and, together with the deployment logic, the water 

injection begins at about 9.5 hours. In this case, mitigation is direct makeup to the pool 

 25 The initial water level is assumed to be 11 m. The initial water temperature is 82 degrees F (28 degrees C). 

Both these initial conditions are applied to all accident scenarios in this report. Based on a teleconference 

with the licensee held on April 24, 2012, this is the postoutage water temperature under steady-state 

conditions where the heat exchangers are working (prior to postulated accident). During an outage (OCP1 

and OCP2), the water temperature could vary between approximately 80 degrees F and 100 degrees F. 

The higher temperature affects the sensible heating of the pool and is not expected to change the overall 

conclusion of boiloff scenarios given the significant margin observed.
120

(injection) since the water level at the time of deployment is more than 1 m above the top of the 

racks. For this small leak, the initial water flow rate is about 250 gpm (0.016 cubic meters per 

second (m3

/s)), which is much lower than the makeup capacity, and the water level is quickly 

restored. This calculation is only run for 24 hours to show the effectiveness of mitigation. 

Therefore, it is concluded that for all slow leak scenarios, the fuel never becomes uncovered 

since the makeup capacity is twice the leak rate. The leak rate is only a function of the water 

level (hydrostatic head) and is independent of the SFP configuration as long as the water level 

remains above the top of the racks.

Figure 50 Water level for mitigated low-density OCP1 (small leak) scenario
121

Figure 51 Flow rates for mitigated low-density OCP1 (small leak) scenario

Unsuccessful Deployment of Mitigation for OCP4 and OCP5 Scenarios

For OCP4, the decay heat is between 37 percent to 48 percent lower than for OCP3. None of 

the scenarios in OCP4 or OCP5 leads to a release from the fuel.

26 Figure 52 through Figure 55

illustrate the thermal-hydraulic response of the high-density pool to a small leak and a moderate 

leak. It takes less than 6 hours to clear the rack baseplate and initiate airflow for the moderate 

leak, while for the small leak case, the rack baseplate does not clear until about 39 hours. In 

both cases, there is a heatup of the fuel as the water level is reduced below approximately half 

the height of the fuel. For the small leak case, it takes longer and the heatup is slower since

there is some steam cooling of the fuel. 

The heatup rates for the low-density cases are somewhat similar to the high-density cases (see 

Figure 56 or Figure 57). The maximum clad temperature and the initial heatup rate in Ring 1 is 

actually higher for the low-density cases because of reduced heat transfer from Ring 1 to 

Ring 2.

27 Even though the total decay heat in the pool for low-density case is only 77 percent of 

the high-density case, the decay heat in Ring 1 is identical in both cases.

 26 The start of the release of radionuclides from the fuel is modeled based on a temperature of 900 degrees C 

(1,173 K). At this temperature, the cladding is assumed to fail and the gap inventory from the fuel is 

released. Further release from the fuel is based on the CORSOR-Booth model and is a function of fuel 

temperature (Gauntt, 2010). 27 The reduced mass in Ring 2 (only racks) initially limits heat transfer from Ring 1 until a sustained natural 

circulation is established.
122

Figure 52 Water level for unmitigated high-density moderate leak (OCP4)

Figure 53 Ring 1 temperature for unmitigated high-density moderate leak (OCP4)
123

Figure 54 Water level for unmitigated high-density small leak (OCP4)

Figure 55 Ring 1 temperature for unmitigated high-density small leak (OCP4)
124

Figure 56 Ring 1 temperature for unmitigated low-density moderate leak (OCP4)

Figure 57 Ring 1 temperature for unmitigated low-density small leak (OCP4)
125

Mitigated Moderate Leak Scenarios in OCP2, OCP3, OCP4, and OCP5

Mitigation for moderate leak cases involves actuation of the sprays for the postoutage scenarios

(OCP3, OCP4, and OCP5) and direct injection in OCP1 and OCP2. The moderate leak is 

modeled in MELCOR with a 11.4-cm- (4.5-in.-) diameter hole at the bottom of the pool based on 

the structural analysis and damage to the pool (effective size of cracks in the liner and the 

concrete). Section 6.1.3 of this report discussed the MELCOR modeling of the sprays and 

presented two modeling options (i.e., simple flow regime model on or off). Only high-density 

OCP3 results28 are presented since the unmitigated scenarios in later phases do not lead to 

release, and the moderate leak size is large enough to avoid the baseplate blockage resulting 

from quasi-steady water level at the bottom of the pool in response to the 200-gpm (0.013-m3

/s) 

spray water. The results of the OCP2 calculation showed no release from the fuel resulting 

from various heat transfer mechanisms (see also discussion for OCP1 in Section 6.3.2 of this 

report).

Figure 58 shows the water level for the moderate leak, high-density OCP3 scenario. Because 

of the spray activation at 3 hours (see Figure 59), the bottom of the racks clears for natural 

circulation airflow more than 1 hour later compared to an unmitigated case (see Figure 52). 

Finally, the spray flow rate and the leak rate are equilibrated by about 8 hours as required by the 

hydrostatic head at the bottom of the pool. The actual spray water reaching the bottom of the 

pool is somewhat less than 200 gpm (0.013 m3

/s) in Figure 59 because of heat transfer from 

spray droplets to the atmosphere and fuel rods.29 Figure 60 shows the response of the clad in 

Ring 1 for the case in which the simple flow regime model is active. As expected, the top cells 

experience more cooling as there is more water coverage. The temperatures reach a quasisteady state by about 10 hours30 and the maximum clad temperature is about 850 K. Figure 61

shows the clad temperatures for the case in which the simple flow regime model is disabled. In 

this mode, the main cooling mechanism is by convection from the fuel rods to the atmosphere, 

and none of the axial segments experience quenching. The maximum clad temperature is 

about 840 K, which is comparable to the previous case. Thus, even though the details of heat 

transfer and fuel heatup differ, the maximum clad temperatures are almost the same and well 

below the gap release criterion. This is partially because of the importance of the heat removal 

by natural circulation of air through the racks. If there was no natural circulation of air through 

the racks, the cooling of the fuel by the spray flow (i.e., modeled with the simple flow regime 

map) would be the only effective cooling mechanism, and therefore would be very important to 

the coolability of the fuel. 

To further test the impact of the modeling assumptions, two additional calculations were 

performed by assuming an additional 3-hour delay in the actuation of the spray as shown in 

Figure 62.

31 Both Figure 63 and Figure 64 show that (for OCP3), following the initial heatup of 

the fuel and reaching a maximum clad temperature (just below 900 K) at about 6 hours, the 

spray flow rate is sufficient to cool the fuel and avoid release.

 28 The low-density case is similar to the high-density case, and there is no release.

29 It would take about 15 gpm of water to remove the entire decay heat in the pool. However, some of the 

decay heat is being removed by natural circulation through the assemblies and leaking out of the reactor 

building. 30 The calculation fails shortly after 10 hours from numerical problems.

31 These cases were actually run based on an earlier logic for spray actuation that assumed a 3-hour 

additional delay at the end of deployment.
126

Figure 58 Water level for mitigated high-density moderate leak (OCP3)

Figure 59 Water flow rates for mitigated high-density moderate leak (OCP3)
127

Figure 60 Ring 1 clad temperatures for mitigated (simple flow regime active) highdensity moderate leak (OCP3)

Figure 61 Ring 1 clad temperatures for mitigated (simple flow regime inactive) highdensity moderate leak (OCP3)
128

Figure 62 Flow rates for mitigated high-density moderate leak (OCP3) with late actuation 

of sprays

Figure 63 Ring 1 clad temperatures for mitigated (simple flow regime active) highdensity moderate leak (OCP3) with late actuation of sprays
129

Figure 64 Ring 1 clad temperatures for mitigated (simple flow regime inactive) highdensity moderate leak (OCP3) with late actuation of sprays

6.3.2 Sequences That Do Lead to a Release

All the sequences in OCP1, OCP2, and OCP3 lead to release without successful deployment of 

mitigation. This section will only discuss representative scenarios to illustrate the accident 

progression phenomenology. One of the phenomena that have a significant impact on the 

overall release is the failure of the reactor building as a result of failure of the blowout panels or 

the roof. Failure of the reactor building introduces additional air that results in further oxidation 

of the hot fuel leading to enhanced release and fuel failure. The refueling room with the SFP at 

the top of the reactor building is modeled as a single volume (Figure 42), and hydrogen 

released from the SFP is assumed to mix with the entire volume. It is assumed that the 

hydrogen will combust at a 10-percent concentration if there is adequate oxygen (oxygen

concentration is greater than or equal to 5 percent) and no steam inerting (steam concentration 

is less than or equal to 55 percent). The analysis considered the sensitivity of the ignition 

assumptions and potential for reactor building refueling bay failure on a case-by-case basis (see 

Section 9.1 of this report).

Unsuccessful Deployment of Mitigation for Moderate Leak (OCP1) Scenario

The water level for the high-density scenario (Figure 65) shows that it takes about 8.5 hours to 

clear the rack baseplate and establish natural circulation in the pool. The timing is longer 

compared to postoutage scenarios (see Figure 52) because of the additional water in the 

reactor well connected to the SFP. The reactor power (Figure 66) is assumed to go to zero as 

the water level reaches the SFP gate and the pool is disconnected from the reactor well.

As the water level decreases, the clad temperatures (Figure 67) start to increase initially as a 

result of decay heat and then by clad oxidation as air is circulated through the assemblies. The 
130

heatup of the cladding in Ring 1 results in a zirconium fire that starts near the top of the full rod 

region (see Figure 41) and propagates downward. The heatup in Ring 1 (fuel, cladding, 

canister, and racks) propagates to Ring 2 assemblies (Figure 68) and leads to the failure of the 

racks in Rings 1 and 2. The failure of racks at about 12 hours results in formation of a debris 

bed in the bypass and relocation to the baseplate, but the channel boxes are still intact at this 

time. Between 13.6 and 14.2 hours, the channel boxes in Rings 1 and 2 fail, which allows 

additional cooling of the debris through flow diversion from the bypass region. As a result, the 

oxidation power is reduced and the heat transfer from the hot inner assemblies is propagated 

outward and starts to gradually heat up the SFP wall liner.32 Natural circulation and radial heat 

transfer throughout the SFP keeps the temperatures relatively low following the initial heat up in 

Rings 1 and 2. However, the fuel continues to slowly heat until a second zirconium fire initiates 

at the top of the fuel in Ring 4 at about 42 hours in the upper levels which then propagates 

downward. The second heatup is more intense and involves the other rings as indicated by 

both the oxidation power (Figure 66) and the clad temperatures in the outer rings (Figure 68).

OCP1 had a relatively rapid draindown in which an air natural circulation flow developed through 

the racks before significant oxidation of the fuel. As a result of a relatively short duration of the 

steam oxidation phase, there was relatively little hydrogen generation.33 The peak 

concentration in the refueling floor was only 5 percent, which is well below the minimum 

threshold for combustion and below a quantity that would lead to a significant pressurization of 

the reactor building. Consequently, there was no potential for a burn inside the refueling bay, 

which remained intact.

The fission product releases began at about 12 hours. Because the reactor building remained 

intact, all releases to the environment are limited by the nominal leakage (see Figure 42). The 

reactor building DF is shown in Figure 71.

34 Aerosols also begin to deposit inside the building 

and the DF for cesium and iodine aerosols remains between 3 to 4 for much of the accident. 

The DF is defined as the ratio of fractional release from the fuel to the fractional release to the 

environment. As discussed before, MELCOR keeps track of the fuel releases from individual 

rings. The fuel releases are divided by the overall DF to arrive at the environmental release for 

each ring. MELCOR mechanistically models all deposition mechanisms; however, because of 

the mixing within the reactor building, only an overall DF can be defined for all rings.

Figure 72 depicts the cesium environmental release fraction for individual rings. The release 

starts at about 9 hours from Ring 1 followed by the release from Ring 2 at 12 hours. The 

release profiles are consistent with the heatup in Figure 68. The later releases result from the 

second heatup and involve all of the outer rings (Ring 3 is empty for OCP1). The total release 

fraction is the input to the MACCS2 code for consequence analysis and is defined by Equation 

 32 The initial heatup of the liner is caused by heat transfer from the water. There is an initial cooldown as 

cooler air circulates before the heatup from the fuel caused the temperature to increase. 33 Hydrogen generation only occurs by oxidation of the SFP Zircaloy and steel with steam. Hydrogen is 

disassociated from the steam and released into the building, which can lead to combustion. If oxygen is 

present, then only air oxidation occurs and there is no hydrogen generation. In a larger leak, the water level 

drops below the bottom of the racks and allows natural circulation of air, which will preclude steam oxidation. 34 The integral DF is the ratio of the fission products released from the fuel to the amount that reaches the 

environment. Upon the start of fission product releases, the quantity is infinite until the release to the 

environment begins. Consequently, the initial peak is an artifact of the definition, whereas the long-term 

value is best characteristic of the reactor building performance.
131

11.

35 The DF is a dynamic quantity as the outer rings start to release (see the fluctuations in 

Figure 71); therefore, care is taken to allow the earlier releases from inner rings preserve their 

release history so that the total release fraction does not decrease at any time as the release 

progresses.

Figure 73 illustrates the results of the low-density case. Comparing the heatup with the 

high-density case (see Figure 67), the Ring 1 low-density case clearly heats up more rapidly 

initially since there are a lot of empty cells surrounding it (with the exception of the rack 

component) and heat is not very efficiently transferred radially, which results in slower heatup of 

Ring 5, as shown in Figure 74. Even though the racks fail in this low-density case, the canisters 

remain intact and the zirconium fire moves down initially and then upwards, as shown in Figure 

73. The cesium environmental release fraction for Ring 1 shown in Figure 75 is comparable to 

the high-density case (Figure 72), but since no release occurs from the older assemblies, the 

total release fraction for the low-density case is lower. 

Figure 65 Water level for unmitigated high-density moderate leak (OCP1)

 35 This activity-weighted release is a function of the inventories in each ring. Therefore, there is more 

contribution from the outer rings that have higher inventories even though the release from these rings is 

smaller compared to Ring 1.
132

Figure 66 SFP power for unmitigated high-density moderate leak (OCP1)

Figure 67 Ring 1 clad temperature for unmitigated high-density moderate leak (OCP1)
133

Figure 68 Midplane clad temperature for unmitigated high-density moderate leak (OCP1)

Figure 69 Midplane rack temperature for unmitigated high-density moderate leak (OCP1)
134

Figure 70 SFP wall liner temperature for unmitigated high-density moderate leak (OCP1)

Figure 71 Reactor building DF for unmitigated high-density moderate leak (OCP1)
135

Figure 72 Cesium environmental release fraction for unmitigated high-density moderate 

leak (OCP1)

Figure 73 Ring 1 clad temperature for unmitigated low-density moderate leak (OCP1)
136

Figure 74 Midplane clad temperature for unmitigated low-density moderate leak (OCP1)

Figure 75 Cesium environmental release fraction for unmitigated low-density moderate 

leak (OCP1)
137

Mitigated Moderate Leak (OCP1) Scenario

Figure 76 illustrates the response of the pool to the mitigated scenario. The connectivity 

between the reactor and the SFP, as well as the additional volume of water, results in a 

relatively slow draindown. Thus, at the end of the mitigation deployment, the water level in the 

pool is more than 0.9 m above the top of the rack.36 Therefore, instead of spray, mitigation is by 

direct injection into the pool. After about 12 hours, the water level remains relatively constant 

and the leak rate is balanced by the injection into the pool. The lower portions of the fuel remain 

cool and covered with water. Although heatup of the fuel occurs (see Figure 77), there is no 

indication of a zirconium fire and propagation through the pool. The peak fuel temperature 

reached 1200 K at 16 hours and remained near that value through 72 hours. A combination of 

radial heat transfer within the assembly; radial heat transfer from the recently discharged, 

high-temperature fuel to adjacent fuel assemblies; and steam cooling from boiling in the bottom 

of the assemblies between cells keep the fuel temperature near 1200 K. Only Ring 1 had 

cladding failure and subsequent releases of the gap inventory, as shown in Figure 78. All other 

fuel was below the threshold for cladding failure and fission product releases.

Figure 79 shows the clad temperature in Ring 1 for the low-density case. The heatup rate for 

the low-density case is more extreme than the high-density case, as was observed for the 

mitigated cases. Unlike the high-density case, the low-density case did not have low decay 

heat fuel assemblies adjacent to the recently discharged assemblies. Since an air natural 

circulation pattern through the racks was not established, the empty cells isolated the high 

decay heat assemblies and contributed to the higher heatup. The fuel in Ring 1 went through 

an oxidation transient, which led to peak fuel temperatures of 1800 K. However, once the 

steam in the assembly was consumed, the fuel temperatures dropped to 1200 K. The 

subsequent behavior was driven mainly by the decay heat, which was very similar to the 

high-density case. Higher fuel temperatures during the initial oxidation transient led to slightly 

more release in the low-density case.

 36 The level was close to 0.9 m above the top of the fuel of the fuel at the timing of the deployment of the 

sprays (i.e., 9.5 hours). If the spray system was used, cooling would be provided to the uncovered portion of 

the fuel. The accident could have benefitted from natural circulation of air through the racks once the water 

level dropped below the rack baseplate and spray cooling from the top.
138

Figure 76 Water level for mitigated high-density moderate leak (OCP1)

Figure 77 Ring 1 clad temperature for mitigated high density moderate leak (OCP1)
139

Figure 78 Cesium environmental release fraction for mitigated high density moderate 

leak (OCP1)

Figure 79 Ring 1 clad temperature for mitigated low density moderate leak (OCP1)
140

Figure 80 Cesium environmental release fraction for mitigated low density moderate leak 

(OCP1)

Unsuccessful Deployment of Mitigation for Small Leak (OCP2) Scenario

This scenario involves a hydrogen combustion that caused a late rapid air oxidation. Figure 81

shows the slow draindown of the pool exposing the top of the racks at 42.6 hours compared to 

18.7 hours for postoutage scenarios (see Figure 54). Figure 82 illustrates the decay power and 

the oxidation power. The air oxidation power reaches an order of magnitude higher than the 

decay heat during the oxidation transient after 60 hours. The fuel heatup begins after the water 

level reaches about the fuel midplane (see Ring 1 response in Figure 83). The 

high-temperature fuel in Ring 1 heats the surrounding low decay heat fuel in Ring 2, as shown 

in Figure 84.

The evolution of reactor building steam and air shows that, by the time the water level reached 

the SFP gate and the SFP is disconnected from the reactor, the building is filled with steam 

which continues to decrease as it is condensed on structures. The hydrogen concentration 

builds up until it reaches 10 percent at 65 hours and combusts. At the time of combustion, all 

the necessary conditions are satisfied; the hydrogen concentration is 10 percent, the oxygen is 

10 percent, and the steam is less than the 55-percent threshold for inerting. The hydrogen 

combustion is sufficient to fail the blowout panels and the roof allowing fresh air to enter the 

refueling room. The fresh air circulates into the SFP, which leads to a rapid fuel heatup and 

failure in Ring 1 and then in Ring 2. The reactor building decontamination factor approaches 

unity (Figure 86) resulting in about a 17-percent cesium release to the environment (Figure 87).

The response for the low-density case was similar to, but less severe than the high-density 

case. The spacing of the fuel with empty rack cells reduced the propensity for propagation of 

the heat from the highest decay heat assemblies to the other assemblies in the SFP. Figure 88

shows the response of the highest decay heat assemblies in Ring 1. The peak fuel 
141

temperatures were less than 1,400 K. As shown in Figure 89, the fuel in Ring 3 had a similar 

response, but the fuel in Ring 5 was substantially lower. Fewer fuel assemblies and lower peak 

temperatures resulted in less oxidation and less hydrogen generation. The peak hydrogen 

concentration was well below the threshold for combustion. The overall cesium release is an 

order of magnitude lower (1.7 percent) than in the high-density case.

Figure 81 Water level for unmitigated high-density small leak (OCP2)

Figure 82 SFP power for unmitigated high-density small leak (OCP2)
142

Figure 83 Ring 1 clad temperature for unmitigated high-density small leak (OCP2)

Figure 84 Midplane clad temperature for unmitigated high-density small leak (OCP2)
143

Figure 85 Reactor building mole fraction for unmitigated high-density small leak (OCP2)

Figure 86 Reactor building DF for unmitigated high-density small leak (OCP2)
144

Figure 87 Cesium environmental release fraction for unmitigated high-density small leak 

(OCP2)

Figure 88 Ring 1 clad temperature for unmitigated low-density small leak (OCP2)
145

Figure 89 Midplane clad temperature for unmitigated low-density small leak (OCP2)

Unsuccessful Deployment of Mitigation for Moderate Leak (OCP3) Scenario

Figure 90 (compare to Figure 67 for OCP1) depicts the response of the fuel temperature in 

Ring 1. The heatup starts sooner because there is less water to drain and the approach to a 

zirconium fire is more gradual because of lower decay heat (i.e., by a factor of 2.5—see Table 

25) and the natural circulation of air through the assemblies. However, once the zirconium fire 

is started, the maximum temperatures are comparable in both cases. As shown in Figure 90, 

the zirconium fire starts at Level 5 but then moves slowly to Level 4, Level 3, and Level 2. After 

the peak temperature at Level 4, the peak temperature in the zirconium fire front decreases with 

each successive level. Radial heat transfer from the fuel racks to the SFP wall (Figure 91), the 

buildup of the oxide layer on the fuel, and the depletion of the oxygen in the reactor building 

(Figure 92) cause the clad temperature to decrease. After 24 hours, the fuel temperatures in 

Ring 1 are relatively stable. There was no hydrogen combustion in this calculation. When the 

hydrogen concentration peaks at 8 percent, the oxygen concentration is only 3 percent (well 

below an amount sufficient for combustion as shown in Figure 92). 

Figure 93 shows the temperature profiles for the low-density case. The low-density 

temperatures are about 400 K lower than the high-density case, with the total cesium release 

being about 0.1 percent compared to 0.7 percent in the high-density case. Similar to the 

previous OCP2 case, the low amount of fuel and the empty rack cells reduced the magnitude of 

hydrogen and the cesium release. A sensitivity analysis was performed to examine the effect of 

higher vapor pressure for the air-oxidizing ruthenium releases. Figure 94 (the default ruthenium 

release model) and Figure 95 (the enhanced ruthenium release model used in the present 
146

study) show that the ruthenium release differs by an order of magnitude.

37 All the calculations 

with moderate leaks were based on the enhanced ruthenium release model.

Figure 90 Ring 1 clad temperature for unmitigated high-density moderate leak (OCP3)

 37 However, ruthenium release differences could be higher for scenarios in OCP1 and OCP2.
147

Figure 91 SFP wall liner temperature for unmitigated high-density moderate leak (OCP3)

Figure 92 Reactor building mole fractions for unmitigated high-density moderate leak 

(OCP3)
148

Figure 93 Ring 1 clad temperature for unmitigated low-density moderate leak (OCP3)

Figure 94 MELCOR default ruthenium release for unmitigated high-density moderate 

leak (OCP3)
149

Figure 95 MELCOR enhanced ruthenium release under air oxidizing conditions for 

unmitigated high density moderate leak (OCP3)

6.3.3 Source Terms for Offsite Consequence Analysis

Table 27 summarizes the release characteristics and key events for the high-density scenarios, 

and Table 28 summarizes these factors for the low-density scenarios. Previous sections of this 

report provided a more detailed discussion of key phenomena for selected sequences. The 

releases are binned for offsite consequence analysis, which Section 7 describes.

For the high-density loading, all of the mitigated scenarios (except OCP1) have no release, 

either because the makeup exceeds the leak rate, as in the small leak cases, or the mitigation is 

successful in limiting the fuel heatup and avoiding gap release. All the scenarios that do not 

involve a hydrogen deflagration have relatively low releases since the depletion of the oxygen 

limits clad oxidation and fuel heatup. A building failure results in air ingress into the assemblies 

and late-phase rapid oxidation.

None of the scenarios in the low-density cases had hydrogen combustion, and the releases 

were relatively small. In the absence of hydrogen deflagration, the release fractions for both 

high-density and low-density cases are generally comparable. One exception is the low-density 

OCP1 cases which had higher release fractions than the high-density cases in some instances. 

This difference resulted from more rapid heatup of the fuel in Ring 1 because of less efficient 

heat transfer to the outer assemblies. As shown above, the inventories in the low-density 

configuration are lower and, for the same release fractions, the released activity would be lower. 

Overall, for the moderate leaks, the low-density cases lead to earlier gap release because of a 

larger inventory of water (assemblies removed) resulting in longer times for clearing the 

baseplate. The gap release first occurs in Ring 1 (hot assemblies), which has the same decay 

heat in both high-density and low-density configurations.
150

6.3.4 Accumulation of Water Elsewhere in the Reactor Building

There are approximately 50 floor drains on the refueling floor, both at floor level and in the 

lower, recessed areas of the floor. The two stair towers are fully enclosed and will not be 

subjected to condensation. The doors to the stair towers are secondary containment doors, and 

so they have air seals (weather stripping, but not watertight seals). The open crane hatch in the 

refueling floor has a surrounding 6-in. (0.15-m) curb, so condensation on the floor will run to the 

floor drains and not the hatch. However, condensation forming directly over the hatch, which is 

17 ft 0 in. (5.2 m) by 21 ft 9 in. (6.6 m), will fall to Elevation 135 ft (41.2 m). There is a 

4-in. (0.1-m) floor drain directly under the hatchway at Elevation 135 ft (41.2 m), with no 

equipment in the footprint of the hatch.

If the stainless steel liner plate and 6-ft- (1.83-m-) thick reinforced concrete slab of the SFP leak 

through, the water will fall onto Elevation 165 ft (50.3 m) of the reactor building. Directly 

beneath the SFP on Elevation 165 ft (50.3 m) are the three fuel pool cooling pumps, the three 

fuel pool cooling heat exchangers, and the three fuel pool service water booster pumps. There 

are several floor drains in this area. Equipment adjacent to this area that could be affected by a 

large volume of water includes the fuel pool equipment panel and the reactor level and pressure 

instrument racks. If the floor drains on Elevation 165 ft (50.3 m) cannot keep up with the flow, 

then the alternate flow paths would be the crane hatch or the door to each of the two stair 

towers, having the same configuration as described for the refueling floor above. A significant 

flow rate could also affect the emergency auxiliary load centers on Elevation 165 ft (50.3 

m). Water flowing over the curb of the crane hatch would reach Elevation 135 ft (41.2 m), 

where it would either enter the floor drains, flow through the grating to the torus room floor, or 

exit the building under the doors of the equipment access lock. Water reaching the stair towers 

would travel to the bottom of the stair tower. Water in the W stair tower would reach the 

residual heat removal pump room which has its own floor drains (procedurally controlled, 

normally closed) and sump pump. Water in the east stair tower would reach the core spray 

pump room or elevator shaft bottom which have floor drains (procedurally controlled, normally 

closed) that run to the reactor building main sump.

The reactor building MELCOR model is simplified (see Figure 42). Therefore, all water 

leakages corresponding to the SFP damage and draindown and overflow from water 

accumulation from condensation are directed to the environment. The model does not track the 

flow of the water and accumulation in other parts of the reactor building.
151

Table 27 Summary of Release Characteristics for High-Density Scenarios

High 

Density 

Case #

Scenario Characteristics Release Characteristic

SFP 

Leakage?

50.54(hh)(2) 

Equipment?

Fuel 

Uncovery

(hr)

Gap 

Release

(hr)

Hydrogen 

Deflagration

(hr)

Cs release 

at 72 hours

Cs-137 

(MCi) 

Released

I release 

72 hour

OCP1

None Yes

None No

Small Yes

Small No 39.7 54.2 No 0.6% 0.33 3.5% 

Moderate Yes 7.4 15.1 No 0.5% 0.26 5.0% 

Moderate No 5.9 8.7 No 1.5% 0.80 2.1% 

OCP2

None Yes

None No

Small Yes

Small No 42.6 60.5 64.8 17.1% 7.90 17.1% 

Moderate Yes

Moderate No 5.9 11.6 No 1.6% 0.73 2.0% 

OCP3

None Yes

None No

Small Yes

Small No 18.7 40.6 47.3 42.0% 24.20 51.2% 

Moderate Yes

Moderate No 2.5 16.9 No 0.7% 0.39 0.7% 

151
152

Table 28 Summary of Release Characteristics for Low-Density Scenarios

Low 

Density 

Case #

Scenario Characteristics Release Characteristic

SFP 

Leakage?

50.54(hh)(2) 

Equipment?

Fuel 

Uncovery

(hr)

Gap 

Release

(hr)

Hydrogen 

Deflagration

(hr)

Cs release 

at 72 hours

Cs-137 

(MCi) 

Released

I release 

72 hour

OCP1

None Yes

None No

Small Yes

Small No 40.3 54.7 No 3.1% 0.33 4.6% 

Moderate Yes 7.4 12.6 No 1.8% 0.19 7.0% 

Moderate No 5.9 8.7 No 0.5% 0.05 1.7% 

OCP2

None Yes

None No

Small Yes

Small No 43.1 59.2 No 1.7% 0.28 3.3% 

Moderate Yes

Moderate No 5.9 10.5 No 0.4% 0.07 0.7% 

OCP3

None Yes

None No

Small Yes

Small No 18.8 41.6 No 0.6% 0.10 1.2% 

Moderate Yes

Moderate No 2.5 15.2 No 0.1% 0.02 0.2% 

152
153

7. OFFSITE CONSEQUENCE ANALYSIS

In the unlikely event of a severe accident that might damage the SFP (as detailed in the 

previous sections), a release of radioactive material from the nuclear power plant site into the 

atmosphere could occur. Such a release of radioactive material is expected to disperse from the 

site through the atmosphere and to the surrounding population, by expanding and moving 

downwind. After modeling the onsite accident progression and potential mitigation measures, 

the MELCOR Accident Consequence Code System, version 2 (MACCS2) code is used to model 

offsite release and consequences of radioactive material. MACCS2 (SNL, 1997) has been 

developed by SNL for the NRC over the past two decades. It has the ability to evaluate the 

impacts of atmospheric releases of radioactive aerosols and vapors on human health and on 

the environment. The MACCS2 code can use site-specific weather conditions, population data, 

and evacuation plans to calculate and model the radiation exposure of the population through all 

of the relevant dose pathways—cloudshine, inhalation, groundshine, and ingestion. Along with 

MACCS2, SNL has also developed WinMACCS for the NRC. WinMACCS is a user friendly 

graphical interface to MACCS2 that facilitates selection of input parameters and sampling of 

uncertain input parameters and performs post processing of results. 

MACCS2 rev. 3.7.0 was used for the offsite consequence analysis in this study. In addition, 

many of the input values for offsite release and consequence modeling are based on

approaches developed in the “State-of-the-Art Reactor Consequence Analyses” research 

project (NUREG-1935). These approaches are documented in greater detail in NUREG/CR7009, “MACCS2 Best Practices as Applied in the State-of-the-Art Reactor Consequence 

Analyses Project”, (expected to be published in 2013), The modeling for NUREG-1935 was, in 

turn, based on previous studies such as NUREG-1150, an expert elicitation of the 

NRC/Commission of the European Communities (CEC) to update certain transport and dose 

parameters in NUREG/CR-7161 (NRC, 2013), and an update of the dose coefficients and doseresponse modeling to be consistent with the latest Federal guidance report (FGR) at the time 

(FGR-13, “Cancer Risk Coefficients for Environmental Exposures to Radionuclides,” issued in 

2002 (EPA, 2002)). Differences between the approaches used in NUREG-1935 and the 

approaches used in this study are documented below. 

7.1 Offsite Consequence Modeling

7.1.1 Radiological Source Term

A source term definition for MACCS2 was created for each accident consequence calculation. 

The activity levels of different radionuclides from the fuel in the pool were supplied by ORIGEN 

calculations. The physical state of the plume, including information on the chemical group 

release rates, aerosol size distributions, density, and mass flow rate was supplied by MELCOR. 

The MELCOR analyses provided a release rate for each chemical group. Because the amount 

of release may differ for different sections of the pool, a new methodology was developed for 

this study to account for the distribution of radionuclides in the pool as well as radionuclidespecific release magnitudes. For instance, recently discharged fuel, which has more short-lived 

radionuclides, is more likely to release before and to greater magnitudes than older fuel. This 

process is described in more detail in Section 6.1.5. 

Because explicit modeling with MACCS2 of all release sequences generated by MELCOR 

analyses is computationally expensive, the MELCOR sequences were binned by their Cs-137 

and I-131 release activities (see Table 29). The first criterion used to bin the sequences was 
154

Cs-137 release, because Cs-137 is the most significant contributor to long-term consequences. 

I-131 was also chosen as a criterion to bin the sequences, because I-131 is a good indicator for 

short-lived radionuclides that may be released from recently discharged spent nuclear fuel. The 

tally into each of these bins can be seen in Table 30.

Table 29 Release Category Types

Release Category 

Binning

Cesium-137 Release 

Activity (MCi)

0 to 0.5 0.5 to 5 5+

Iodine131 

Release 

Activity 

(MCi)

0 to 0.25 RC11 RC21 RC31

0.25 to 

0.55 RC12 RC22 RC32

0.55+ RC13 RC23 RC33

Table 30 Release Category Tally

Release 

Category RC11 RC12 RC13 RC21 RC22 RC23 RC31 RC32 RC33 Total

Sequence 

Tally 5 5 0 2 0 0 0 0 2 14

One sequence was chosen from each bin (not including bins with no contributing accident 

sequences) to represent the entire release category, and the offsite consequences of these 

sequences were analyzed. The study considered a number of different factors to determine 

which sequence should represent each bin, including the release frequency, the relative Cs-137 

and I-131 release for the bin, the start time of release, the SFP loading configuration, and the 

availability of the source term data (some accident progression calculations were still ongoing at 

the time the selection was made). In addition, because of the significant differences in release 

category 33 relative to the other bins, both of these sequences were analyzed, as identified in 

Table 31. Then, based on their conditional probabilities, all the main MELCOR sequences and 

their associated consequences were applied to the scenarios reported in the results, which are

high-density and low-density loading both with and without successfully deployed 10 CFR 

50.54(hh)(2) equipment. Sequences with no release were not included, as they do not have 

offsite consequences. Section 6.3 contains more information regarding which sequences do 

and do not have a release. For all sequences, successful deployment of 10 CFR 50.54(hh)(2) 

equipment prevents release of radioactive material, except for a moderate size leak during 

OCP1 (as defined in Section 5.2), which is when newly discharged fuel is first loaded from the 

reactor. Without successful deployment of 10 CFR 50.54(hh)(2) equipment, the predicted 

scenario-specific release frequency is 10-7 per year.
155

Table 31 Listing of Scenario-specific Release Sequences

High Density (1x4) Fuel Loading

Unsuccessful mitigation Deployed 50.54(hh)(2)

Sequence

Release 

Frequency 

(/yr)1

Release 

Category Sequence

Release 

Frequency 

(/yr)

Release 

Category

OCP1

small 

leak 6E-09(2) RC12* OCP1 mod leak 6E-09 RC12

mod 

leak 6E-09 RC21

OCP2

small 

leak 2E-08 RC33*

mod 

leak 2E-08 RC21* No Release

OCP3

small 

leak 4E-08 RC33*

mod 

leak 4E-08 RC11

Total 1E-07 Total 6E-09

Low Density Fuel Loading

Unsuccessful mitigation Deployed 50.54(hh)(2)

Sequence

Release 

Frequency 

(/yr)

Release 

Category Sequence

Release 

Frequency 

(/yr)

Release 

Category

OCP1

small 

leak 6E-09 RC12 OCP1 mod leak 6E-09 RC12

mod 

leak 6E-09 RC11

OCP2

small 

leak 2E-08 RC12

mod 

leak 2E-08 RC11 No Release

OCP3

small 

leak 4E-08 RC11*

mod 

leak 4E-08 RC11

Total 1E-07 Total 6E-09 1 Release frequency = initiating event frequency * ac power fragility * OCP probability * liner fragility for the specified 

leak size (see Section 5.6.3 for conditional probabilities) 2 Example calculation: 1.7x10-5

/yr ∙ 0.84 ∙ 0.0086 ∙ 0.05 = 6x10-9/yr

* Sequences marked with an (*) were used in MACCS2 analysis

7.1.2 Atmospheric Modeling and Meteorology

The atmospheric transport and dispersion model in MACCS2 is a straight-line Gaussian plume 

segment dispersion model. For this study, the atmospheric release of radionuclides is

discretized into (at longest) 1-hour plume segments. This accounts for variations in the release 

rate, as well as for changes in wind direction. More plume segments increase the resolution of 

the dispersion modeling to the point the resolution corresponds to the time resolution of the 

weather data, because each segment can travel in a compass direction representative of the 

actual weather data at the time the plume segment is released. 

A set of aerosol deposition velocities, combined with the aerosol size distribution from 

MELCOR, determines the rates aerosols are deposited from the plume to the ground. 

Generally, the larger aerosols deposit more quickly and so are depleted more rapidly from the 

plume. The peak in the aerosol size distribution is usually a few microns in diameter, which 

corresponds to a deposition velocity of about 4 or 5 millimeters per second. Dry deposition 

velocities have been updated to account for a more typical surface roughness of 60 cm for the 

reference plant site. (A surface roughness of 10 cm was used for NUREG-1935 and a 60-cm 

surface roughness was considered in a sensitivity calculation.) The relative aerosol deposition 
156

velocities, as well as much of the non-site-specific data for acute health effects, are developed 

from NUREG/CR-7161 (NRC, 2013).

Because the exact weather conditions that would apply in the case of a potential accident in the 

future cannot be known in advance, MACCS2 accounts for weather variability by analyzing a 

statistically significant set of weather trials. Thus, the modeled results are ensemble averages

of weather that represent of the full spectrum of meteorological conditions. The weathersampling strategy for this study uses a nonuniform weather-binning approach. Weather binning 

is an approach used in MACCS2 to categorize similar sets of weather data based on 

windspeed, stability class, and the intensity and timing of precipitation. This sampling strategy 

was chosen to represent the statistical variations of the weather. Further discussion on this 

approach can be found in NUREG/CR-7009. 

Meteorological data used for this project consisted of one year of hourly meteorological data 

(8,760 data points for each meteorological parameter). The data are from onsite meteorological 

tower observations are the same as those used in NUREG-1935. The site selected for the 

reference plant provided two years of weather data, including directly measured hourly 

precipitation data. Stability class data were derived from temperature measurements at two 

elevations on the site meteorological towers. The specific year of meteorological data chosen 

for the reference plant was 2006, which was based on data recovery (greater than 99 percent 

being desirable) as documented in NUREG/CR-7009. Different trends (e.g., wind rose pattern 

and hours of precipitation) between the years were estimated to have a relatively minor (less 

than 25 percent) effect on the final NUREG-1935 results. More specific details of the weather 

data can be found in NUREG/CR-7009.

7.1.3 Exposure, Dosimetry, and Health Effects Modeling

MACCS2 considers groundshine, cloudshine, inhalation, and ingestion exposure pathways. 

The principal exposure pathway to members of the public occupying land contaminated by 

atmospheric deposition of radioactive materials is expected to be exposure of the whole body to 

external gamma radiation. Although it is normally expected to be of lesser importance, the 

inhalation pathway contributes additional doses to internal organs (EPA, 1992), especially 

during the emergency phase of the accident. The MACCS2 outputs for health effects and 

population dose include doses from exposure via the ingestion pathway. However, the 

MACCS2 code does not represent these consequences in the individual LCF risk results38. 

Food ingestion parameters were chosen to be consistent with Sample Problem A, as 

documented in NUREG/CR-6613, Vol. 1 (Chanin and Young, 1998). Sample Problem A is 

based on NUREG-1150, with the exception that newer options not included in the older MACCS 

model were used to demonstrate new capabilities in MACCS2 (e.g., that the food ingestion 

model is updated to use the newer COMIDA2 rather than the original MACCS food-chain 

model). NUREG-1935 did not include exposure to contaminated food because staff judged it 

not to be a significant contributor to individual risk. 

NUREG/CR-7009 reviews the shielding factors applied to evacuation, normal activity, and 

sheltering for each dose pathway (e.g. groundshine) used in NUREG-1150 (NRC, 1990) and 

NUREG/CR 6953, Volume 1, “Review of NUREG-0654, Supplement 3, ‘Criteria for Protective 

Action Recommendations for Severe Accidents’—Focus Group and Telephone Survey,” issued 

 38 Including the ingestion pathway is predicted to increase health effect risks in this study by about 5% 

with an LNT dose response model, depending on the scenario.
157

October 2008 (NRC, 2008c). This study uses the same shielding factors as updated in 

NUREG/CR-7009. 

The site file containing population and economic data was created for 16 compass sectors and 

then interpolated onto a 64 compass-sector grid for better spatial resolution for the 

consequence analysis. Site population data have been projected to the target year 2011 using 

the latest version of the computer code SECPOP2000 (SNL, 2003). SECPOP2000 uses 2000 

census data and applies a multiplier to account for population growth and an economic 

multiplier to account for the value of the dollar to create site data for MACCS2. A multiplier 

value of 1.1051 from the U.S. Census Bureau was used to account for the average population 

growth in the United States from 2000 to 2011. Consistent with the approach used in NUREG1935, the economic values from the database in SECPOP2000 (which uses an economic 

database based on the year 2002) were scaled to account for price escalation between the 

years 2002 and 2011. A scaling factor of 1.250 was derived based on the Consumer Price 

Index (CPI). 

Consistent with NUREG-1935, the dose and risk coefficients and relative biological 

effectiveness values used in this study are based on FGR-13 (EPA, 2002). The dose 

coefficients allow organ-specific doses to be calculated from exposure to radiation. The risk 

factors in FGR-13 are based on the risk coefficients for the U.S. population detailed in the BEIR 

V report (NAS, 1990). As implemented in MACCS2, these factors include seven organ-specific 

cancers plus residual cancers not accounted for directly. The inhalation factors in FGR-13 were 

processed to account for a distribution of particle sizes. An activity median aerodynamic 

diameter of 1 micron was assumed with a log-normal form for the distribution and with a 

geometric standard deviation of about 2.5. Parameters that relate to acute health effects in this 

study, as well as much of the nonsite-specific data used for consequence analysis were taken 

from NUREG/CR-7161 (NRC 2013). All of the input parameters extracted from the expert 

elicitation are median values. 

The FGR-13 coefficients, as implemented in MACCS2, include a dose and dose rate 

effectiveness factor (DDREF), which has been incorporated in the dose-response modeling for 

the long-term phase of the offsite consequences and to the dose-response modeling for the 

early-phase (i.e. the first week) for doses less than 20 rem. This factor accounts for the fact that 

protracted, low doses are estimated to be less effective in causing cancer than more acute 

doses. The DDREFs for all cancer types, except for breast, were 2.0; the DDREF for the breast 

was 1.0, as in NUREG/CR-7009. 

To provide perspective on uncertain low-dose health effects, the results also include dose 

truncations that limit the quantified health effects to those arising from higher doses. Dose 

truncation values used here include 620 mrem/year (representative background radiation 

including average annual medical exposures), and 5 rem/year with a 10-rem lifetime cap (based 

on the Health Physics Society’s position that there is a dose below which, because of 

uncertainties, a quantified risk should not be assigned). This approach is consistent with the 

approach used in NUREG-1935.

7.1.4 Emergency Response Modeling

The MACCS2 models were set up to calculate exposures in two distinct phases: the emergency 

phase and the long-term phase. The emergency-phase models calculate the dose and 

associated health effects to the public, as well as the effects of emergency preparedness 
158

actions that protect the public. The emergency phase is defined as the seven day period 

following the start of the release.

The staff modeled offsite response organization (ORO) decision making based upon the 

accident sequences, timing, radiological release, and knowledge of response activities and the 

availability of emergency response technical support. Since actions beyond the emergency 

planning zone (EPZ)

39 would be ad hoc, there is no procedural guidance or exercise 

performance documentation upon which to base assumptions. However, state and local OROs 

have shown long standing capability and understanding of response to hypothetical radiological 

accidents. The accidents modeled in the SFPS are slow to develop relative to the accident 

scenarios used in evaluated exercises. Additionally, there would be national level assistance to 

help civil authorities with protective action decision making. While alternative timing could be 

assumed the staff used a best estimate approach to modeling ORO decision making for 

protective actions beyond the EPZ. 

For each of the accident sequences, staff determined that a General Emergency would be 

declared promptly (within 15 minutes), based on the emergency action levels for the operating 

reactor. The timing of significant radiological release varied among the accident sequences and 

was an important factor in the response modeling. A release from a SFP with a moderate leak 

begins earlier than a damage state with a small leak, but these still do not begin until evacuation 

is well underway or completed within the EPZ.

A number of different protective actions can be modeled in MACCS2. The residents are 

modeled as groups (known as cohorts) and have different types of protective actions and 

associated response timings. The actions that can be taken include:

Shelter-in-place (SIP): For certain areas where dose may be reduced below the PAG through 

sheltering, SIP is modeled as an expected protective action consistent with the emergency 

plans. In other areas, sheltering can occur prior to evacuation.

General Public Evacuation: Residents evacuate the affected area when the official order to 

evacuate is received. 

Early Evacuation: Residents evacuate after the earthquake, but before the official order to 

evacuate is received.

Shadow Evacuation: Residents evacuate from areas that are not under an official evacuation 

order. A shadow evacuation typically begins when a large scale evacuation is ordered (NRC, 

2005b). In a national telephone survey of residents of EPZs, about 20 percent of people that 

had been asked to evacuate had also evacuated for situations in which they were asked not to 

evacuate (NRC, 2008c). In the SFP project, the initiating event is an earthquake that would be 

felt by residents of the EPZ. The event would be followed with media information related to an 

accident at the nuclear power plant, widespread loss of power and damage to some buildings. 

It was assumed that these factors would increase the shadow evacuation to 30 percent of the 

public in the environs of the plant. 

 39 EPZ in this study refers to the plume exposure pathway EPZ with a radius of about 10 miles from the 

reactor site. This should not be confused with the ingestion exposure pathway EPZ with a radius of about 

50 miles from the reactor site. 
159

Hotspot and Normal Relocation: Models are included in the MACCS2 code to reflect emergency 

relocation of people from areas that were not included in the evacuation order where the dose 

exceeds emergency-phase PAGs. Within the MACCS2 calculation, individuals who would be 

relocated because their projected total committed dose is projected to exceed the protective 

action criteria are assumed to not receive any additional dose following relocation for the 

duration of the emergency phase. This relocation dose criterion is applied at a specified time 

after plume arrival within the affected area and is applied to the entire population within the 

analysis area, including the nonevacuating cohort (0.5% of the population) within the EPZ. The 

hotspot and normal dose and time values were developed for each evacuation model. They

were established with the assumption that relocation begins after the evacuation is substantially 

complete which depends on the timing of the first plume for each sequence. For the larger 

release sequences which affect areas beyond 30 miles, the normal relocation time was 

assumed to be 12 hours after the hotspot relocation time. This assumption provides time for 

offsite response organizations to address the higher priority hotspot areas first.

The detailed emergency plans developed for the EPZ provide a substantial basis for expansion 

of response efforts if necessary (NRC, 1980a). This study identified many potential accident 

sequences and performed preliminary consequence modeling to establish baseline dose 

projections as a function of distance. This information was used to develop the appropriate 

emergency response parameters for the release being modeled. The distance to which the 

PAG may be exceeded assisted in determining the extent of offsite protective actions and the 

type of protective actions (sheltering or evacuation) that would be implemented. In the event of 

model predictions of elevated doses at distances beyond the plume exposure pathway EPZ, a 

review of the State emergency response plans was performed to determine the types of 

protective actions that would be implemented in these areas. The results of the dose projections

were binned based on the EPA’s PAGs (EPA, 1992) to support an efficient use of detailed 

consequence modeling to determine the potential effects of such accidents. For this analysis,

the PAG was considered to be exceeded if the four day projected dose is expected to exceed 

one rem for a member of the public. Using the dose projections, the three evacuation models 

presented in the table below were developed for analysis. Detailed information on the 

implementation of these evacuation models is provided in APPENDIX A.
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Table 32 Summary of Evacuation Models

Evacuation 

Model

4-Day Dose 

Projection EPZ Area beyond EPZ 

1 Small: Does not 

exceed PAG 

beyond EPZ.

General public

evacuation, 

including early 

evacuation of 

30% of the 

public. 

Shadow evacuation of 30% of the public from 

immediately beyond the evacuation area. Hotspot 

relocation is 5 rem at 4 hours after plume arrival. 

Normal relocation is 1 rem at 8 hours after plume 

arrival.

2 Large (48 hour): 

Exceeds PAG 

beyond EPZ.

General public

evacuation, 

including early 

evacuation of 

30% of the 

public. 

Shadow evacuation of 30% of the public from 

immediately beyond the evacuation area. Delayed 

evacuation to a distance of 30 miles. Shelter in 

place (SIP) for the 30 to 40-mile area. Shadow 

evacuation of 20% of the public from the SIP area. 

Hotspot relocation is 5 rem at 4 hours after plume 

arrival. Normal relocation is 1 rem at 16 hours

after plume arrival. (Rapid implementation of 

relocation is based on having 48 hours to prepare 

before release begins)

3 Large (24 hour): 

Exceeds PAG 

beyond EPZ.

General public

evacuation, 

including early 

evacuation of 

30% of the 

public. 

Shadow evacuation of 30% of the public from 

immediately beyond the evacuation area. Delayed 

evacuation to a distance of 30 miles. Shelter in 

place of 30 to 40-mile area. Shadow evacuation of 

20% of the public from this Shelter-in-place (SIP)

area. Hotspot relocation is 5 rem at 26 hours after 

plume arrival. Normal relocation 1 rem at 38 hours

after plume arrival.

The population was divided into multiple cohorts to better represent the response of the public. 

A cohort is a population group that mobilizes or moves differently from other population groups. 

The site specific evacuation time estimate provides information on population characteristics, 

mobilization of the public, special facilities, transportation infrastructure and other information 

used to estimate the time to evacuate the EPZ. The evacuation time estimate was used to 

inform the development mobilization times and travel speeds for the public. To model 

evacuation in MACCS2, each cohort was loaded onto the roadway network at a specified time, 

and a set of speed values were applied per cohort for the early, middle and late periods of the 

evacuation. However, evacuations occur as a distribution in which the percent of public 

evacuating the area increases over time until all members of the public have evacuated. The 

rate of evacuating the public is typically represented as a curve that is relatively steep at the 

beginning and tends to flatten as the last members of the public exit the area. The point at 

which the curve tends to flatten occurs when approximately 90 percent of the population has 

evacuated. The last 10 percent of the population is called the evacuation tail (Wolshon, 2010)

and was modeled as a separate cohort. 

An assessment of travel distance and time was initially used to develop the speed of the general 

public cohorts. A distance of 13 miles was assumed as a maximum travel distance to provide for 

the fact that roadways are not necessarily oriented directly outward from the plant. Consistent 

with the location of the reference plant, the analysis includes the State of Pennsylvania position 

that, if an evacuation is ordered, it will include the entire EPZ. This position differs from other 

states, where evacuation of downwind areas would be implemented rather than the full EPZ. 

For this project, a full evacuation was modeled assuming that the offsite response organizations 

from neighboring states would adopt the same protective action decisions. 
161

The following general assumptions were applied in this analysis:

• The EPZ is modeled as the area within 10 miles of the site, as an approximation.

• Protective actions would be implemented within the EPZ were an accident to occur.

• Protective actions would be expanded beyond the EPZ as necessary.

• Dose projections would be developed and available to support protective action 

decisions.

• Residents would expect they cannot return and would take more belongings with them, 

than what was considered in the past, e.g. NUREG/CR-7009, thereby increasing 

mobilization times.

• Residents would generally be aware of an impending emergency through media 

broadcasts.

• For the delayed release sequences in which a releases do not start for more than 24 

hours, schools beyond the EPZ would be closed rather than evacuated.

• Evacuees are transported to safe distances. 

• There is no loss of power beyond 20 miles. Communications, traffic signals, and 

emergency alert system messaging are not impacted in this area.

The chosen time period for the emergency phase begins with the initiating event and continues 

for one week following the initial release. This assumption gives time for the plume to pass and 

deposit radioactive material onto the ground so that all the calculated acute exposures are

captured. The one-week period for the emergency phase is different than the four-day period

used for emergency-phase dose projections, which were used to inform the evacuation models. 

The four-day period was chosen to be consistent with the EPA PAGs (EPA, 1992). 

The roadway network within the EPZ was reviewed against the site-specific evacuation plan to 

determine the likely evacuation direction in each grid element. Travel directions were input at 

the grid level to approximate travel along evacuation routes and primary roadways. For 

evacuations beyond 20 miles, travel directions were chosen to be radially outward to simplify 

modeling of evacuation in these areas. Speed adjustment factors were applied at the grid 

element level to speed up vehicles in the rural uncongested areas and to slow vehicles in more 

urban settings in which the modeling indicates that speeds are lower than the average values 

used in the analyses. 

The MACCS2 potassium iodide (KI) model used in this analysis assumes that KI would be 

distributed only within the EPZ. Half the residents within the EPZ are assumed to have access 

to their KI and to take it within the specified timeframe. 

Adverse weather is typically defined as rain, ice, or snow that affects the response of the public 

during an emergency. Adverse weather was addressed in the movement of cohorts within the 

analysis using an evacuation-speed multiplier to reduce travel speed when precipitation is 

occurring (indicated from the meteorological data file). The evacuation speed multiplier factor 

was set to be 0.7, which effectively slows down the evacuating public to 70 percent of the fairweather travel speed when precipitation exists.

The analyses of the effect of the seismic event on emergency response developed for NUREG1935 were applied in this analysis, as the reference plant in this study was one of the plants 

studied in NUREG-1935. The evaluations of the potential failure of roadway infrastructure 

conducted for NUREG-1935 identified 12 bridges and roadway segments that could fail under 
162

the postulated conditions. The EPZ evacuation routes identified in the emergency plan indicate 

that evacuees west of the river would typically evacuate in a westerly or southerly direction, and 

evacuees east of the river would evacuate in a northerly or easterly direction. Thus, the loss of 

bridges and roads would have a minimal effect on the evacuation time. The other bridges and 

roadways that fail in the earthquake serve sparsely populated areas where alternative roads are 

available. Alternate routes out of the EPZ have more than sufficient capacity to support the 

evacuating population.

The seismic event is assumed to cause the loss of all onsite and offsite power within the EPZ, 

which can affect the response timing and actions of the public. Sirens would be sounded 

following the GE declaration, and because the reference plant will have a fully backed up siren 

system in 2013, it is assumed sirens sound for this analysis. The residents within the EPZ 

would have felt the earthquake, which effectively serves as the initial warning; however, the loss 

of power would affect the number of residents receiving instructions via emergency alert system

messaging. It is expected that the residents use multiple methods of communication, such as 

cell phones, telephones, websites (where power is available), and direct interface to 

communicate the emergency message. 

A review of the roadway network within the EPZ indicates that there are only a few traffic signals 

and that most intersections are controlled with stop signs. The loss of power would cause traffic 

signals to default to a four-way stop mode, which is less efficient than normal signalization. It is

expected that emergency response personnel would respond to these intersections and direct 

traffic as indicated in the site evacuation time estimate. Therefore, the loss of signalization 

should have a limited impact on the evacuation. It is assumed that at distances beyond 20 

miles, there is no loss of power and traffic signals, and emergency alert system messaging is

not impacted.

7.1.5 Long-Term Protective Action Modeling

The long-term phase is the period following the seven-day emergency phase and is modeled for 

50 years. The 50 year duration of the long-term phase has been chosen to provide a 

reasonable time period for calculating consequences from exposure for the average person. 

Exposure is mainly from external radiation from trace contaminants that remain after the land is 

decontaminated, or in lightly contaminated areas where no decontamination was required. 

However, internal exposures may also occur during this period, including inhalation of 

resuspended radionuclides and ingestion of food and water with trace contaminants. 

Depending on the relevant PAGs and the level of radiation, food and water below a certain limit 

could be considered adequately safe for ingestion, and lightly contaminated areas could be 

considered habitable.

A long-term cleanup policy for recovery after a severe accident does not currently exist. The 

actual decisions regarding how land would be recovered and populations relocated after an 

accident would be decided by a number of local, state, and federal jurisdictions and would most 

likely be based on a long-term cleanup strategy, which is currently being developed by the NRC, 

EPA, and other Federal agencies. Furthermore, a cleanup standard may not have an explicit 

dose level for cleanup. Instead, the cleanup strategy may give local jurisdictions the ability to 

develop localized cleanup goals after an accident, to allow for a number of factors that include 

sociopolitical, technical, and economic considerations. Three protective actions were modeled 

to occur for contaminated land during the long-term phase: interdiction, decontamination, and 

condemnation. As used in the MACCS2 model, interdiction and condemnation refer to the 

relocation of people from contaminated areas according to the habitability criterion. Interdiction 
163

is defined in the MACCS2 model as temporary relocation during which the contamination levels 

are reduced by decontamination, natural weathering, and radioactive decay. Condemnation is 

defined in the MACCS2 model as a permanent relocation when contamination levels cannot be 

adequately reduced by decontamination, natural weathering, and radioactive decay. 

Decontamination is modeled in a manner consistent with both NUREG-1150 and NUREG-1935. 

Two levels of decontamination (a decontamination factor of 3 and 15) are each assumed to take 

one year, but the cost of the higher decontamination factor (15) is assumed to be greater, 

reflecting the greater effort needed to achieve the higher level of decontamination. This study 

uses the values in NUREG/CR-7009 for the cost of decontamination. During the 

decontamination period, the land is temporarily interdicted (e.g. the population is temporarily 

relocated), and may be interdicted for an additional period to allow for radioactive decay and 

natural weathering to reduce contamination levels if needed to restore habitability. If land 

cannot be restored to habitability in 30 years, the MACCS2 model defines the land as 

condemned and residents are modeled to not return during the long-term phase. The MACCS2 

models assume that decontamination will only take place if it is projected to make land habitable 

and if the value of the land is greater than the cost to decontaminate. If the level of 

contamination is too high, or if the cost of decontamination is projected to be higher than the 

land value, the individuals on that land are assumed to be permanently relocated. Because 

both the land values and the level of decontamination affect decisions on whether contaminated 

areas can be restored to habitability, they affect predicted long-term doses, health effects, and

economic costs.

Site-specific values are used to determine long-term habitability, whereas farmability is defined 

to be consistent with NUREG-1150. For habitability, most states adhere to EPA guidelines that 

allow a dose of 2 rem in the first year and 500 mrem each year thereafter. However, consistent 

with the location of the reference plant, the analysis includes the State of Pennsylvania position 

using a habitability criterion of 500 mrem per year beginning in the first year, which is the value 

that is used for this study. For consistency and practicality reasons, the same standard for 

estimating habitability is applied to all affected areas in this study. The values used to define 

farmability were taken from NUREG-1150. During the year of the accident, the allowable 

committed dose equivalent from consumption of dairy products to an organ or tissue is 2.5 rem 

(7.5 rem for the thyroid), as well as an additional dose of 2.5 rem (7.5 rem for the thyroid) for all 

other foods. In subsequent years, the maximum allowable dose to the organ or tissue from all 

foods, including dairy products, is 500 mrem (1.5 rem for the thyroid). Agricultural lands 

projected to be contaminated to such an extent that agricultural products would exceed these 

levels are defined to be unfarmable, and the crops growing on these lands at the time of the 

accident are assumed to be disposed. No farming is allowed until the farmability criterion is 

satisfied.

7.2 Offsite Consequence Results

Several consequence metrics have been selected to characterize the impacts resulting from a 

severe spent fuel pool accident. Individual risk of early and latent cancer fatality, as well as 

societal risk of latent cancer fatalities, are measures of the radiological health impact of the 

accident and consistent with NRC’s safety goals (NRC, 1986). In this study, collective dose is 

used as a surrogate for the societal impact of latent cancer fatalities. In addition, certain metrics 

that would influence the values considered by the NRC in regulatory analysis and documented 

in NUREG/BR-0058, such as measures of offsite property damages, the number of displaced 

individuals (either temporarily or permanently), and the extent over which such actions may be 

needed, are also presented. These metrics provide a benchmark for understanding the nature 
164

and extent of a severe spent fuel pool accident. These measures are subject to considerable 

uncertainty, as the details of how long-term protective actions would be carried out would have 

a significant effect on the actual values reported herein. 

All results presented in this section are conditional upon a pool leak following a specified severe 

(0.71g peak ground acceleration) seismic event on the SFP at the reference plant. In the event 

of a pool leak following a severe seismic event, a number of potential outcomes could occur, 

depending upon when in the operating cycle the event occurred, the severity of the leak, and 

whether effective mitigation (in the form of either makeup water or pool sprays) was able to be 

successfully deployed prior to the beginning of the release. Staff has evaluated the likelihood of 

these different conditions. The relative likelihood of a seismic event during a particular 

operating cycle phase is simply proportional to the duration of the phase. The relative likelihood 

of significantly different leak rates is discussed in Section 4. Because these probabilities can be 

quantified with a reasonable degree of certainty, the offsite consequence results are weighted 

by the relative likelihood of these factors to yield an average over the different operating cycle

phases and leak rates.

In contrast, the likelihood of successful deployment of 10 CFR 50.45(hh)(2) mitigation has not 

been quantified. NRC staff judgment is that the likelihood of successful mitigation can in many 

cases be high, but that it is affected by a number of factors that are difficult to quantify (see 

Section 5.3). Related to this, a human reliability analysis (HRA) is provided in Section 8. 

Although the HRA does not provide a quantitative value required to determine the overall 

likelihood of mitigation, it does provide significant insights into the likelihood of mitigation during 

this seismic event for certain damage states. To quantify the overall likelihood of successful 

mitigation, a PRA type analysis would be required. For this reason, the results of the study are 

presented as a range of mitigation effects related to successfully deployed mitigation and 

mitigation that is unsuccessful for 3 days.

This analysis examines the relative effects of a low-density and a high-density fuel loading 

configuration. Therefore, results are reported for two configurations, those being a high-density 

loading case with a 1x4 loading pattern and for a low-density loading case with a mixture of 1x4 

and checkerboard loading patterns, as portrayed in Figure 44 through Figure 48. 

In this chapter, the results for each selected metric are discussed for each loading configuration 

(high-density and low-density). In addition, the factors that affect the results and how those 

results vary with dose truncation assumptions and with distance are discussed. To the extent 

possible, the relationship between the results presented here and the results obtained in 

previous studies is discussed.
165

Table 33 Overall Consequence Results

SFP Fuel Loading High Density (1x4) Low Density

Seismic Hazard Frequency1 (/yr)

(PGA of 0.5 to 1.0g) 1.7E-05 1.7E-05

50.54(hh)(2) Mitigation Credited Yes No Yes No

Conditional2 Probability of Release 0.036% 0.69% 0.036% 0.69%

Hydrogen Combustion Event “Not 

Predicted” “Possible” “Not Predicted”

“Not 

Predicted”

Conditional3 Consequences (Release Frequency-Averaged4)

Cumulative Cs-137 Release at 72 

hours (MCi) 0.26 8.8(8) 0.19(7) 0.11

Measures Related to Health and Safety of 

Individuals

Individual Early Fatality Risk 0 0 0 0

Individual Latent Cancer Fatality 

Risk5 Within 10 Miles 3.4E-04 4.4E-04 3.4E-04 2.0E-04

Measures Related to Cost Benefit Analysis

Collective Dose (Person-Sv) 47k 350k 47k 27k

Land Interdiction6 (mi2) 230 9,400 230 170

Long-term Displaced Individuals6 120k 4,100k 120k 81k

Consequences per year (Release Frequency-Weighted4)

Release Frequency (/yr) 6.1E-09 1.2E-07 6.1E-09 1.2E-07

Measures Related to Health and Safety of 

Individuals

Individual Early Fatality Risk (/yr) 0 0 0 0

Individual Latent Cancer Fatality 

Risk5 Within 10 Miles (/yr) 2.1E-12 5.2E-11 2.1E-12 2.4E-11

Measures Related to Cost Benefit Analysis

Collective Dose (Person-Sv/yr) 2.9E-04 4.1E-02 2.9E-04 3.2E-03

Land Interdiction6 (mi2/yr) 1.4E-06 1.1E-03 1.4E-06 2.0E-05

Long-term Displaced Individuals6

(Persons/yr) 7.1E-04 4.9E-01 7.1E-04 9.5E-03

1 Seismic hazard model from USGS (Peterson et al., 2008) 2 Given specified seismic-event occurs 3 Given atmospheric release occurs 4 Results from a release are averaged over potential variations in leak size, time since reactor shutdown, population 

distribution, and weather conditions (as applicable); additionally, “release frequency-weighted” results are multiplied 

by the release frequency.

5 LNT and population-weighted (i.e., total amount of latent cancer fatalities predicted in a specified area, divided by 

the population that resides within that area.) 6 1st year post-accident; calculation uses a dose limit of 500 mrem per year, according to Pennsylvania Code, Title 

25 § 219.51 

7 Mitigation can moderately increase release size (see Section 6.3); the effect is small compared to the reduction in 

release frequency. 

8 Largest releases here are associated with small leaks (although sensitivity results show large releases are possible 

from moderate leaks). Assuming no complications from other SFPs/reactors or shortage of available equipment/staff, 

Section 8 shows that there is a good chance to mitigate the small leak event.
166

7.2.1 Individual Early Fatality Risk 

For all scenarios, no offsite early fatalities attributable to acute radiation exposure are predicted 

to occur. Due to radioactive decay, spent fuel pools tend to have significantly less shorter-lived 

radionuclides (e.g. I-131) than reactors. Despite this, in at least one case that was analyzed, 

doses close to the site did reach levels that can induce early fatalities. Therefore, the potential 

(although remote) for early fatalities exists. However, emergency response as treated in this 

study effectively prevents any early fatality risk from acute radiation exposure, at least in part 

because the modeled accident progression results in releases that are long compared to the 

implementation of emergency response in the areas of most concern.

The projections of no early fatalities in this study is lower than that reported in some previous 

studies of risks from spent fuel pool accidents, such as NUREG/CR-6451 and NUREG-1738, 

and consistent with the earlier studies documented in NUREG-1353. Tables 4.1 and 4.2 of 

NUREG/CR-6451 project anywhere from approximately one to one hundred early fatalities 

within a 500 mile radius in the event of an accident involving the full spent fuel pool, with the 

higher values associated with high release fractions. NUREG-1738 (Table 3.7-1 and Table 3.7-

2) reported similar values, ranging from no fatalities for low Ruthenium source terms with early 

evacuation to up to 192 early fatalities for an accident shortly (30 days) after shutdown with high 

Ruthenium source terms and late evacuation. NUREG-1353 does not provide quantitative 

estimates of early fatality risk but states that “...there are no “early” fatalities and the risk of early 

injury is negligible”. On balance, the scenarios analyzed here are consistent with the lower end 

of the reported range from previous studies, in that no early fatalities are projected to occur. 

7.2.2 Individual Latent Cancer Fatality Risk

Despite the large releases in certain circumstances, the risk of latent cancer fatality to the 

average individual within 10 miles of the plant is low. When averaged over the likelihood of 

different event timings and leak sizes, the conditional risks within 10 miles are in the 1E-04 to 

1E-03 range for cases both with and without successful 50.54(hh)(2) mitigation and for highdensity and low-density cases, when using an LNT dose response model. This range does not 

appreciably increase even if the releases for different leak sizes or operating cycle phases are 

shown separately. 

Individual latent cancer fatality risk is low because:

• The predicted release frequency of this event is very small

• Protective actions, especially those for long-term chronic doses, are estimated to avert 

significant amounts of public exposure.

Because of the nature of the event, this risk is predominantly from long-term chronic exposures.

With effective long-term protective measures (e.g. temporary and permanent land interdiction), 

essentially no individuals receive any long-term risks greater than those associated with the 

dose limits for protective actions. Therefore, independent of the release magnitude of the event, 

these dose limits form an upper limit to individual long-term risk. In addition, emergency 

response is assumed to be very effective in evacuating and relocating the public. For instance, 

individuals within the 0-10 mile distance (representative of the plume exposure pathway EPZ) 

essentially only receive LCF risk if they return to low risk, habitable areas. The conditional 

individual LCF risks within ten miles are comparable to or lower than the projections from earlier 

studies of spent fuel pool accident risk. For example, NUREG-1738 reports conditional 
167

individual latent cancer fatality risks ranging from 8E-4 to 8E-2 for a range of initiating events 

where large seismic events contributed the most to the overall estimate of risk. These 

conditional risks were driven largely by the previous estimates of ruthenium volatility and by the 

effectiveness of evacuation.

When the release frequency is considered, the latent cancer fatality risks from the events 

analyzed in this study are very small, in the 1E-12 to 2E-11 per year range, when using an LNT 

dose response model. For perspective, the Commission’s safety goal policy related to the 

cancer fatality quantitative health objective (QHO) represents a 2E-6 per year objective for an 

average individual within 10 miles of the nuclear plant site (NRC, 1983). While the results of 

this study are scenario-specific and related to a single spent fuel pool, staff concludes that since 

these risks are several orders of magnitude smaller than the QHO, it is unlikely that the results 

here would contribute significantly to a risk that would challenge the Commission’s safety goal 

policy (NRC, 1986).

Because the health effects that would be induced by low dose radiation are uncertain, staff 

performed a sensitivity analysis to understand how the risks would change if computed health 

risks were limited to those arising from higher doses. The upper truncation level used in this 

sensitivity analysis corresponds to a treatment consistent with the HPS position statement (5 

rem annually and 10 rem lifetime). The second truncation level corresponds to the average 

annual dose to the public from medical and background radiation exposures in the United 

States (620 mrem annually). 

Using truncation levels that do not quantify the effects of doses below the dose levels chosen 

here significantly reduces the estimated individual LCF risk. This is because individual LCF risk 

using an LNT dose response model mainly comes from doses less than those specified in 

protective action guidelines. Table 34 (which shows risk to residents living within ten miles, not 

including risk from ingestion or risk to decontamination workers) shows the use of the dose 

truncations that are analyzed here lowers the estimated individual LCF risk within 10 miles by 

several orders of magnitude. Because the dose truncations are greater than the dose limits for 

land interdiction, it is difficult for doses from the long-term phase to contribute to the quantified 

LCF risk. Therefore, emergency-phase exposures play a more significant role in the doses that 

exceed the truncation levels. However, the amount of early phase exposures that exceed the 

dose truncations is very small within 10 miles because emergency response is effective in 

protecting the evacuees. 

A number of factors can affect quantified individual LCF risks, particularly the very small values 

from dose truncation results. These include potential variations of the real application of 

protective actions, different protective action levels, or consideration of ingestion doses. 

Nevertheless, the overall conclusions that with an LNT calculation, individual LCF risk is mainly 

from long-term chronic exposures, and that dose truncations significantly lower the estimated 

individual LCF risk, remain valid. 
168

Table 34 Dose-Response Model Results (LNT) and Dose Truncation Comparison

Dose-Response High Density (1x4) Low Density

50.54(hh)(2) Mitigation 

Credited Yes No Yes No

Conditional1 Individual Latent Cancer Fatality Risk Within 10 Miles

(Release Frequency-Averaged2)

Linear, No Threshold 3.4E-04 4.4E-04 3.4E-04(3) 2.0E-04

620 mrem/yr truncation 6.1E-08 1.2E-07 6.1E-08(3) 3.4E-08 

5rem/yr or 10rem lifetime 

truncation 1.8E-08 1.4E-07 1.8E-08(3) 5.6E-09

Individual Latent Cancer Fatality Risk Within 10 Miles (/yr)

(Release Frequency-Weighted2)

Linear, No Threshold 2.1E-12 5.2E-11 2.1E-12 2.4E-11

620 mrem/yr truncation 3.8E-16 1.4E-14 3.8E-16 4.0E-15

5 rem/yr or 10 rem 

lifetime truncation 1.1E-16 1.6E-14 1.1E-16 6.6E-16

1 Conditional on a release occurring

2 Results from a release are averaged over potential variations in leak size, time since reactor shutdown, population 

distribution, and weather conditions; additionally, “release frequency-weighted” results are multiplied by the release 

frequency.

3 Mitigation can moderately increase release size (see Section 6.3); the effect is small compared to the reduction in 

release frequency. 

While individual latent cancer fatality risk is very low, it decreases slowly with distance, 

particularly for large releases such as may occur from an accident in a high-density pool with 

unsuccessful mitigation for 3 days. This is because offsite release models predict significant 

spread of contamination to far distances, mainly because of the slow deposition of aerosols 

from the plume. Increasing the magnitude of the release extends the range over which a plume 

can travel before the radioactive inventory of the plume is significantly depleted by deposition. 

Furthermore, because protective actions such as land interdiction are modeled to occur 

wherever the model predicts that the dose limits are exceeded, most distances are held to 

comparably low levels of individual LCF risk regardless of the magnitude of the deposition, as 

was seen in the results for individual LCF risks in Table 33. This can be seen in 

Figure 96, which like the table, is also weighted by the release frequency.
169

1

Linear-no threshold, weather-averaged, release frequency-weighted, and population-weighted

Figure 96 Individual Latent Cancer Fatality Risk (per year)

The accuracy of atmospheric transport and deposition (ATD) models (e.g., the Gaussian plume 

segment model used in MACCS2) tend to decrease with distance, and therefore the results 

should be viewed with caution at longer distances. However, MACCS2 has been benchmarked 

against other ATD models, and the staff considers the broad conclusion remains valid—that 

risks would be small but drop slowly with distance in the event of large releases.

For affected areas with large populations, severe accidents can result in significant numbers of 

latent cancer fatalities. However, this should be weighed against the likelihood of the accident. 

Furthermore, because the individual doses are relatively small, this would be a small fraction of 

all cancer fatalities from all causes. This risk mainly comes from doses that are constrained to 

be less than dose limits for protective actions from an LNT dose response model; dose 

truncations predict significantly fewer latent cancer fatalities. 

Figure 97 compares the quantified individual LCF risk for different dose truncations and for a 

variety of reported distance ranges for a high-density (1x4) configuration with unsuccessful 

mitigation for 3 days. The figure shows that dose truncation significantly lowers the quantified 

LCF risk. This is similar to Table 34; however this figure shows risks for a range of distances. 

1.E-13

1.E-12

1.E-11

1.E-10

0-10 10-20 20-30 30-40 40-50 50-100 100-200 200-500

Latent Cancer Fatatlity Risk (/yr)

Distance from Release (miles)

Average Individual Latent Cancer Fatality Risk1 (Per Year)

High Density (1x4) - 50.54(hh)(2)

Mitigation Not Credited

Release Frequency: 1E-7/yr

Low Density - 50.54(hh)(2)

Mitigation Not Credited

Release Frequency: 1E-7/yr

High Density (1x4) - 50.54(hh)(2)

Mitigation Credited

Release Frequency: 6E-9/yr

Low Density - 50.54(hh)(2)

Mitigation Credited

Release Frequency: 6E-9/yr
170

1 High Density (1x4)—Unsuccessful Mitigation, weather-averaged, release frequency-averaged, and populationweighted 

Figure 97 Conditional Individual LCF Risk for Different Dose Truncations and Distances

The effect of protective actions can be observed from Figure 97. For the release modeled in 

this scenario, the LCF risk within 10 miles is slightly less than at the 10–20 mile range. This 

small variation in risk with distance is because different modeled protective actions (such as 

evacuation, sheltering, early relocation, decontamination, temporary interdiction, and permanent 

condemnation) will depend on the level of contamination expected at a particular location. For 

example, the higher contamination levels closer to the source may result in relatively longer 

periods of relocation. Because no exposure to these populations would occur during this period 

these individuals could have lower overall doses than individuals further away under some 

situations. The 620 mrem annual dose truncation in particular demonstrates the effect of 

reduced individual LCF risk at these distances compared to longer distances. In 

Figure 97, the 620 mrem annual dose truncation best illustrates the effect of emergency 

response because this dose truncation does not quantify the significant contributions from 

chronic, long-term exposures. 

7.2.3 Land Contamination

As the values in Table 33 suggest, conditional on a release (with a frequency of 1E-7 per year, 

or lower) occurring, the total land contamination area can be considerable. The low-frequency, 

large releases are significantly affected by hydrogen combustion events, which are currently 

predicted in some high-density loading situations without successful mitigation for 3 days, but 

not in other scenarios. For relatively small releases from a SFP, the extent of contaminated 

land could range to hundreds of square miles. For a large release, such as a release from a 

high-density pool without successful deployment of 50.54(hh)(2) mitigation that leads to a 

hydrogen combustion event, the amount of contaminated land can be two orders of magnitude 

higher (Table 35 partially reflects this range, although it reports average values). The levels of 

potential land contamination in the event of a release should be weighed against the likelihood 

1.E-09

1.E-08

1.E-07

1.E-06

1.E-05

1.E-04

1.E-03

1.E-02

Individual Latent Cancer Fatatlity Risk (/event)

Distance (miles)

Conditional Individual Latent Cancer Fatality Risk1

Linear, No

Threshold

620mrem/yr

truncation

5rem/yr - 10rem

lifetime truncation
171

of the accident. When the amount of contaminated land is weighted by the annual likelihood of 

occurrence (as seen in Table 33), the expected impact is relatively low. In addition, only a small 

portion of these interdicted areas are expected to be permanently interdicted, as the level of 

contamination is expected to significantly decrease with time as decontamination, radioactive 

decay, and weathering occur. 

The amount of land affected depends on the dose criterion selected. For the purposes of this 

study, land contamination is defined as the area impacted by protective actions, specifically 

either temporary or permanent land interdiction. Because of the location of the reference plant, 

the particular protective action level the study uses is the Pennsylvania standard for habitability 

(dose limit of 500 mrem each year). The study uses this measure to estimate land 

contamination starting in the first year after a potential severe accident. In reality, the annual 

dose limit for what is considered “habitable” can change when crossing a state boundary. 

However, for consistency and practicality reasons, the same standard for estimating land 

contamination area is applied to all affected areas in this study, and the measure chosen for this 

study is only meant to be an indicator of land contamination. 

Consistent with the observations of a relatively slow decline in individual latent cancer fatality 

risk with distance, the results of the analysis indicate that protective actions such as temporary 

relocation may be needed at long distances. The table below displays an average amount of 

interdicted land within different distances for high- (1x4) and low-density fuel loading. 

Table 35 Average Land Interdiction* (square miles per event)

SFP Loading Pattern High Density (1x4) Low Density

10 CFR 50.54(hh)(2)

mitigation credited Yes No Yes No

Release Frequency (/yr) 6.1E-9 1.2E-7 6.1E-9 1.2E-7

0-50 miles 210 1,200 210** 160

0-100 miles 230 3,100 230** 170

0-500 miles 230 9,400 230** 170

* Weather-averaged and release frequency-averaged; Dose limit of 500 mrem per year, according to Pennsylvania 

Code, Title 25 § 219.51 

** Mitigation can moderately increase release size (see Section 6.3); the effect is small compared to the reduction in 

release frequency. 

A release in the high-density fuel loading situation without successful 50.54(hh)(2) mitigation is 

capable of large releases, and therefore an average release from this situation is capable of 

causing significantly more land contamination at longer distances than in the other situations. In 

contrast, releases from situations with low density fuel loading (and/or successfully deployed 

50.54(hh)(2) mitigation equipment) cause a relatively smaller amount of land contamination 

beyond 50 miles, and none beyond 100 miles when using land interdiction as a measurement of 

land contamination. This is because on average, a release in these situations contaminates 

significantly less area. However, because of the release magnitude of any of the analyzed SFP 

releases, the total amount of land contamination that remains within ten miles is relatively small. 

On land contamination, past results are expected to be broadly consistent with this study. 

However some previous studies did not report land contamination and some reported different 

metrics for estimating areas, so a direct comparison is not possible. NUREG/CR-6451 reports 

values for condemned farmland that includes hundreds of square miles within a 50-mile radius 

and thousands of square miles within a 500 mile radius, albeit for a full core off-load. NUREG-
172

1353 reports values for land contamination based on NUREG/CR-4982 that range into the 

hundreds of square miles, albeit largely within a 50-mile radius of the plant. These differences, 

as well as different choices for the land contamination criteria that can significantly affect the 

estimated areas, make a quantitative comparison less meaningful. However, it is clear that both 

this study and past studies have predicted that SFP accidents can lead to significant land 

contamination.

7.2.4 Displaced Individuals

Consistent with the results for land contamination, relatively large numbers of people may be 

impacted following a large release from a spent fuel pool. Displaced individuals, also known as 

relocated individuals, are people who are predicted to be temporarily or permanently relocated 

due to interdiction of contaminated land, based on the dose limit for land interdiction starting in 

the first year following an accident. These individuals are not necessarily the same as 

evacuees, who evacuate during the emergency phase (although an individual could be both of 

these). 

Conditional on a release (with a frequency of 1E-7 per year or lower) occurring, the total number 

of temporarily relocated individuals could be considerable. For relatively small releases of an 

SFP, the number of displaced individuals could range into the hundreds of thousands. For a 

large release, which is predicted in some high-density loading situations early in the operating 

cycle without successful 50.54(hh)(2) mitigation, the number of displaced individuals can be two 

orders of magnitude higher. (Table 36 partially reflects this range, although it reports average 

values). 

Also consistent with the observations related to the amount of land contamination with distance, 

the results of the analysis indicate that protective actions such as temporary relocation may be 

needed at long distances. The table below displays the average number of displaced 

individuals for different distances for high (1x4) and low density fuel loading.

Table 36 Average Number of Long-term Displaced Individuals* (per event)

SFP Loading Pattern High Density (1x4) Low Density

10 CFR 50.54(hh)(2)

mitigation credited Yes No Yes No

Release Frequency (/yr) 6.1E-09 1.2E-07 6.1E-09 1.2E-07

0-50 miles 100k 780k 100k** 72k

0-100 miles 120k 2,000k 120k** 81k

0-500 miles 120k 4,100k 120k** 81k

* Weather-averaged and release frequency-averaged; Dose limit of 500 mrem per year, according to Pennsylvania 

Code, Title 25 § 219.51

** Mitigation can moderately increase release size (see Section 6.3); the effect is small compared to the reduction in 

release frequency. 

These estimates should be weighed against the likelihood of the accident. When the number of 

displaced individuals is weighted by the annual likelihood of occurrence (with a frequency of 1E7 per year or lower; as seen in Table 33), the expected impact is relatively low. An average

release in the high-density fuel loading situation without successful 50.54(hh)(2) mitigation

causes significantly more relocation at longer distances than in the other situations because it is 

capable of larger releases. In contrast, releases from situations with low density fuel loading 

(and/or successfully deployed 50.54(hh)(2) mitigation equipment) cause a relatively small 
173

amount of relocation beyond 50 miles, and none beyond 100 miles because on average, a 

release from these scenarios contaminates significantly less area. For all situations, the 

number of displaced persons from the 0 to 10 mile area is relatively small because the number 

of people living on this area is relatively small. 

 

7.3 Offsite Consequence Comparison

A goal of the study is to compare the results of the scenario-specific, high- and low- density fuel 

loading seismic events. To facilitate the comparison, results of different scenarios are 

compared to each other by dividing the results from one scenario by another scenario, for a 

variety of consequence metrics. The ratios of the consequence metrics are indicators of the 

scenario specific safety benefit between the two scenarios. 

These comparisons should consider the scenario release frequency as well as conditional on a 

release occurring, appropriate. In the first comparison below, the high-density (1x4) fuel loading 

and low-density fuel loading had the same release frequency. Therefore, for this comparison, 

there is no additional reduction when the likelihood of occurrence is also considered. 

Table 37 Consequence1 Comparison – High (1x4) Density / Low Density Loading without 

Successful 50.54(hh)(2) Mitigation

SFP Fuel Loading High Density (1x4) Low Density Reduction Factor 

(dimensionless)

Release Frequency 1.2E-07 1.2E-07 1.0

Individual Latent Cancer 

Fatality Risk2 within 10 Miles 4.4E-04 2.0E-04 2.1

Collective Dose (Person-Sv) 350k 27k 13

Land Interdiction3 (mi2) 9,400 170 56

Long-term Displaced 

Individuals3 (Persons) 4,100k 81k 51

1 Conditional on a release occurring (frequency of 1E-7 per year, or lower); results from a release are averaged over 

potential variations in leak size, time since reactor shutdown, population distribution, and weather conditions 

2 Linear-no threshold, population-weighted 3 1st year post-accident; calculation uses a dose limit of 500 mrem per year, according to Pennsylvania Code, Title 

25 § 219.51 

The most significant reduction factor in a low-density loading pattern is in the reduction in land 

interdiction and associated displaced individuals. This is because these consequences are 

more closely related to the size of release than the other results. In addition, a small amount of 

contamination can occur before land reaches the dose level for interdiction. This dose threshold 

effect means smaller releases more-than-proportionally reduce the amount of land interdiction. 

The reduction in collective dose (and latent cancer fatalities) in a low density loading pattern is 

also due to the smaller release magnitude. This reduction is significant, although not as 

significant as the reduction in land interdiction. This is because larger releases are predicted to 

have considerably more temporary and permanent interdiction to protect the public. This is 

especially true at shorter distances, as indicated by the reduction factor for LCF risk for 0-10 

miles. One significant reason a smaller release magnitude is expected in the low-density 

loading situations is because hydrogen combustions are currently not predicted in these 

situations. 
174

The next table reports the reduction of the consequences with successful deployment of 

50.54(hh)(2) mitigation equipment. Because successfully deployed mitigation can prevent fuel 

release, it affects the reduction factors for release frequency-weighted consequences (per year)

differently than consequences conditional on a release occurring. For brevity, the consequence 

values are not displayed here, although can be seen in the previous section.

Table 38 Consequence Comparison – Unsuccessful/Successful Deployment of 

50.54(hh)(2) Equipment 

Fuel Loading Density High Density (1x4) Low Density

Reduction Factor (dimensionless)

Change in Release Frequency (/yr) 19 19

Conditional1 Consequences

(Release Frequency-Averaged2)

Type of Consequence Reduction Factor (dimensionless)

Individual Latent Cancer Fatality Risk3

within 10 Miles 1.3 0.61

Collective Dose (Person-Sv) 7.4 0.59

Land Interdiction4 (mi2) 40 0.72

Long-term Displaced Individuals4

(Persons) 36 0.70

Consequences per year

(Release Frequency-Weighted2)

Type of Consequence Reduction Factor (dimensionless)

Individual Latent Cancer Fatality Risk3

within 10 Miles (/yr) 25 12

Collective Dose (Person-Sv/yr) 140 11

Land Interdiction4 (mi2/yr) 780 14

Long-term Displaced Individuals4

(Persons/yr) 690 13 1 Conditional on a release occurring (frequency of 1E-7 per year, or lower) 2 Results from a release are averaged over potential variations in leak size, time since reactor shutdown, population 

distribution, and weather conditions; additionally, “release frequency-weighted” results are multiplied by the release 

frequency.

3 Linear-no threshold, population-weighted 4 1st year post-accident; calculation uses a dose limit of 500 mrem per year, according to Pennsylvania Code, Title 

25 § 219.51 

For both high- and low-density fuel loading, the release frequency was reduced by about a 

factor of 20 with successful deployment of 50.54(hh)(2) mitigation. 

Conditional on a release occurring (middle portion of Table 38), successful deployment of 

50.54(hh)(2) mitigation reduces all of the average consequences of the high-density fuel loading

pattern, although to varying degrees. These varying degrees of consequence reductions are 

similar to that predicted in Table 37 for using a low-density loading pattern, although to a 

somewhat lesser extent. A significant portion of this reduction may be attributable to the fact 

that hydrogen combustions are not predicted with successful deployment of 50.54(hh)(2) 

equipment. 
175

Contrary to what might be expected, 50.54(hh)(2) mitigation is predicted to slightly increase the 

average conditional consequences of a release from a low-density fuel loading pattern. While 

successful deployment of 50.54(hh)(2) equipment is usually effective at preventing releases, it is 

not as effective at mitigating release from the low-density fuel loading pattern when deployed in 

a capacity specifically to provide makeup water through injection, as sometimes assumed. In 

these conditions, release from a SFP can sometimes be somewhat larger with deployed 

mitigation. In addition, the situations for which 50.54(hh)(2) equipment prevented release for 

the low-density loading events were the situations with the smallest release magnitudes, which 

has the non-intuitive effect of increasing the average consequence of a release. 

The bottom section of Table 38 shows the combined benefit of prevention and mitigation from 

successfully deployed 50.54(hh)(2) equipment, which combines the reduction factors of a lower 

release frequency with the changes in the average consequences of a release. In the highdensity loading situation, the overall benefit of 50.54(hh)(2) equipment is very significant (more 

than a factor of 100 reduction in most of the risk metrics) if successfully deployed. For lowdensity loading, the deployment of the 50.54(hh)(2) equipment has a somewhat negative effect 

on the average conditional consequence; however, this is far outweighed by the benefit it

provides in preventing release. 
176
177

8. HUMAN RELIABILITY ANALYSIS 

Consistent with the limited scope of the SFPS, a limited scope human reliability analysis (HRA) 

was performed, to develop initial insights into the likelihood of successful operator actions to 

prevent spent fuel damage for the specific seismic event and consequence scenarios studied. 

A full scope HRA would primarily be useful as part of a PRA analysis. A PRA would necessarily 

consider a much broader scope than the SFPS. Such a scope would include the likelihoods of 

all initiating events, the plant damage states for the two reactors and spent fuel pools, the 

availability of all installed or portable mitigation equipment, and the availability of on-site (and 

possibly off-site) personnel. Thus the limited scope HRA results presented here must be 

viewed from the context of their specific assumptions, including assumptions that remove likely 

complexities that impact operator performance. 

In this context, to perform an HRA for this study, successful mitigation must be defined. For this 

HRA, mitigation success is defined as preventing radioactive release from the fuel rods of the 

Unit 3 SFP fuel (or gap release). The reference plant site has two reactor units (Unit 2 and Unit 

3) in operation. The status of the Unit 2 and 3 reactors, the Unit 2 SFP, and the other plant 

SSCs would affect Unit 3 SFP mitigation, but successful mitigation, as defined in this analysis, is 

only determined by the Unit 3 SFP fuel status. 

The effective SFP mitigation strategies, to prevent fuel overheating and release of radioactive 

material from the damaged fuel rods, are to either inject or spray water into the SFP from the 

refueling floor. The refueling floor on top of the reactor cavity is part of the primary containment 

that insulates the refueling floor from the reactor. In situations involving reactor damage with 

intact primary containment, access to the refueling floor is still possible. Over the refueling floor 

is the secondary containment which is a light-weight steel structure. During an SFP accident, 

the secondary containment can reduce the radioactivity released from the SFP to the 

environment. During refueling, the primary containment head is removed to expose the reactor 

cavity. The reactor vessel head is also removed for defueling and refueling. Therefore, during 

a refueling outage, the refueling floor is no longer insulated from the reactor. Heat and radiation 

generated from the reactor would directly affect the work environment on the refueling floor. In 

addition to the strategies of spraying water from the refueling floor to the SFP, strategies to 

spray water from outside of the secondary containment (e.g., by ladder fire trucks) to the 

secondary containment or the SFP (through containment breaches) are available. However, as 

these strategies are aimed at mitigating releases to outside of the secondary containment and 

not at preventing fuel overheating, they are not credited in this HRA study.

The SFPS ran a number of computer simulations to understand the effects of a set of factors 

affecting SFP fuel radioactive release after an earthquake damaged the normal SFP cooling 

system. The set of factors include SFP leak size, spent fuel loading pattern, OCP, mitigation 

deployment, mitigation flow rate, and types of mitigation (i.e., injection or spray). These 

simulations generated information that served as the foundation for the HRA study. Section 8.1 

summarizes the SFPS results relevant to the HRA study and discusses their implications to the 

HRA study. Section 8.2 discusses the staffing, mitigation equipment, strategies, and 

procedures of the reference plant relevant to the SFP mitigation. Section 8.3 discusses the 

HRA study framework, scope, and approach. The conduct of an HRA is normally done in 

conjunction with a PRA to identify each event sequence (i.e., scenario) following an initiating 

event. For each event sequence, the PRA model would explicitly specify the status 

(i.e., success or failure) of each component, system, and human action that affects the event 

sequence’s progression and end consequence. For this reason, the development of a PRA 
178

model would require significant effort. For this limited scope HRA study, a detailed PRA (i.e., 

using event trees branched to represent various possible scenarios) was not performed. 

Instead, scenarios are classified based on the status of a few key SSCs (e.g., electric power 

availability, and the status of the Unit 3 reactor and primary containment). The Unit 3 reactor 

and primary containment status are included in the HRA study because of their significant 

effects on the Unit 3 refueling floor work environment (i.e., where the SFP mitigation strategies 

are performed). Table 39 summarizes the scope and assumptions applied to the HRA study. 

Section 8.4 summarizes the insights of this study.

Table 39 The scope and assumptions of the HRA study

# Scope and assumptions Notes

1 Success criterion: prevent radioactive 

release from the fuel rods of Unit 3 SFP fuel

- Do not include strategies designed to 

reduce radioactivity released to the 

environment. The effective mitigation 

is to inject or spray water into the

SFP.

- The status of fuel in the Unit 2 and 3 

reactors and Unit 2 SFP are not 

considered in the success criterion.

2 Classify plant damage states as a result of 

the earthquake and estimate the mitigation 

failure probability for each plant state.

The probabilities of the plant damage states 

as a result of the earthquake were not 

estimated. 

3 The installed equipment for SFP mitigation is 

not available. Operators have to use the 10

CFR 50.54(hh)(2) equipment for mitigation.

If the installed equipment (e.g., fire system 

and residual heat removal system) is 

available, the SFP mitigation would have a 

much higher success likelihood than this 

study’s estimates.

4 The SFP mitigation uses the minimum flow 

rate specified in NEI 06-12 guidance for 

complying with 10 CFR 50.54(hh)(2).

The actual flow rate is expected to be higher 

than the minimum NEI recommended flow 

rate (i.e., 500 gpm of injection or 200 gpm of 

spray)

5 10 CFR 50.54(hh)(2) equipment and water 

sources for Unit 3 SFP mitigation is 

available.

Earthquake-caused damage to the 10 CFR 

50.54(hh)(2) equipment is not included in the 

study. Further, dividing equipment to 

mitigate multiple reactor and SFP problems 

is not considered.

6 Sufficient plant staff is available to perform 

the Unit 3 SFP mitigation.

Staffing information is discussed but different 

staffing scenarios are not factored into the 

analysis. For multiple reactors and SFPs 

damaged by the hypothetical earthquake, 

the personnel sufficiency would be a key 

factor affecting mitigation success.

7 Non-plant, off-site support (e.g., off site fire 

trucks) are not considered.

For an SFP event, the primary function of 

off-site support is to keep radioactivity 

release within the plant site. Off-site support 

for preventing SFP fuel rod damage is not 

credited.
179

8.1 Summary of Spent Fuel Pool Study Analysis Results Relevant to 

Human Reliability Analysis

8.1.1 High Level Scenarios Classification

The SFPS concludes that the following four scenarios do not lead to gap release with a 

72-hour-truncated simulation time (see Table 40 for a tabulate classification):

(1) Boil off Scenario with No SFP Leaks. As mentioned earlier, the SFP water level in this 

scenario would take more than 7 days to decrease to the top of the fuel rack. Because 

of the long time available for response, multiple opportunities are available to prevent 

damage to the SFP; therefore, the human error probability (HEP) (in this study the HEP 

is equivalent to mitigation failure probability) is negligible. 

(2) Mitigated Scenario for Small Leaks. No fuel damage occurred when the makeup water 

was injected into the SFP at the time specified by the SFPS. The SFPS suggests that, 

as long as the spent fuel is covered with water, SFP failure would not occur. Therefore, 

the available time for operators to respond is longer than the SFPS injection time in 

some scenarios. For these scenarios, the HEPs were calculated.

(3) Unmitigated Scenario in Late Phases (i.e., OCPs 4 and 5). These scenarios have low 

decay heat. Even when relying only on natural air circulation, heat convection and 

radiation, and other natural means of heat transfer, overheating of the spent fuel can be 

prevented. For these scenarios, the HEPs were not calculated.

(4) Mitigated Moderate Leak Scenarios in OCP2, OCP3, OCP4, and OCP5. When SFP 

water is drained to the top of fuel rack, the radiation level is considered too high to 

deploy SFP mitigation strategies on the refuel floor (discussed in section 8.2.3). In 

OCP2, the SFP water takes almost 6 hours to drain to the top of the fuel rack but only 

about 2.5 hours in OCP3. The HEPs were calculated for OCPs 2 and 3. The HEPs for 

OCPs 4 and 5 were not calculated because SFP decay heat is insufficient to cause fuel 

damage, as noted above. 

The SFPS shows that the SFP status is stable in the four scenario classes listed above 

following termination of the computer simulations at 72 hours after the initiating event. This 

result implies that, for the unmitigated scenarios, if spent fuel damage does not occur within the 

first 72 hours, spent fuel damage would not occur afterward because the decrease in decay 

heat and the long time available for response.

Table 40 The SFPS Simulation Results.

No Leak

(90%)

Small Leak

(5%)

Moderate Leak

(5%)

OCP 1 (0.9%) ~ 0.05%

OCP 2 (2.4%)

OCP 3 (5.0%)

OCP 4 (25.7%)

OCP 5 (66%)

- OCP: Operating Cycle Phase

- Percentages above are percent of the time for corresponding condition.

~ 99.2%

~0.8%
180

Table 40 provides an overview for performing an HRA as follows:

• The green cells represent that either the HEP is negligible or mitigation does not affect 

the end consequence. For the SFP no leakage scenario, the SFPS calculated that SFP 

water would take more than 7 days to boil to the top of the fuel rack. Because of the 

long time available for mitigation, the HEP is negligible. For the scenario in which the 

earthquake occurs during OCPs 4 and 5, the SFP fuel decay heat is insufficient to cause 

a gap release event even without the provision of SFP makeup flow; therefore, mitigation 

does not affect the end consequence. These two scenario classes (i.e., no leakage and 

the occurrence of the earthquake during OCPs 4 and 5) are colored as green cells and 

total about 99.2 percent of the conditional probability. An HRA is not performed for the 

scenarios in the green cells.

• The OCP 1 moderate leakage scenario (i.e., the red cell with a ~0,05% conditional 

probability in Table 40) would result in a gap release regardless of whether mitigation 

has taken place because the current NEI guidance for complying with 

10 CFR 50.54(hh)(2) is insufficient (providing at least 500 gpm of injection flow or 

200 gpm of spray flow40). The flow rates are provided by two flow paths using fire 

hoses. Significantly increasing the mitigation flow rate requires setting up additional fire 

hoses to provide additional flow paths. Because the procedures do not provide 

instructions on when additional flow paths should be established, this study concludes 

that no additional flow path other than the two procedure-instructed flow paths will be 

used for SFP mitigation. Therefore, gap release would occur in the OCP1 moderate 

leak scenarios. This is not because the mitigation flow cannot be deployed in time, but 

is because the flow rate is insufficient for the assumed OCP 1 decay heat load as 

determined by SFPS section 6.3.2.

41 

• The yellow colored cells represent conditions where gap release can be prevented if the 

minimum NEI recommended SFP makeup flow (i.e., 500 gpm of injection or 200 gpm of 

spray) is deployed in time. This HRA focuses on these scenarios for which mitigation 

would prevent gap release. 

8.1.2 Key Factors Affecting Available Time for Mitigation

The SFPS divides the reference plant operation cycle into five OCPs. OCPs 1 and 2 occur

during refueling in which the SFP and reactor cavity are hydraulically connected. Because the 

reactor cavity and SFP are located within the same reactor building and they are hydraulically 

connected, a reactor problem would affect the refueling floor work environment in which the 

effective mitigative actions to prevent SFP fuel damage are performed. OCPs 3, 4, and 5 occur

during at-power operations in which the SFP and reactor cavity are hydraulically disconnected. 

 40 NEI 06-12, “B.5.b Phase 2 and 3 Submittal Guideline,” issued December 2006 (ADAMS Accession 

Nos. ML070090060 and ML070080351) recommends minimum of 500 gpm of injection and 200 gpm of 

spray for implementation of the requirements in 10 CFR 50.54(hh)(2). 41 In comparison with OCP 1, moderate leakage, and mitigated scenarios, the OCP 2 scenario has the same 

makeup type (i.e., injection), makeup flow rate, and makeup deployment time. However, gap release did not 

occur in the OCP 2 scenario because the hottest 88 assemblies for OCP 1 at approximately 4 days have a 

decay heat of 1,927 kilowatts (kW) or 65 percent of the whole SFP (2,951 kW), whereas the hottest 

assemblies for OCP 2 at 13 days have a decay heat of 1,143 kW or 32 percent of the whole SFP 

(3,567 kW).
181

This HRA assumes different rates of spent fuel decay heat for each OCP, which in turn affects 

the required mitigation flow and, to some degree, the available time for mitigation necessary to 

prevent SFP damage. 

The SFPS groups the SFP damage caused by the earthquake into three classes: (1) no 

leakage, (2) small leakage, and (3) moderate leakage with a corresponding conditional 

probability of 90 percent, 5 percent, and 5 percent, respectively. The small leakage scenario is 

represented by 40 small tears in the stainless steel liner at the backup bar locations. The small 

cracks create an initial leakage rate of about 250 gpm. The leakage flow rate depends on the 

SFP water level. As the SFP water level decreases, the leakage rate reduces. The moderate 

leakage is represented by a long crack with a combination of the stainless steel SFP liner tear 

and a through-wall concrete crack at the bottom of the SFP wall. Section 4.1.5 of this report 

discusses the SFPS damage states in detail. The moderate leak creates an initial leakage rate 

of about 1,900 gpm. 

The HRA assumes that the SFP leak rate affects the available time necessary for mitigation 

because, when the SFP fuel is not covered by water, the radiation level at the locations in which 

mitigative equipment is stored and mitigative actions are performed is assumed to be too high 

for performance of the mitigative actions in this study. Thus, the SFP leak rate directly affects 

the SFP fuel uncovery time. Table 41 shows the time to SFP fuel uncovery in the various 

scenarios. 

Table 41 Approximate Time of Fuel Uncovery

Time No Leak Small Leak Moderate 

Leak

OCPs 1 and 2 > 7 days 40 hours 6 hours

OCPs 3, 4, and 5 > 7 days 19 hours 2.5 hours

Figure 98 shows the approximate dose rate contours in the refueling area at the time of

defueling when the SFP water level is at the top of the fuel rack. The radiation at the mitigation 

equipment storage location ranges from 3–30 rem per hour and the radiation level at the 

locations of the spray nozzles for SFP makeup is in the range of 10 to 300 rem per hour. 

Working at this radiation level could cause emergency responders who perform mitigation 

actions to receive doses greater than those in EPA’s PAGs (EPA, 1992). This radiation map is 

the basis for specifying that the SFP makeup must be deployed before the SFP water level 

reaches the top of the fuel rack in order to credit mitigation success. 

In addition to radiation, high temperature on the refueling floor is another factor that affects 

mitigation success. In this study, 140 °F (60 °C) is used as the temperature threshold. The 

refueling floor reaches 140 °F before the SFP water level is drained to the top of fuel rack only 

in the OCP 1 and 2 small leak scenarios. In these scenarios, the reactor head is open. Boiling 

in the reactor cavity significantly increases the temperature on the refueling floor. Figure 99

shows the time history of the refueling floor temperature of the OCP 1 small leak scenarios. 

The temperature reaches 140 °F in about 13.5 hours. Figure 100 shows the time history of the 

refueling floor temperature of the OCP 2 small leak scenarios. The temperature reaches 140 °F 

in about 26 hours. Because of the long available response time and steep temperature 

increase at the time of 140 °F reached, changing the temperature threshold to a higher 

temperature does not affect the HRA results.
182

Figure 98 Approximate dose rate of elevation contours, water at the top of fuel hardware, 

around the time of defueling (rem per hour).

Figure 99 The refueling floor temperature of OCP1 small leak scenarios.
183

Figure 100 The refueling floor temperature of OCP 2 small leak scenarios.

In summary, successful deployment of the mitigation strategy has to be done before the earliest 

of either the SFP water reaching the top of the fuel rack or the reactor building atmosphere 

reaching 140 °F. Table 42 shows these available times for the scenarios of interest.

Table 42 The available time*

Small Leak (hr) Moderate Leak (hr)

OCP 1 13.5** 6***

OCP 2 26** 6***

OCP 3 19*** 2.5***

*

Assume Unit 3 reactor is not damaged

**Due to refueling floor temperature reaching 140°F ***Due to SFP water level draining to the top of fuel rack

8.2 Staffing, Mitigation Equipment, Strategies, and Procedures

8.2.1 Staffing, Procedures, Training, and Response Time

Staffing

This HRA assumes that sufficient plant staff is available for Unit 3 SFP mitigation. In the 

situation that the hypothetical earthquake causes damage to multiple SSCs, additional events 

(e.g., fire), and personnel injury, the assumption may not be applicable to some scenarios. 

The reference plant uses a combined main control room for its two reactor units. Consistent 

with NEI 12-01, the on-shift personnel are assumed to be limited to the minimum complement 
184

allowed by the site emergency plan. This represents a staffing level during backshift, weekend 

or holiday. The staffing level of the reference plant, Units 2 and 3, during the backshift, 

weekend, and holiday includes the following: 

• Main Control Room

− One Shift Manager (Licensed Senior Reactor Operator (SRO)). The shift 

manager oversees the control room activities and assesses the emergency 

action level.

− One Shift Technical Advisor (Licensed SRO). The shift technical advisor 

performs independent plant status assessment.

− Two Control Room Supervisors (Licensed SROs). The control room supervisors 

implement procedures as a team with the reactor operators (ROs).

− Two Licensed ROs. The ROs perform control board actions according to the 

control room supervisors’ instructions and answer emergency phone calls.

− Two Assistant (or Spare) Licensed ROs. The assistant operators perform the 

same functions as the ROs.

• On Site

− One Field Supervisor (Licensed SRO). The field supervisor oversees onsite 

activities.

− Nine Auxiliary Operators. The auxiliary operators will report to the main control 

room after the earthquake to obtain the master keys for the assigned tasks. Five 

of the nine auxiliary operators are on the fire brigade.

− Additional Staff. Additional staffing comprises health physicists, chemical staff, 

maintenance personnel, and security staff onsite who can support mitigation

(e.g., health physicists will provide refueling floor radiation information). 

However, these personnel are not expected to directly perform SFP mitigative 

actions. 

The above summary describes a typical staffing level during backshift, weekend or holiday of 

the reference plant instead of the minimum staffing requirement, or during a normal weekday or 

refueling. If the earthquake occurs during normal working hours or if either Unit 2 or 3 is in a 

refueling outage, the staffing level would be significantly higher. 

To augment staffing, except calling for the off-site plant staff (e.g., to mobilize emergency 

response facilities), the reference plant can also call for the nearby Delta-Cardiff Volunteer Fire 

Company to assist in tasks such as SFP mitigation, fire mitigation, and treatment of injured 

personnel. The fire company could send engines, tankers, a ladder fire truck, an air unit, an 

ambulance and personnel to the reference plant site. Based upon the above assumptions, this 

analysis assumes that there is sufficient staff for Unit 3 SFP mitigation. No detailed analysis is 

performed on the staffing situation for all scenarios.
185

Procedures and Operator Initial Responses

In the hypothetical earthquake that causes a station blackout (SBO), the general response is 

that the control room supervisors work with the ROs to implement the emergency response 

procedures. In this case, the entry conditions of the following three procedures are met: 

(1) SE-11, “Loss of Offsite Power”

(2) SE-5, “Earthquake”

(3) TSG-4.1, “Operational Contingency Guideline”

The control room supervisors work with ROs to implement the above three procedures in 

parallel. The immediate objectives are to ensure that the reactor is properly tripped and ensure 

sufficient electricity, equipment, and water to maintain reactor cooling. Because a high-priority 

task in an SBO scenario is the provision of emergency electric power, the control room 

supervisors would send two auxiliary operators to inspect the emergency diesel generators and 

would direct one assistant (or auxiliary) operator to implement SE-11 to connect the dedicated 

power supply from the Conowingo Hydroelectric Generating Station (Conowingo) to the 

reference plant. If the earthquake has not affected Conowingo, connecting its supply power to 

the reference plant would take about 1 hour during normal conditions. The other auxiliary 

operators will be tasked with performing a plant walkdown and SFP inspection in accordance 

with SE-5.

Training

Training related to the implementation of TSG-4.1 and 10 CFR 50.54(hh)(2) includes the 

following:

• annual training in emergency response organization mobilization and implementation of 

the TSG-4.1 and TSG-4.2, “Extreme Damage Mitigation Guidelines for Loss of Large 

Area of the Plant,” procedures and the related requirements in 10 CFR 50.54(hh)(2)

• biannual training on security threat responses

• initial training on procedures and equipment related to 10 CFR 50.54(hh)(2)

Response Time

NEI 06-12, Revision 2, “B.5.b Phase 2 & 3 Submittal Guidance,” states that plants should be 

able to deploy a flexible means of providing SFP makeup (i.e., either 500 gpm of injection or 

200 gpm of spray per unit) within 2 hours from the time in which plant personnel diagnose that 

external SFP makeup is required. This HRA study uses the 2-hour deployment time as the 

action time for deploying mitigation. The total mitigation time is the sum of delay time, diagnosis 

time, and action time (discussed in Section 8.3.2.2).

The analysis in Volume 1 of NUREG/CR-7110, “State-of-the-Art Reactor Consequence 

Analyses Project,” estimates that, given the hypothetical earthquake event that causes SBO but 

with dc power, the technical support center (TSC) is assumed operational within 2.25 hours 

after the earthquake. The NEI 12-01 guideline assumes the following site accessibility: (1) no 

site access within the first 6 hours; (2) limited site access between 6 to 24 hours; and (3) 

improved site access after 24 hours. The assumptions apply to a large-scale external event that 

occurs that results in: (1) all on-site units affected; (2) extended loss of AC power, and (3) 
186

impeded access to the reactor buildings. The emergency response facilities most relevant to 

responding to the hypothetical earthquake are the operational support center (OSC), which is an 

onsite assembly area separate from the control room, and the TSC to which licensee operations 

support personnel report in an emergency. NUREG/CR-7110 does not provide an estimated 

time in which the OSC will be operational. Therefore, for the purposes of this study the TSC 

assumption of 2.25-hours is also used for the OSC when neither Unit 2 nor Unit 3 is in a 

refueling outage. The OSC provides additional man power to mitigate plant damage, but this 

additional staff is not considered in this HRA study.

8.2.2 Mitigation Equipment

This HRA study assumes that portable mitigation equipment is available but the installed 

equipment is not available for Unit 3 SFP mitigation. The portable equipment includes the two 

portable diesel pumps discussed in this section. The installed equipment includes the fire 

system and residual heat removal system. If the earthquake causes damage to multiple 

reactors and SFPs that consequently requires mitigation equipment, there may not be sufficient 

portable equipment for the Unit 3 SFP mitigation. For the purposes of this study, portable 

mitigation equipment was assumed to be available. 

The reference plant relies on the following onsite equipment and systems for SFP makeup:

• Fire System: One motor-driven fire pump and one diesel-driven fire pump are necessary 

to pressurize the fire header. The diesel-driven fire pump is designed to operate for 

6.4 hours at full load. Making up the diesel fuel requires the use of a temporary 120-volt

ac power source to restore a fuel oil transfer pump to deliver fuel for the diesel-driven fire 

pump. In situations in which both fire pumps are lost and cannot be repaired within 

1 hour, the reference plant will contact the York County 911 center for a fire engine to 

pressurize the onsite fire header. If the reference plant cannot obtain offsite support and 

if the situation allows, the reference plant can use one of the two portable diesel pumps 

to pressurize the fire header. 

The two fire pumps are housed in a seismic Class I tornado-resistant structure. 

Therefore, the diesel-driven fire pump is assumed functional after the earthquake. 

However, the underground fire pipes may be damaged by the earthquake. Depending 

on the damage, the fire system may still be available by isolating the damaged section

or sections or by using a fire hose in place of the fire main. The fire system is the 

preferred water source for the most effective mitigation strategies necessary to prevent 

spent fuel gap release. If the fire system is not available, the Conowingo pond or torus 

storage tank is the alternative water source.

• Diesel-Driven Portable (DDP) Pump. The diesel-driven portable pump has the 

capability of delivering 600 gpm of water. A trailer stationed near the pump stores all 

piping, connectors, and spray nozzle. A dedicated pickup truck will be used to tow the 

pump and trailer to the specified location for operation. With a full tank, which is the 

normal condition, the pump can continue to run for more than 12 hours. The DDP pump 

has a 4” discharge connection. To deliver the flow rate of 500 gpm of injection or 250

gpm of spray the plant staff uses a wye adapter to connect the 4” discharge to two 2.5” 

hoses. The reference plant demonstrated that the combination flow rate met the 500 

gpm of injection and 200 gpm of spray requirements. 
187

• Diesel-Driven High-Capacity Portable (DDHCP) Pump. The diesel-driven high-capacity 

portable pump has the capability of delivering 1,300 gpm of water. The DDHCP pump 

has two 4” discharge connections. To deliver the NEI recommended flow rate, the plant 

staff uses a wye adapter to connect a 4” discharge to two 2.5” hoses. Four 2.5” 

discharging hoses would be needed to reach the pump maximum discharge capacity. 

The TSG-4.1 instructs the plant staff to connect two 2.5” hoses to a 4” discharge 

connection for SFP makeup. The reference plant demonstrated that the combination 

flow rate of using two hoses exceeds the 500 gpm of injection and 200 gpm of spray 

requirements. Like the diesel-driven portable pump, dedicated pickup trucks will be used 

to tow the portable pump to the designated location for operation.

The HRA team identified the three systems listed above during a site visit to the reference plant 

in July 2012. The HRA team was aware that the reference plant planned to purchase more 

equipment to address Order EA-12-49 mitigating strategies; however, this HRA study does not 

credit the additional equipment.

The reference plant stores much of its mitigation equipment at grade level. Section 2.4.3.5 of 

reference plant’s’ FSAR discusses the effect of a simultaneous failure of the upstream Holtwood 

dam on the site. The FSAR indicates that the upstream Holtwood dam failure would not 

increase the level of the Conowingo pond such that it would exceed the grade level at the site. 

Therefore, a simultaneous Holtwood dam failure is not assumed to affect the availability of 

mitigation equipment.

8.2.3 Mitigation Strategies

NEI 06-12 discusses implementation strategies for SFP makeup and spray. The mitigation 

strategy described in NEI-06-12 provides that it should be implemented within two hours after 

the decision of deploying the mitigation strategy. This NEI guidance defines the actions that 

should be taken in situations in which normal procedures or command and control structures 

are not available. The notes in the parentheses below include items not considered applicable 

to the accident scenarios for the SFPS. The assumptions in the guidance include the NEI-06-

12:

• An immediate threat warning does not occur.

• Access to the control room is lost (not expected in SFPS scenarios).

• Equipment or supplies normally located in the control room or in the building that houses 

the control room are lost (not expected in SFPS scenarios).

• Access to the building that contains the control room is lost (not expected in SFPS 

scenarios).

• All personnel normally in the control room are lost (not expected in SFPS scenarios).

• All ac and dc power required for operation of plant systems is lost (i.e., both class 1E 

and non-class 1E sources).

• Only minimum site staffing levels are available (i.e., weekend/backshift). Note: the 

minimum staff mentioned in the NEI guidance is not the minimum staff requirements. 
188

Instead, it refers to the normal staffing level during weekend or backshift. This 

assumption does not apply when either Unit 2 or Unit 3 is in refueling outage.

• Other onsite control rooms and personnel in separated building are unaffected. 

(Personnel injury is likely to occur given the hypothetical earthquake.)

• Operations personnel who are not normally located in the control building are available 

for implementation of extensive damage mitigation guidelines.

• Nonlicensed personnel, typically an auxiliary operator, can perform the mitigative

actions.

• The level of training on implementing procedures and guidance is consistent with actions 

under severe accident management guidelines and is consistent with utility 

commitments made under B.5.b Phase 1.

• Before the event, the plant systems are in a normal configuration with the reactor at 

100-percent power. (This SFP safety analysis includes refueling outages (i.e., OCPs 1 

and 2).)

The above items that are noted in parentheses as not expected in the SFPS scenarios apply to 

TSG-4.2. TSG-4.2 may not apply to the SFPS scenarios. Instead, TSG-4.1 is the most 

applicable procedure for the SFPS scenarios. The sections below discuss the SFP mitigation 

strategies in accordance with TSG-4.1. 

Internal Makeup

This strategy connects two fire hoses to the two existing fire system standpipes on the refueling 

floor to provide a minimum of 500-gpm total injection flow to the SFP. The fire system must be 

pressurized to implement this strategy. To implement this strategy, the operators need to 

remove the existing 1.5-in reducer from the two fire standpipes, connect two 2.5-in fire hoses to 

the two standpipes, and route the two fire hoses to the SFP. Operators can deliver makeup flow 

by fastening the hose to the SFP side for direct injection into the SFP. Operators can also 

deliver makeup flow by connecting the two fire hoses to the two spray nozzles to spray water 

into the SFP. This strategy will deliver a total spray flow of more than 200 gpm. All equipment 

mentioned is available on the refueling floor. This strategy assumes that the refueling floor is 

accessible for local makeup. 

External Makeup and Spray

This strategy uses any of the two portable diesel pumps (Section 8.2.2) to inject or spray water 

into the SFP. This strategy requires the plant staff to (1) tow the portable diesel pump to the 

desired location at grade level, (2) lay two approximately 200-ft fire hoses that are connected by 

two sections from the refueling floor through a stairwell to the grade level (about 100 ft in 

elevation difference) to connect the hoses to the charging output of the portable pump, (3) 

connect the hose end on the refueling floor to an spray nozzle, and (4) connect the portable 

pump’s suction to a fire hydrant. The hoses for connecting the pump discharge to the spray 

nozzles are stored on the refueling floor. Each spray nozzle can be adjusted for spray or to 

obtain full flow (i.e., injection). 
189

The external makeup and spray mitigation strategy uses the fire water system as the default 

water source to the portable diesel pumps. Under situations in which the fire water system is 

not available, the reference plant’s procedure SO-37L.1.a, “Diesel Driven Portable and Diesel 

Driven High Capacity Portable Pump Startup and Shutdown,” identifies additional water 

sources, including the inner pond, discharge pond, and Conowingo pond. In the event of a 

seismically induced failure of the Conowingo pond, the loss-of-pond procedure provides cooling 

water management strategies. In addition, the reference plant Assignment Report 

No. 01001590 identifies the candidate water sources, including high-pressure service water, fire 

water, residual heat removal water, condensate transfer water, and cross connections to the 

opposing unit’s spent fuel water supply. Detailed step-by-step instructions for using water from 

the alternative water sources are not available. However, this HRA study credits the use of the 

alternative water sources in the situations when the fire water is not available because of the 

similarity of using water from these sources to drafting fire water.

External Local Spray or Scrub 

This strategy uses any of the following three procedures, individually or in combination, to 

provide spent fuel cooling or secondary containment spray to scrub potential radionuclides

released from the SFP primary or secondary containment structures:

(1) Use the portable diesel pump to provide water to the two spray nozzles on the refueling 

floor to spray water into the SFP. This strategy requires operators to lay out fire hoses 

from the refueling floor to grade level, as described in the section above entitled, 

“External Makeup and Spray.” 

(2) Use the portable diesel pump to provide water to one or two of the two spray nozzles

located on the turbine building roof to spray water to the secondary containment or the 

refueling floor through building breaches. This strategy requires operators to lay out 

hoses from the turbine building roof to grade level to connect the portable pump and the 

spray nozzle.

(3) Use a ladder truck to spray water into the SFP area through building breaches from the 

steel structure surrounding the SFP floor. This strategy requires the use of an offsite fire 

company’s 100-ft ladder fire truck. Exelon Generation Company, LLC (owner of the 

reference plant), has a letter of agreement with nearby Delta-Cardiff Volunteer Fire 

Company. Upon dispatch and without additional complexity, the fire truck could arrive at 

the reference plant within 30 minutes (Assignment Report No. 01001590). The 

Delta-Cardiff Volunteer Fire Company possesses two fire trucks from Seagrave Fire 

Apparatus, LLC, either one of which can perform the portable diesel pump function. 

Other nearby fire companies could support the reference plant mitigative efforts. 

Items (2) and (3) above are useful primarily to mitigate the release of radioactivity off-site, but 

are not effective in preventing radioactive release from the SFP fuel rods. Because the 

mitigation success of this HRA study is to prevent radioactive release from the SFP fuel rods, 

items (2) and (3) are not credited in this HRA study. TSG-4.1 requires the plant to use the 

external local spray or scrub strategy after it has attempted the internal makeup and external 

makeup and spray strategies (as discussed earlier). For consistency with the NEI guidance, the 

reference plant uses the flowchart shown in Figure 101 as general guidance for deployment of 

SFP mitigation strategies.
190

Makeup with Residual Heat Removal Pump from the Torus

This strategy requires that electrical power is available for a residual heat removal pump to 

pump torus water into the SFP at a flow rate of 10,000 gpm. This flow rate is much larger than 

the maximum moderate leakage flow rate (i.e., approximately 1,900 gpm). In an SBO scenario, 

this strategy is not available because power is not available for the residual heat removal 

pumps. this HRA study assumes this strategy is not available.

Leakage Control

The reference plant has a list of stocked materials that could help to reduce the leakage flow 

rate, including steel plates, plywood, bag stopper, sealants, ropes, and rubber matting. Certain 

materials would require a crane for moving (e.g., 5/8-in by 4-ft by 4-ft steel plates. Based on its 

emphasis on the initiation of makeup strategies and the 72-hr scope of the analysis, the study 

did not consider repair options.

Indication of Damage to SFP 

Resulting from Large Leak1

SFP Area 

Accessible2

Makeup Using All 

Available Means

Local Spray 

Possible

Deploy 

External Spray

Deploy 

Spray Locally

SFP Leakage 

Excessive3

Makeup Using Internal 

Strategy, If Available

Makeup Using External 

Strategy, If Necessary

SFP Level Still 

Dropping

Local Spray 

Possible

Deploy External Spray

Refer to Leakage 

Control Strategies

Deploy

Spray Locally

Yes Yes Yes

Yes

Yes

No

No

No

No

Notes:

1. A large leak is one that 

exceeds available, normal 

makeup or one in which the 

effectiveness of normal makeup 

cannot be determined.

2. Accessibility can be affected 

by damage conditions or local 

dose rates.

3. Excessive leakage is an SFP 

leak rate estimated to be 

greater than 500 gpm, or high 

dose rates indicate excessive 

loss of inventory.

No

Figure 101 Generalized guidance for SFP makeup and spray decisions

8.3 Study Framework, Scope, and Approach

8.3.1 Study Framework and Scope

Preventing gap release of the Unit 3 SFP fuel is the success criterion defined in this HRA study. 

Because human performance is sensitive to the extent of earthquake damage to the plant, the 

study identifies a set of plant damage states and estimates an HEP for each damage state. 

Identification of the damage state is based on the status of a few key SSCs that include electric 
191

power availability, reactor status, and fire system availability. Figure 102 illustrates the 

framework and scope of this HRA study. The large rectangle with dashed lines shown in Figure 

102 represents the scope of this HRA study. Each box within the dashed lines represents a 

probability. The partial enclosure of the damage state signifies that the scope of this HRA study 

only identifies the plant damage states; it does not estimate the probabilities of the damage 

states. Therefore, the gap release (and no gap release) probabilities could not be estimated, 

and the HEPs were computed to only provide initial general insights.

Figure 102 The study framework

8.3.2 Approach to Human Error Probability Estimates

8.3.2.1 A Two-Phased Approach to HEP Estimates

This HRA study used a two-phase approach to estimate mitigation failure probabilities or human 

error probabilities (HEPs). Phase 1 estimates HEPs for mitigating the reference plant SFP leak. 

These estimates consider the status of the electric power, reactor and primary containment, and 

fire system of Unit 3, the Unit 3 OCP, and Unit 3 SFP leakage rate. These are the dominant 

factors that affect mitigation of a single SFP leak. In Phase 2, other damage to the site that 

affects mitigation is discussed. Phase 2 involves situations that combine reactor and SFP 

problems or multiple unit problems caused by the same earthquake. The four discrete steps 

listed below represent the HEP estimation process used in this analysis. The four sections that 

follow discuss these steps in detail.

(1) Identify the time required for deployment of SFP makeup.

(2) Identify the damage states and corresponding available time.

(3) Estimate the HEP of each damage state.

(4) Identify additional feasibility considerations.
192

8.3.2.2 Step 1—Identify the Time Required

In this study, the total time necessary for deployment of the SFP makeup is the sum of the 

following three time segments:

(1) Delay Time: In an earthquake-induced SBO scenario, the control room operator’s 

primary focus is on reactor safety. Although the SFP trouble alarm is triggered soon 

after the earthquake, a time delay occurs for starting a diagnosis process to investigate 

an SFP problem. The cue for starting to investigate the SFP is the earthquake 

procedure SE-5. Step 9 of the procedure instructs the operators to check the SFP, SFP 

cooling system, and fuel floor blowout panels. Based on an interview with PBAPS staff, 

the delay time ranges between 30 minutes and 1 hour. This study uses 45 minutes for 

the SBO scenarios, 30 minutes for LOOP scenarios, and 60 minutes for SBO without dc 

power scenarios because, based on crew interviews, the control room supervisor would,

at a minimum, simultaneously implement SE-11 and SE-5. When less electricity is 

available for maintaining reactor safety margin, the operators would put more effort into 

restoring electricity (i.e., SE-11). As a result, less time is spent on SE-5, which 

consequently would delay implementation of Step 9 in SE-5 to send auxiliary operators 

to check the SFP status. 

(2) Diagnosis Time: Diagnosis time is the time between when auxiliary operators are 

deployed to inspect the SFP and when they report SFP leakage back to the control room 

operators. Based on the leakage rate (both small leakage and moderate leakage) and 

leakage locations (i.e., the SFP bottom at the elevation of a few inches above the 195-ft 

floor), detecting SFP leakage is not a challenging task. Based on the same interview 

with PBAPS staff and a plant walkdown of the path that the auxiliary operators would 

normally take to inspect the SFP, the diagnosis time was determined to be 15 minutes.

(3) Action Time: The 2-hr implementation expectation in NEI 06-12 is used for deployment 

of the portable diesel pump to provide SFP makeup. The HRA uses the 2 hours as the 

action time at which the fire system is available because TSG-4.1 instructs the staff on 

how to use the fire system as the water source. When the fire system is not available, 

using water from the alternative water sources would require additional time. An 

additional 1 hour of action time is necessary when the fire system is not available. 

Table 43 and Table 44 summarize the time estimates based on the above discussion. Table 43

shows the mitigation time estimates in the scenarios for which fire water is sufficient for 

mitigation. Table 44 adds 1 hour of action time to Table 43 to account for the effect of 

unavailable or insufficient fire water. In Table 43 and Table 44, note that the total time 

difference between the LOOP scenarios and SBO without dc scenarios is only 30 minutes. 

However, the conditional reactor core damage probability in these two scenarios would be 

significantly different. Further, the conditional reactor core damage probability in these two 

scenarios would have a significant difference. That difference directly affects the refueling floor 

accessibility and, in turn, the mitigation success probability. This HRA study assesses HEPs for 

the SBO and SBO without dc scenarios with and without reactor core damage separately.
193

Table 43 Estimates of the Time Required for the Operator to Deploy SFP Makeup If Fire 

Water Is Available

Delay Time Diagnosis Time Action Time Total Time

Required

LOOP 30 minutes 15 minutes 2 hours 2 hours

45 minutes

SBO 45 minutes 15 minutes 2 hours 3 hours

SBO without 

dc

60 minutes 15 minutes 2 hours 3 hours

15 minutes

Table 44 Estimates of the Time Required for the Operator to Deploy SFP Makeup

If Fire Water Is Not Available or If It Cannot Deliver Sufficient Flow

Delay Time Diagnosis Time Action Time Total Time

Required

LOOP 30 minutes 15 minutes 3 hours 3 hours

45 minutes

SBO 45 minutes 15 minutes 3 hours 4 hours

SBO without 

dc

60 minutes 15 minutes 3 hours 4 hours

15 minutes

8.3.2.3 Step 2—Identify the Damage States and Available Time

The key factors that affect the likelihood of successful mitigation of the Unit 3 SFP include SFP 

leakage size; OCP; and the status of the electric power, reactor and primary containment, and 

fire system of Unit 3. These factors characterize the damage states (as shown in Table 47). 

SFP leakage size and whether the SFP and the reactor cavity are hydraulically connected 

(i.e., during refueling and nonrefueling) largely determine available time. As discussed earlier, 

the available time is determined by the shorter time of either the SFP water reaching the top of 

the fuel rack or the refueling floor reaching 140°F. Table 45 shows the time required and time 

available of the damage states of interest assuming the Unit 3 reactor is not damaged.

Table 45 Estimates of time required and time available for mitigation

Small Leak Moderate Leak

Time 

Required(hr)

Time 

Available(hr)

Time 

Required(hr)

Time 

Available(hr)

OCP 1

LOOP 2.75(3.75)

13.5

2.75(3.75)

SBO 3.0(4.0) 3.0(4.0) 6

SBO w/o DC 3.25(4.25) 3.25(4.25)

OCP 2

LOOP 2.75(3.75)

26

2.75(3.75)

SBO 3.0(4.0) 3.0(4.0) 6

SBO w/o DC 3.25(4.25) 3.25(4.25)

OCP 3

LOOP 2.75(3.75)

19

2.75(3.75)

SBO 3.0(4.0) 3.0(4.0) 2.5

SBO w/o DC 3.25(4.25) 3.25(4.25)
194

*The numbers outside the parentheses are the time required when the fire system is available. The numbers inside 

the parentheses are the time required when the fire system is not available.

**These values assume that the Unit 3 reactor is not damaged and the staff uses the portable diesel driven pumps for 

SFP mitigation

8.3.2.4 Step 3—Estimate Basic HEPs of a Single Unit Event

This step estimates the basic HEPs for each damage state based on the following assumptions 

and practices:

• The required mitigative equipment stored outside of the reference plant, Unit 3, and 

water sources are available. Step 4 considers equipment and water unavailability and 

other factors. 

• The plant staff is available for performing the mitigation activities. 

• The earthquake damaged much of the nonsafety piping and equipment but the 

workplace is accessible (with additional difficulties from normal situations).

• The purpose of including some situations in the HRA (e.g., core damage within the 

specified available time) is to explicitly identify the key factors that affect human 

performance. Estimating the likelihood of the occurrence of these situations is outside 

the scope of this HRA study. Estimating the likelihood of each situation would require a 

PRA.

The main considerations necessary for assessing HEPs are based on the time margin and 

workload that affect staffing availability. Electric power availability strongly affects workload. 

The power availability is classified into: (1) LOOP only, (2) SBO, and (3) SBO without dc. The 

three classes of power availability impose significant differences in operator workload that, in 

turn, affect personnel availability to perform all required tasks. The flow diagram in Figure 103

shows the HEP estimation procedure, which is based on NUREG-6883 “The SPAR-H Human 

Reliability Analysis Method” issued in 2005, supplemented with the NRC staff’s expert 

judgment. 

SPAR-H’s low power and shut down diagnosis worksheets classify time margin effects into five 

classes as shown in Table 46.

Table 46 Time margin effects on human error probability in the SPAR-H HRA method for 

cognitive activities in low power /shutdown operations 

Class HEP or HEP Multiplier Note

Insufficient time HEP = 1.0 Less than 2/3 of normally time required

Barely adequate 

time

HEP multiplier = 10 ~2/3 of normally time required

Nominal time HEP multiplier = 1 About the normally time required

Extra time HEP multiplier = 0.1 Equal to or greater than 5 times of normal time

required

Expansive time HEP multiplier = 0.01 Equal to or greater than 50 times of normal time

required
195

The SPAR-H’s action worksheets use slightly different time scales to adjust the HEP. The

adjusting factor 1 in Figure 103 represent time margin effects on HEP based on SPAR-H’s 

classification. The adjusting factor 2 in Figure 103 represents the performance shaping factors 

of “complexity” and “ergonomics/human machine interface” of SPAR-H. Table 47 shows the 

HEP calculation results. Note that OCP1 moderate leak scenarios are likely to have gap 

release because the NEI recommended minimum mitigation flow rates are insufficient to prevent

gap release. This is not reflected in Table 47 because this is not considered as a human error 

in a typical HRA application.

Table 47 shows that fire system availability in general does not have significant effects on 

human error probability. Table 48 summarizes the qualitative results of the HRA with respect to 

the likelihood of gap release in various plant states with the assumption of no reactor core 

damage. Two plant states have an HEP of 1.0: moderate leak in OCP1 and moderate leak in 

OCP3. In OCP1 moderate leak scenarios, the high likelihood is because the NEI recommended

minimum mitigation flow rates are insufficient to prevent gap release. The high likelihood is not 

shown in Table 47 because the failure is considered as a design issue rather than a human 

error from a conventional HRA perspective. In OCP3 moderate leak scenarios, the high 

likelihood is because of the short time available for response (i.e., about 2.5 hours). 
196

Basic HEP =

1.0E-2

Available time is 

greater than 72 

hours?

HEP is negligible

Yes

No The Available Time/

Required Time is

IF available time < 0.8 THEN HEP = 1.0;

IF 0.8 <= available time < 1.0 THEN AF1 = 5.0;

IF 1.0 <= available time < 1.5 THEN AF1 = 1.0; 

IF 1.5 <= available time < 3.0 THEN AF1 = 0.5;

IF 3.0 <= available time < 7.0 THEN AF1 = 0.1;

IF 7.0 <= available time < 50 THEN AF1 = 0.05;

IF 50 <= available time THEN AF1 = 0.01;

(AF: Adjusting Factor)

Earthquake occurs 

during refueling?

Unit 3 Reactor Core 

Damaged?

IF DC is okay

 THEN AF2 = 2.0;

IF DC failed 

 THEN AF2 = 5.0;

HEP = 

basic HEP × AF1 × AF2;

If HEP > 1.0 THEN HEP = 1.0

Yes

No

If DC is okay

 THEN SF2 = 2.0;

If DC failed THEN

 THEN SF2 = 5.0 

Unit 3 Reactor Core 

Damaged?

This is a LOOP only 

scenario?

HEP = 

basic HEP × AF1

This is a SBO 

scenarios

Determine the time 

margin’s effects.

Earthquake occurs 

during at-power 

operation.

HEP = 1.0

Unit 3 Primary 

Containment Failed?

If DC is available

 THEN AF2 = 10.0;

If DC failed THEN

 THEN AF2 = 50.0 

HEP = 1.0

Yes

Yes Yes

No

No

Yes

No

Determine the task 

complexity effects.

Figure 103 Flow chart for estimating HEPs for a single reactor unit event
197

Table 47 Human error probability estimates of a single unit event

Small Leak Moderate Leak

OCP 1

LOOP* 0.001

(0.001)

0.003

(0.003)

SBO

No CD 0.006

(0.006)

0.002

(0.002)

CD 1

(1)

1

(1)

SBO w/o 

DC

No CD 0.015

(0.015)

0.005

(0.005)

CD 1

(1)

1

(1)

OCP 2

LOOP* No CD 0.003

(0.003)

0.0005

(0.001)

SBO

No CD 0.006

(0.006)

0.001

(0.002)

CD 1

(1)

1

(1)

SBO w/o 

DC

No CD 0.015

(0.015)

0.0025

(0.005)

CD 1

(1)

1

(1)

OCP 3

LOOP* No CD 0.01

(0.1)

0.001

(0.001)

SBO

No CD 0.2

(1)

0.002

(0.002)

CD; CTM 

intact

1

(1)

0.05

(0.05)

CD, CTM 

breach

1

(1)

1

(1)

SBO w/o 

DC

No CD 0.5

(1)

0.005

(0.005)

CD; CTM 

intact

1

(1)

0.05

(0.05)

CD, CTM 

breach

1

(1)

1

(1)

*Assume no reactor core damage (CD)

**The numbers outside the parentheses are the HEPs when the fire system is available. The numbers inside the 

parentheses are the HEPs when the fire system is not available
198

Table 48 The likelihood of gap release*

Small Leak

(5%)**

Moderate Leak

(5%)**

OCP 1 (0.9%)** Low1 High2

OCP 2 (2.4%)** Low1 Low3

OCP 3 (5.0%)** Low1 High4

*Assumes only one SFP damaged without concurrent reactor core damage

**The probabilities are conditional probabilities given that the studied earthquake occurs 1

The available time for response is long. The SFP fuel is submerged if SFP makeup is deployed in time. 2

The NEI recommended minimum mitigation flow rate is insufficient to prevent gap release. 3

The NEI recommended minimum mitigation flow rate is sufficient to prevent gap release. 4

The available time for response is short so that the SFP makeup will likely not be deployed in time to prevent gap 

release.

8.3.2.5 Step 4—Additional Feasibility Considerations

This final step (i.e., Step 4) identifies situations that occur outside of the reference plant, Unit 3, 

that would have adverse effects on Unit 3 SFP mitigation. These effects are not considered in 

Step 3. These additional considerations include the following:

• Equipment demand cannot be met: When the earthquake causes extensive damage to 

the reactors and SFPs of Unit 2 and Unit 3 and the normal reactor and SFP cool down 

mechanisms are not available, the two portable pumps may not be available for Unit 3 

SFP makeup given the multiple demands. The DDHCP pump has two 4” discharge 

connections, and the DDP pump has one 4” discharge connection. In combination, the 

two portable diesel pumps can deliver three times the NEI recommended minimum 

mitigation flow rate. The operators have to decide how to use the limited equipment for 

multiple problems for the reference plant’s two reactors and two SFPs. The decision will 

strongly depend on the situation. 

• Damage to the mitigation equipment (e.g., the DDHCP pump and DDP pump) and 

support equipment (e.g., pump accessories and the designated truck to tow the pumps) 

would reduce the available equipment or delay mitigation.

• Simultaneous large or multiple fire events that demand more plant staff personnel than 

those available.

• Structural damage causes plant personnel injury that could result in less than adequate 

personnel available for SFP mitigation.

• Unit 3 Refueling floor is inaccessible for reasons such as Unit 3 reactor damage causing 

high radiation in the access path.

The cells in Table 48 with low gap release likelihood can be split into two groups:

• Greater than 13 hours for all small leak scenarios

• About 6 hours for OCP 2 moderate leak scenarios
199

In either situation, sufficiency of plant response personnel is likely not an issue because of the 

long available time of the small leakage scenarios and in refueling outage of the OCP 2 

scenarios. Though not accomplished through a full scope PRA, this HRA attempted to account 

for the complexity of handling multiple reactors and SFP damage events. As such, an adjusting 

factor of 50 (based on the SPAR-H’s performance shaping factors of “high complexity” and “low 

experience/training”) was applied to Table 47. The results are summarized in Figure 104 and 

Table 49.

Damage to 

multiple reactors 

and SFPs

Unit 3

Refueling floor 

accessible? 

Yes

No

Equipment 

available for 

Unit 3 SFP 

mitigation? 

Yes

No

OCPs 1, 2,or 3 

and Unit 3 SFP 

leaks

Small leak

No Unit 3 SFP 

leakage or

OCPs 4 or 

OCP 5

Yes

No

Moderate 

Leak?

Gap release probability is:

- 0.05 if LOOP only

- 0.1 if SBO

- 0.25 if SBO w/o DC

OCP 2?

Yes

No

Gap release probability is:

- 0.15 if LOOP only

- 0.3 if SBO

- 0.75 if SBO w/o DC

Gap release 

probability Is 1.0.

Gap release 

probability Is 1.0.

Gap release 

probability Is 1.0.

Gap release 

probability Is 

negligible.

Yes

No

OCP 1 or OCP3

Figure 104 The gap release probability assessments given damage to multiple reactors 

and SFPs.

Table 49 shows three levels of likelihood of having radioactive release from the Unit 3 SFP fuel 

rods. Three colored coded regions are discussed below:

• Green Cells:

Two sub groups in the green coded cells: (1) the “no leak” scenarios have long available

time (greater than 7 days) for response. The mitigation failure probability is determined 

to be negligible; and (2) The OCP4 and OCP5 have low decay heat. Even without 

mitigation, radioactive release is not expected.

• Yellow Cells:

For the small leak scenarios, the available time ranges from more than 13 hours to more 

than 1 day. Give the long time available, time is not a critical factor affecting mitigation 

success. The mitigation failure probability is estimated to range from one failure out of 

twenty to one failure out of four. For the OCP2 moderate leak scenarios, the time 
200

available is 6 hours. This increases the mitigation failure probability compared to the 

small leak scenarios. The mitigation failure probabilities for OCP 2 moderate leak 

scenario range from one failure out of twenty to three failures out of four. 

• Red Cells:

Two red cells are in Table 49. The OCP1 moderate leak scenario is red because the

500 gpm of injection or 200 gpm spray is insufficient to prevent fuel overheating. The 

OCP3 moderate leak scenario has only a short time available (2.5 hours), and mitigation 

is not expected to be deployed in time. 

Table 49 Scenario Specific Human Error Probability Estimates*. 

No Leak

(90%)

Small Leak

(5%)

Moderate Leak

(5%)

OCP 1 (0.9%)

Negligible

- 0.05 if LOOP only

- 0.1 if SBO

- 0.25 if SBO w/o DC

1.0**

OCP 2 (2.4%) - 0.15 if LOOP only

- 0.3 if SBO

- 0.75 if SBO w/o DC

OCP 3 (5.0%) 1.0***

OCP 4 and

OCP 5 (91.7%) Inconsequential

- OCP: Operating Cycle Phase

- Percentages above are the percent of the time for the corresponding condition.

* Assume mitigating equipment is available for Unit 3 SFP, and Unit 3 reactor status does not deny access to the Unit 

3 refueling floor.

**The NEI recommended minimum mitigation flow rate is not sufficient to prevent gap release. The procedure (i.e., 

TSG-4.1) does not instruct operators to establish an additional SFP makeup flow path to significantly increase the 

SFP makeup flow rate to be greater than the minimum flow rate recommended by NEI. The HEP is set to 1.0 to 

indicate that gap release would occur.

***Primarily due to short time available for response (i.e., ~ 2.5 hours). OCPs 1 and 2 (i.e., during refueling) have the 

reactor cavity and SFP hydraulically connected, which provides more time than OCP3. 

8.4 Discussion and Summary

This SFP HRA study identifies a set of plant damage states and calculates the corresponding 

HEPs; however, it does not calculate the conditional probabilities of the damage states. The 

following information summarizes the human performance insights:

• The HEPs of the SFP no leakage scenarios are negligible because of the long time 

available for response. The scenarios in OCPs 4 and 5 would not lead to gap release of 

the SFP fuel because of the low spent fuel decay heat. These two groups of scenarios 

share 99.2 percent of the probability (i.e., 0.992). In other words, given the 0.5–1.0g 

earthquake, the SFPS estimates minimum 99.2-percent conditional probability that a gap 

release would not occur.

• 500 gpm of injection is not sufficient to prevent gap release in the OCP1 moderate leak 

scenarios, as determined in the SFPS. The SFPS did not perform sensitivity 

calculations to determine the NEI recommended flow rates (either injection or spray) to 

prevent gap release in this case. Therefore, this HRA study assumes that the plant staff 

would need to connect more than the procedurarlized two hoses to the portable diesel 

pump and use more than two spray nozzles to provide sufficient cooling. Because TSG4.1 only provides instructions on establishing two hoses and two spray nozzles, the lack 

of procedures and insufficient equipment (i.e., hoses and spray nozzles) are assumed to 
201

cause the mitigation to fail in the OCP1 moderate leak scenarios. Even though the 

reference plant flow rates are greater than the NEI recommended minimum flow rate, 

sensitivity calculations on the actual flow rate from the spray nozzle would be needed for 

a more detailed assessment. 

• The available time for SFP mitigation is determined by the shorter time of either the SFP 

water draining to the top of the fuel or the refueling floor reaching 140°F. In the OCP 1 

and 2 small leakage scenarios, the refueling floor reaches 140°F earlier than the time 

necessary for the SFP water to drain to the top of the fuel rack thus causing refueling 

floor temperature to become the limiting factor for determining HEPs

• The two spray nozzles (as illustrated in TSG-4.1) for SFP makeup are set up in high

radiation areas. Delivering the same amount of flow from a low-dose area (e.g., near the 

wall next to the storage pool) would significantly increase the available time because 

using the time necessary for the SFP water level to reach the top of the fuel rack as a 

criterion is based on the radiation level at the locations of the spray nozzles, as specified 

in TSG-4.1. Moving the spray nozzles setup locations to a lower dose area would 

significantly increase the mitigation success probabilities for moderate leak scenarios for 

which the time necessary for SFP water to drain to the top of the fuel rack is the limiting 

factor.

• The fire system availability (from earthquake-induced fire piping rupture) affects OCP 3 

moderate leakage scenarios but not small leak scenarios because the small leak 

scenarios have at least 13 hours for mitigation deployment. Instructions on how to 

quickly determine whether the fire system can deliver sufficient flow for mitigation may

improve the probability of successful mitigation.

The success criterion of this HRA study is to prevent a radioactive release from the Unit 3 SFP 

fuel rods. The mitigation strategies that emphasize keeping the radioactivity released from the 

fuel rods on site are not within this HRA scope. Deploying these strategies could mitigate 

radioactive releases to the environment. 

The HEP results shown in Table 49 are based on the assumptions that mitigation equipment is 

available, there is no combination of Unit 3 reactor core damage and primary containment 

failure that causes inaccessibility of the refueling floor, and there is sufficient staff to deploy for 

the Unit 3 SFP mitigation. If the earthquake damages multiple reactors and SFPs some of the 

above assumptions may not apply. An analysis of these issues would require the performance 

of a PRA and associated HRA.
202
203

9. CONSIDERATION OF UNCERTAINTY

This section catalogues a set of sensitivity analyses to better understand the potential effect of 

certain assumptions on the results of this study. The sensitivity analyses include those for 

analyzing additional plant states (e.g., 1x8 pattern in a high-density loading configuration) and 

for analyzing parameter/model uncertainties (e.g., hydrogen combustion ignition). The 

assumptions analyzed were chosen from the list of key assumptions compiled in Section 2, 

based on their perceived importance and project constraints.

9.1 Sensitivity to Hydrogen Combustion (MELCOR)

A sensitivity calculation was performed to examine the response of the SFP to the hydrogen 

combustion ignition criterion. This calculation involved reducing the hydrogen concentration 

from 10 percent to 7 percent given the inherent uncertainties in this parameter discussed 

before. The case that showed the strongest sensitivity to this parameter is the unmitigated 

high-density, moderate leak size scenario from OCP 2. The base case reactor building 

concentration of gases in Figure 105 shows that, by the time the hydrogen concentration 

exceeds the ignition criterion of 10 percent, the oxygen concentration is below the 5-percent 

limit and no hydrogen combustion is predicted. However, at about 18 hours, both the hydrogen 

and oxygen concentrations are above 7 percent, which can support a hydrogen combustion. 

Figure 106 shows the mole fraction of gases for this sensitivity case. At about 18 hours, the 

hydrogen combustion consumes the hydrogen in the building as evidenced by the rapid 

decrease in the hydrogen concentration and is accompanied by a sudden increase in the 

oxygen concentration as the failure of the reactor building causes the outside air to enter. 

Following the air ingress, the clad oxidation power significantly increases (compare the base 

case in Figure 107 with the sensitivity case in Figure 108). The higher oxidation power leads to 

higher clad temperatures42 (Figure 109 and Figure 110) and additional release of fission 

products from the fuel and release to the environment (Figure 111 and Figure 112). The cesium 

release fraction of 50 percent for this sensitivity is much higher than the base case of 

1.6 percent (see Table 27), and it is comparable to the release fraction of 49 percent for the 

uniform pattern (see Table 50).

 42 The failure of the fuel rods leads to formation of debris that continues to release fission products.
204

Figure 105 Reactor building mole fraction for unmitigated high-density moderate leak 

(OCP2)

Figure 106 Reactor building mole fraction for unmitigated high-density moderate leak 

(OCP2-S)
205

Figure 107 SFP power for unmitigated high-density moderate leak (OCP2)

Figure 108 SFP power for unmitigated high-density moderate leak (OCP2-S)
206

Figure 109 Ring 1 clad temperature for unmitigated high-density moderate leak (OCP2)

Figure 110 Ring 1 clad temperature for unmitigated high-density moderate leak (OCP2-S)
207

Figure 111 Cesium environmental release fraction for unmitigated high density moderate 

leak (OCP2)

Figure 112 Cesium environmental release fraction for unmitigated high density moderate 

leak (OCP2-S)
208

9.2 Sensitivity to 1x8 Fuel Assembly Pattern (MELCOR)

This sensitivity involves a more favorable fuel pattern in which the hot assemblies are 

surrounded by eight cold assemblies. Figure 113 shows the assembly layout in a 1x8 pattern in 

which the 284 assemblies from the last offload are grouped into Rings 1, 3, and 5 (see Figure 

46 for the 1x4 pattern). Rings 2, 4, and 6 contain all of the old fuel and have a total of 2,771

assemblies with their total decay heat distributed in each ring scaled by the number of 

assemblies.

Figure 113 Layout of assemblies for OCP3 high density (1x8) model

A number of sensitivity calculations were performed for the high density, small leak scenarios in 

OCP2 and OCP3 (which had the highest release as a result of hydrogen combustion). Figure 

114 and Figure 115 show the thermal response of the fuel in the 1x8 configuration for OCP3. 

Figure 114 shows that the highest power fuel assemblies in Ring 1 do not undergo a zirconium

fire and the temperatures remain low enough to avoid gap release for the duration of the 

transient. The midplane fuel temperatures in the pool shown in Figure 115 have a more uniform 

heat up of the fuel assemblies than the comparable 1x4 pattern. There is more mass of the cold 

assemblies in the 1x8 pattern, which leads to lower heatup of the fuel. The fuel thermal 

response in the 1x8 pattern can be contrasted to the 1x4 pattern as shown in Figure 116 and 

Figure 117. For the 1x8 calculation, no release occurs from the fuel through 72 hours. In the 

1x4 layout, a zirconium fire propagation began at 40 hours, which led to a 42-percent release of 
209

cesium inventory to the environment. For the OCP2 configuration results shown in Figure 118

and Figure 119, the decay heat is high enough to cause a zirconium fire in the hottest 

assemblies, even though the peak fuel temperatures in the 1x8 pattern are somewhat lower. 

The beneficial effect of the 1x8 pattern is also evidenced by the lower release fractions, as 

shown in Figure 120 and Figure 121.

Figure 114 Ring 1 clad temperature for unmitigated high-density small leak (OCP3; 1x8) 
210

Figure 115 Midplane clad temperature for unmitigated high-density small leak (OCP3; 

1x8)

Figure 116 Ring 1 clad temperature for unmitigated high-density small leak (OCP3; 1x4) 
211

Figure 117 Midplane clad temperature for unmitigated high-density small leak (OCP3; 

1x4) 

Figure 118 Ring 1 clad temperature for unmitigated high-density small leak (OCP2; 1x8) 
212

Figure 119 Ring 1 clad temperature for unmitigated high-density small leak (OCP2; 1x4) 

Figure 120 Environmental release fractions for unmitigated high-density small leak 

(OCP2; 1x8) 
213

Figure 121 Environmental release fractions for unmitigated high-density small leak 

(OCP2; 1x4) 

9.3 Sensitivity to a Contiguous (Uniform) Fuel Pattern during an Outage

(MELCOR/MACCS2)

The reference plant studied has prearranged the SFP such that discharged assemblies can be 

placed directly into a 1x4 (actually 1x8 in the case of PBAPS) arrangement for the last two 

outages for both operating units. This approach is consistent with the requirements previously 

discussed in Section 5.1. However, those requirements do allow for the fuel to be stored in a

less favorable configuration for some time following discharge if other considerations prevent 

prearrangement. A requirement is associated with the time window by which the 

1x4 arrangement must be achieved; however, the specific time requirement is not publicly

available information (because it could be potentially useful to an adversary). This section 

posits a situation in which the fuel is unfavorably arranged during the outage to demonstrate the 

effect of this aspect on the results.

Figure 122 and Figure 123 show the layout of assemblies for the OCP1 and OCP2 uniform 

configuration. For the 1x4 pattern (see Figure 44), the effective area between Rings 1 and 2 

was determined by the number of panels (i.e., 352 panels for 88 assemblies), since each 

assembly in Ring 1 is completely surrounded by Ring 2 assemblies. In the uniform pattern

(Figure 122), the surface areas between Rings 1 and 2 and between Rings 3 and 4 were 

effectively reduced by about an order of magnitude, assuming that all of the assemblies in 

Rings 1 and 3 formed an approximate square. In the 1x4 pattern, the boundary area (per unit 

axial length) for Rings 1 and 3 was based on four panels per assembly. In the uniform pattern, 

the number of panels per assembly is estimated as 0.4 for Ring 1 (4(88)1/2/88) and 0.3 

(4(196)1/2/196) for Ring 3. This is a stylized representation of a uniform configuration which

limits the areas (and thus total heat transfer) between the hot rings and the rest of the 

assemblies in the pool. 
214

Figure 122 Layout of assemblies for OCP1 high-density (uniform) model

Figure 123 Layout of assemblies for OCP2 high-density (uniform) model
215

Unmitigated Moderate Leak (OCP1 Uniform) Scenario

Figure 124 and Figure 125 show the results of the calculation for the uniform OCP1. A 

comparison of the heatup with the 1x4 geometry (Figure 67) shows the higher temperatures in 

the uniform Ring 1 configuration because there is less surface area between Ring 1 and the 

colder assemblies in Ring 2. The overall thermal response, however, is comparable. At about 

30 hours, Ring 1 experiences a gradual heatup as the oxygen in the building is depleted, and 

formation of debris restricts airflow through the assemblies. Eventually, all of the fuel in Ring 1 

collapsed and formed a debris bed. There is continuous release from Rings 1 and 2 and the 

overall cesium release to the environment is about twice of that in the 1x4 geometry (see Figure 

72).

Figure 124 Ring 1 clad temperature for unmitigated uniform high-density moderate leak 

(OCP1)
216

Figure 125 Cesium environmental release fraction for unmitigated uniform high-density 

moderate leak (OCP1)

Mitigated Moderate Leak (OCP2 Uniform) Scenario

For the mitigated case in the OCP2 uniform pattern that had the highest cesium release fraction 

(1.2 percent), a number of calculations were performed to determine the effectiveness of 

mitigation. The same scenario in the 1x4 pattern did not have any release. The overall 

behavior of fuel temperature is similar to the 1x4 pattern cases in OCP2 (not shown) and OCP 1 

(Figure 77), but the fuel is experiencing a higher temperature that gradually decreases. For this 

base case (Figure 126), temperatures are high enough to cause a gap release and more 

gradual release of fission products from the fuel. Figure 127 illustrates the calculation for the 

200-gpm spray instead of the 500-gpm makeup water, which actually shows a rapid heatup 

before the temperatures are stabilized.43 A calculation was performed to test the effectiveness 

of a higher spray flow rate of 500 gpm and, as indicated in Figure 128, the fuel temperature is

stabilized at much lower temperatures without release of fission products from the fuel. In all of 

the spray calculations performed in this study, the simple flow regime model was disabled 

because of a more stable and faster calculation, and the previous results from OCP3 had 

already demonstrated that both models predict comparable maximum clad temperatures.

 43 The initial higher temperature spike for the 200 gpm spray, as compared to lower temperatures for the 500 

gpm injection case, results from a combination of the leakage versus makeup rate for this particular 

scenario. For the 500 gpm injection case, the lower portions of the assemblies are covered with water and 

high decay heat promotes steam cooling of the exposed portions of the fuel. A larger hole size would not 

have the benefit of steam cooling, and the spray is expected to perform better for a wide range of conditions. 

Even in this particular case, under quasi-steady conditions the fuel temperatures are generally lower for the 

spray case. 
217

Figure 126 Ring 1 clad temperature for mitigated uniform high-density moderate leak 

(OCP2) with 500 gpm injection

Figure 127 Ring 1 clad temperature for mitigated uniform high-density moderate leak 

(OCP2) with 200 gpm spray
218

Figure 128 Ring 1 clad temperature for mitigated uniform high density moderate leak 

(OCP2) with 500 gpm spray

Table 50 Summary of Release Characteristics for High-Density, Uniform Pattern

High 

Density 

Case #

Scenario Characteristics Release Characteristics

SFP 

Leakage?

50.54

(hh)(2) 

Equipment?

Fuel 

Uncovery

(hr)

Gap 

Release

(hr)

Hydrogen 

Deflagration

(hr)

Cs 

release 

at 72 

hours

Cs-137 

(MCi) 

Released

I 

release 

at 72 

hours

I-131 

(MCi) 

Released

OCP1

Small No 39.7 52.3 No 0.8% 0.41 4.8% 0.38

Moderate Yes 7.4 11.7 No 0.6% 0.32 0.6% 0.05

Moderate No 5.9 8.2 No 3.6% 1.88 12.4% 0.97

OCP2

Small No 42.6 55.2 65.4 4.2% 1.93 5.5% 0.61

Moderate Yes 7.3 12.7 No 1.2% 0.55 5.0% 0.56

Moderate No 5.9 8.8 21.6 49.1% 22.71 68.4% 7.65

For the offsite consequence analysis, the sequences with recently discharged fuel in a uniform 

configuration were binned in a similar manner to the low-density and high-density (1x4) loading 

scenarios. Since the licensee must either preconfigure the SFP to allow direct placement of 

discharged fuel in or move their recently discharged fuel to a more favorable configuration after 

a certain amount of time, this sensitivity simply assumes that the high-density uniform case 

becomes identical to the high-density (1x4) case after OCP2 (i.e., that the actions to meet the 

requirements on fuel pattern discussed in Section 5.1 are taken at the end of OCP2). While the 

uniform case has different release categories, the situations that lead to release are largely the 

same as the low-density and high-density (1x4) base cases. The one exception is for OCP2 

with a moderate leak and deployed 10 CFR 50.54(hh)(2) equipment, in which case a successful 
219

deployment of mitigation equipment is expected to prevent release for the high-density (1x4) 

and low-density scenario, but not for the sensitivity scenario of recently discharged fuel in a 

uniform configuration. 

Table 51 Listing of Uniform Pattern Release Sequences

High Density (uniform) Loading

Unsuccessful mitigation Deployed 50.54(hh)(2)

Sequence

Release 

Frequency 

(/yr)*

Release 

Category Sequence

Release

Frequency 

(/yr)

Release 

Category

OCP1

small 

leak 6E-09** RC12 OCP1 mod leak 6E-09 RC11

mod 

leak 6E-09 RC23 OCP2 mod leak 2E-08 RC23

OCP2

small 

leak 2E-08 RC23

mod 

leak 2E-08 RC33 No Release

OCP3

small 

leak 4E-08 RC33

mod 

leak 4E-08 RC11

Total 1E-07 Total 2E-08

* Release frequency = initiating event frequency * ac power fragility * OCP probability * liner fragility for the specified 

leak size (see Section 5.6.3 for conditional probabilities)

** Example calculation: 1.7x10-5

/yr ∙ 0.84 ∙ 0.0086 ∙ 0.05 = 6x10-9/yr
220

Table 52 reports the consequence results for the sensitivity scenario of recently discharged fuel 

in a uniform configuration. It is similar to Table 33 for the base scenarios.

Table 52 Uniform Pattern Consequence Results

SFP Fuel Loading High Density (uniform)

Seismic Hazard Frequency1 (/yr)

(PGA of 0.5 to 1.0g) 1.7E-05

50.54(hh)(2) Mitigation Credited Yes No

Conditional2 Probability of Release 0.14% 0.69%

Hydrogen Combustion Event “Not Predicted” “Possible”

Conditional3 Consequences (Release Frequency-Averaged4)

Cumulative Cs-137 Release at 72 hours (MCi) 0.5 11

Measures Related to Individual Health and Safety

Individual Early Fatality Risk 0 0

Individual Latent Cancer Fatality Risk5 Within 

10 Miles 7.3E-04(7) 6.9E-04

Measures Related to Cost Benefit Analysis

Collective Dose (Person-Sv) 1.4E+05 4.9E+05

Land Interdiction6 (mi2) 1.1E+03 1.3E+04

Long-term Displaced Individuals6 6.2E+05 5.6E+06

Consequences per year (Release Frequency-Weighted4)

Release Frequency (/yr) 2.3E-08 1.2E-07

Measures Related to Individual Health and Safety

Individual Early Fatality Risk (/yr) 0 0

Individual Latent Cancer Fatality Risk5 Within 

10 Miles (/yr) 1.7E-11 8.1E-11

Measures Related to Cost Benefit Analysis

Collective Dose (Person-Sv/yr) 3.1E-03 5.7E-02

Land Interdiction6 (mi2/yr) 2.5E-05 1.5E-03

Long-term Displaced Individuals6 (Persons/yr) 1.4E-02 6.7E-01 1 Seismic hazard model from USGS (Peterson et al., 2008) 2 Given specified seismic-event occurs 3 Given atmospheric release occurs 4 Results from a release are averaged over potential variations in leak size, time since reactor shutdown, population 

distribution, and weather conditions (as applicable); additionally, “release frequency-weighted” results are multiplied 

by the release frequency. 5 LNT and population-weighted 6 1st year post-accident; calculation uses a dose limit of 500 mrem per year, according to Pennsylvania Code, Title 

25 § 219.51

7 Slightly higher conditional risk with mitigation is due to a more prolonged release (allowing changes in wind direction 

to affect additional portions of the 10 mile area) and effective protective actions that limit individual risk (regardless of 

release magnitude); the difference is small compared to the reduction in release frequency.
221

The insights of the high density 1x4 scenario are also applicable here to the uniform pattern: 

There is very small likelihood of release. When there is a release, no offsite early fatalities

attributable to acute radiation exposure are predicted. On average, significant land 

contamination is predicted when there is a release with unsuccessfully deployed mitigation. A

significant numbers of latent cancer fatalities are also estimated; however, this is a small 

fraction of cancer fatalities from all causes, because protective actions are expected to keep 

doses below limits for habitation and ingestion. Overall, individual latent cancer fatality risk is 

very low, mainly because of the very small likelihood of release and protective actions. 

Health effects that would be induced by low dose radiation are uncertain, and insights from a 

dose truncation for the uniform pattern scenario are similar to those for the high density 1x4 

scenario. As can be seen in Table 53, dose truncation significantly lowers the estimated 

number of total latent cancer fatalities because the uncertain effects of small individual doses

are excluded.

Table 53 Dose Truncation Comparison for Uniform Pattern

Dose-Response High Density (1x4)

50.54(hh)(2) Mitigation Credited Yes No

Conditional1 Individual Latent Cancer Fatality Risk Within 10 Miles

(Release Frequency-Averaged2)

Linear, No Threshold 7.3E-04(3) 6.9E-04

620 mrem/yr truncation 3.2E-07 1.1E-06

5rem/yr or 10rem lifetime truncation 2.3E-07 1.1E-06

Individual Latent Cancer Fatality Risk Within 10 Miles (/yr)

(Release Frequency-Weighted2)

Linear, No Threshold 1.7E-11 8.1E-11

620 mrem/yr truncation 7.3E-15 1.3E-13

5 rem/yr or 10 rem lifetime truncation 5.3E-15 1.3E-13

1 Conditional on a release occurring

2 Results from a release are averaged over potential variations in leak size, time since reactor shutdown, population 

distribution, and weather conditions; additionally, “release frequency-weighted” results are multiplied by the release 

frequency.

3 Slightly higher conditional risk with mitigation is due to a more prolonged release (allowing changes in wind direction 

to affect additional portions of the 10 mile area) and effective protective actions that limit individual risk (regardless of 

release magnitude); the difference is small compared to the reduction in release frequency. 

Similar to the high density (1x4) scenario without deployed 50.54(hh)(2) equipment, the uniform 

scenario is sometimes predicted to have significant releases when there is a hydrogen 

combustion. Once again, this is because hydrogen combustion leads to much more zirconium 

oxidation from the influx of air, as well as a much smaller building decontamination factor. A 

comparison of this sensitivity analysis of a uniform fuel pattern and those of the base case (i.e. 

the high density 1x4 pattern and the low density configuration) are quantified in the Table 54

and Table 55.
222

Table 54 Consequence Comparison – High Density (1x4 and Uniform) Loading Without 

Successful 50.54(hh)(2) Mitigation

Benefit of High Density (1x4) vs. High Density (uniform) Fuel Loading

(Scenario Specific, Weather-Averaged, Release Frequency-Averaged, Unsuccessful Deployment of 50.54(hh)(2))

Type of Consequence Consequences** (/yr) Conditional* 

Consequences

Reduction Factor (dimensionless)

Release Frequency 1.0 -

Individual Latent Cancer Fatality Risk***

for 0-10 Miles 1.6 1.6

Collective Dose (Person-Sv) 1.4 1.4

Land Interdiction (mi2) 1.4 1.4

Displaced Individuals (Persons) 1.4 1.4

* Conditional on a release occurring (frequency of 1E-7 per year, or lower)

** Release Frequency-Weighted

*** Linear-No Threshold, Population-Weighted

As can be seen in Table 54, without mitigation in the high-density configurations, consequences 

of the uniform pattern are discernibly higher than the 1x4 pattern. While other contributors could 

be partially responsible for this difference, this is largely because the accident progression 

analysis predicts a uniform pattern to sometimes have more detrimental hydrogen combustion 

events than the 1x4 pattern. 

Table 55 compares consequences of high and low density with a uniform pattern for the high 

density loading, without mitigation. This is similar to Table 37 which uses a 1x4 pattern for the 

high density loading; however, Table 55 has larger differences because of the larger 

consequences predicted from a uniform pattern. 

Successfully deployed mitigation in the high density configuration lowers the release frequency 

and most conditional consequences for both uniform and 1x4 patterns. For both patterns, 

hydrogen combustions are not predicted with MELCOR when 50.54(hh)(2) mitigation is 

successfully deployed, and therefore the relatively large releases are also not predicted. 

However, deployed mitigation is not quite as effective in the uniform pattern as it is for the 1x4

pattern. Additionally, deployed mitigation is predicted to be unsuccessful at preventing an 

additional release in the uniform pattern scenario as compared to the 1x4 pattern. The 

differences in the release frequencies and conditional consequences can be seen by comparing 

Table 52 and Table 33.
223

Table 55 Consequence Comparison – High (Uniform) Density / Low Density Loading

Without Successful 50.54(hh)(2) Mitigation

Benefit of High (Uniform) Density / Low Density Loading

(Scenario Specific, Weather-Averaged, Release Frequency-Averaged, Unsuccessful Deployment of 50.54(hh)(2))

Type of Consequence Consequences** (/yr) Conditional* 

Consequences

Reduction Factor (dimensionless)

Release Frequency 1.0 -

Individual Latent Cancer Fatality Risk***

for 0-10 Miles 3.4 3.4

Collective Dose (Person-Sv) 18 18

Land Interdiction (mi2) 78 78

Displaced Individuals (Persons) 70 70

* Conditional on a release occurring (frequency of 1E-7 per year, or lower)

** Release Frequency-Weighted

*** Linear-No Threshold, Population-Weighted

9.4 Sensitivity to Multiunit or Concurrent Accident Events (MELCOR)

These sensitivity calculations are intended to show the importance of the reactor building in the 

progression of accident in the SFP and the source term with a concurrent reactor accident. In 

the base calculation for a high-density, moderate leak scenario in OCP3 (see Figure 129), the 

fuel heats up and a zirconium fire is initiated. The reactor building refueling bay remains intact 

during the rapid draindown of the pool and there is very low hydrogen generation in the SFP. 

As the accident progresses, the atmosphere of the reactor building heats up as air is circulated 

through the assemblies. The oxidation of the SFP fuel depletes the oxygen in the reactor 

building and limits any long-term air oxidation and the associated exothermic power. 

Consequently, the long-term fuel heatup is limited primarily by decay heat, and the source term 

is relatively small (1.7-percent cesium release to the environment). The sensitivity calculations 

assume failure of the reactor building as a result of the hydrogen combustion caused by leakage 

from the containment (as evidenced from the SOARCA analysis and the Fukushima accident). 

The failure of the reactor building is based on the results of the PBAPS short-term SBO 

calculations for SOARCA (with and without reactor core isolation cooling (RCIC) blackstart). 

The reactor building failure times are at 8.5 hours (without RCIC blackstart) and 16.9 hours (with 

RCIC blackstart). It is further assumed that the failure of the reactor building and formation of 

debris in the pool results in a reduction of flow area at the exit of the assemblies (50 percent of 

nominal flow area) and increased flow losses. 

Figure 130 and Figure 131 show the thermal response of the SFP with early (8.5 hours) and late 

(16.9 hours) failure of the reactor building. With early failure of the reactor building (before 

significant fuel heatup), the circulation of the cool air limits the fuel heatup and there is no 

release from the fuel. With the reactor building intact (Figure 129), the reactor building

atmosphere keeps heating up, which limits the convective cooling of the assemblies. With late 

failure of the reactor building, the fuel becomes hot enough that a sudden increase in the flow of 

oxygen through the assemblies ignites and rapidly leads to significant air oxidation (zirconium

fire). The fuel heats up leading to degradation and finally relocation (see Figure 131). This 

leads to 60-pecent cesium release to the environment. Finally in OCP4 (see Figure 132 and 

Figure 133), the decay heat and peak fuel temperatures are lower. The reactor building failure 
224

has no impact on the accident progression because the accident is not oxygen-limited. In fact, 

the reactor building failure (and thus lower temperature of circulating air) leads to lower fuel 

temperatures. 

Table 56 compares the source term for low-density OCP1 for unmitigated small and medium 

leaks. In both cases, the loss of the reactor building, and thus the effectiveness of natural 

decontamination, leads to higher release by a factor of 2 to 4 depending on the radionuclide 

class.

Figure 129 Ring 1 clad temperature for unmitigated high-density moderate leak (OCP3) 
225

Figure 130 Ring 1 clad temperature for unmitigated high-density moderate leak (OCP3; 

early reactor building failure)

Figure 131 Ring 1 clad temperature for unmitigated high-density moderate leak (OCP3; 

late reactor building failure)
226

Figure 132 Ring 1 clad temperature for unmitigated high-density moderate leak (OCP4) 

Figure 133 Ring 1 clad temperature for unmitigated high-density moderate leak (OCP4; 

late reactor building failure) 
227

Table 56 Comparison of Low-Density OCP1 Release Fractions for a Concurrent Reactor 

Accident

Environmental release 

fraction

Small Leak Moderate Leak

Base case Late reactor 

building failure

Base case Late reactor 

building failure

Xe/Kr 1.39E-01 1.85E-01 8.54E-02 8.87E-02

Cs 3.13E-02 1.18E-01 4.58E-03 1.49E-02

Ba 4.39E-03 9.66E-03 1.08E-03 4.45E-03

I 4.55E-02 1.41E-01 1.66E-02 5.60E-02

Te 4.54E-02 1.40E-01 1.68E-02 5.76E-02

Ru 2.17E-05 9.84E-05 2.09E-05 4.93E-05

Mo 8.86E-03 3.51E-02 2.60E-03 6.13E-03

Ce 1.49E-09 6.08E-09 4.94E-10 1.01E-09

La 1.34E-09 5.67E-09 4.37E-10 8.96E-10

9.5 Sensitivity to Molten Core-Concrete Interaction (MELCOR/MACCS2)

Accident Progression Analysis (MELCOR)

This sensitivity is a variation of the previous sensitivity calculation with late reactor building

failure caused by a concurrent reactor accident. Even without MCCI, the SFP concrete floor 

starts to heat up and by the end of 3 days, a portion of the concrete experiences temperatures 

in excess of its ablation temperature (assumed to be 1500 K). Figure 134 shows the contours 

of temperature in the SFP floor. In the present sensitivity calculation, MCCI is assumed to be 

initiated in a control volume that becomes active once the floor liner melts and the debris 

contacts the concrete. Figure 135 shows the environmental release fraction of cesium. Without 

MCCI, the releases from the fuel are dominated by diffusion from the fuel matrix grain 

boundaries as modeled in the CORSOR-Booth model in MELCOR. The MCCI releases are 

modeled by the VANESA model in MELCOR which takes into account sparging of the concrete 

decomposition of gases and the presence of metal in the melt.44 The release fraction of cesium

is identical in both calculations (see Figure 16) until MCCI starts in Rings 1 and 2 at about 35 

hours. MCCI results in a sudden increase in cesium release (and other fission products) at 35 

hours and then again at 40 hours (start of MCCI in Rings 3 and 4) as soon as zirconium is 

added to the melt interacting with the concrete. In general, the release fractions with MCCI are 

higher, and for cerium and lanthanum groups, the MCCI releases are orders of magnitude 

higher. 

 44 There are some limitations in representing MCCI in a SFP using MELCOR. The MELCOR MCCI model 

was developed to represent a pour of core debris from a failed reactor into a confined reactor cavity. In 

contrast, the relocation of fuel onto the SFP liner could be highly dispersed, especially in a favorable 

configuration. There could be regions of low-decay heat assemblies surrounding failed high-powered 

assemblies or open regions under the racks where only the high-powered assemblies relocated to the 

SFP liner. The MELCOR MCCI model immediately mixes all debris into a uniform debris bed with 

uniform temperature and decay heat power. Nevertheless, the MCCI sensitivity calculations illustrate 

the potential impact of MCCI physics on the radionuclide chemical form (and volatility) and the 

associated release of radionuclides to the environment. Certain radionuclide species can become more 

volatile in the presences of sparging ablation gases, which leads to the differences in Table 57.
228

Figure 134 SFP concrete floor temperature for unmitigated high-density moderate leak 

(OCP3; late reactor building failure) 

Figure 135 Cesium release fraction for unmitigated high-density moderate leak (OCP3; 

late reactor building failure) with and without MCCI
229

Table 57 Comparison of Release Fractions with and without MCCI.

Environmental release fraction Without MCCI With MCCI

Xe/Kr 0.92 0.92

Cs 0.61 0.88

Ba 0.01 0.07

I 0.83 0.91

Te 0.80 0.74

Ru 0.01 0.003

Mo 0.15 0.11

Ce 1.7E-07 0.007

La 1.6E-07 0.0002

Offsite Consequence Analysis (MACCS2)

The sequence used in the accident progression analysis was analyzed with MACCS2 to 

understand how MCCI affects offsite consequences. The sequence analyzed was the OCP3 

moderate leak scenario, with hydrogen combustion in the refueling bay at 16.9 hours as 

predicted from the SOARCA short-term SBO with RCIC blackstart scenario. 

The focus of this study was specifically on the SFP, and therefore, this sequence is not part of 

the main results. Rather, these are part of a different sensitivity investigating the effects of 

concurrent reactor events. Therefore both sequences with and without MCCI were calculated 

with MACCS2 in order to focus on MCCI. The individual consequence results of these 

sequences are not reported; however the effect of MCCI on the offsite results is shown in Table 

58. 

Table 58 Consequence Comparison – Molten Core Concrete Interaction

Molten Core Concrete Interaction Sensitivity

(Weather-Averaged; OCP3 Moderate Leak sequence with reactor 

building failure at 16.9 hours)

Type of Consequence Conditional* Consequences

Percent Increase

Individual Latent Cancer 

Fatality Risk** for 0-10 Miles -17%

Collective Dose 9%

Land Interdiction (mi2) 33%

Displaced Individuals (Persons) 17%

* Conditional on a release occurring (frequency of 1E-7 per year, or lower)

** Linear-No Threshold, Population-Weighted

No early fatalities are predicted for these sequences. These sequences have considerable 

release fractions, including increased contributions from some of the typically non-volatile 

chemical groups. However, these characteristics were not enough to reach the dose 

thresholds associated with early fatalities in OCP3, for which the last fuel offload has cooled for 

37 days since shutdown. 

Although the release is larger with MCCI, the individual LCF risk for 0-10 miles non-intuitively 

decreases. The reason for this is likely due to the significant level of protective actions and the 

type of radionuclides in the different source terms. The radionuclides in the MCCI sensitivity 
230

likely have different dose contributions relative to their LCF risk contribution, and therefore are 

more likely to cause protective actions despite having relatively lower risk factors, which in turn 

causes lower LCF risk. 

To verify this phenomenon, the risk and dose contributions of different radionuclides to offsite 

consequence could be investigated. This is not done here; however, when farther distances are 

included (which are areas where protective actions and this phenomenon are less likely to 

occur), this reduction in the results no longer exists. This can be seen in the effect of MCCI on 

the collective dose in Table 58. Similarly, land interdiction and displaced individuals have higher 

consequences, as one may expect with relatively larger source terms. 

9.6 Sensitivity to Radiative Heat Transfer (MELCOR)

For this sensitivity calculation, the surface area between Rings 2 and 4, Rings 4 and 6, and 

Rings 6 and 7 are modified by plus or minus 25 percent. Although Rings 2, 4 and 6 are empty, 

they still contain rack components and can impact the heatup in Rings 1, 3, and 5. Table 59

reports the results of the calculation for the unmitigated small leak (OCP2), low-density 

configuration. In general, the highest differences are observed for the reduction in the area 

(approximately 30-percent reduction in the release of volatiles), while increasing the surface 

area only has a modest effect on the release. The fuel heatup, radiation between rack 

components, initiation of air cooling, and interaction between different assemblies and racks are 

all complex phenomena that contribute to fission product release in a nonlinear fashion. 

Nevertheless, the releases for the low-density case are still small, and uncertainties in radiation 

modeling do not seem to significantly change the results 

Table 59 Low-density OCP2 Release Fraction Sensitivity to Ring-Ring Radiation

Base Case -25% surface area +25% surface area

Xe/Kr 4.41E-02 4.44E-02 4.40E-02

Cs 1.71E-02 1.23E-02 1.66E-02

Ba 5.19E-03 3.69E-03 5.06E-03

I 3.31E-02 2.50E-02 3.22E-02

Te 3.54E-02 2.53E-02 3.45E-02

Ru 9.27E-07 9.27E-07 8.75E-07

Mo 9.95E-05 9.93E-05 9.39E-05

Ce 3.54E-11 4.03E-11 3.16E-11

La 3.58E-11 4.06E-11 3.19E-11

9.7 Sensitivity to Land Contamination (MACCS2)

The measure of contaminated land area can vary significantly with the criterion used to measure 

or estimate the level of contamination. This study calculates the land that exceeds 500 mrem in 

the first year after the accident as an indicator for land contamination, based on the 

Pennsylvania 500 mrem annual dose limit for habitability. However, other protective action 

levels exist that can also be used as an indicator for measuring land contamination. These 

protective actions tend to be related to dose levels associated with either land interdiction or 

decontamination, but not necessarily. 

Radioactivity levels are not typically used as a basis for protective actions. Instead, activity 

levels are usually measurements for estimating different dose levels, which are in turn used as 

the basis for protective actions. A range of typical activity levels for Cs-137 is included in the 
231

table below. These particular levels have been widely reported as the zoning criteria for the 

Chernobyl nuclear disaster.

The EPA intermediate phase PAG levels are 2 rem in the first year, and 500 mrem annually 

thereafter. Previous studies have typically used one criterion (4 rem in 5 years) to represent 

these PAG levels. How well this represents the actual EPA PAG levels was not analyzed here, 

although this criterion is included here. 

For simplicity, this sensitivity was based on the weather-average results of a single accident 

sequence. This is unlike the results of the production analyses, which are frequency-weighted 

averages of all the release sequences. The sequence chosen was the OCP3 small leak from a 

high density SFP with unsuccessful deployment of 50.54(hh)(2) equipment, which has a cesium 

release fraction of 42% at 72 hours. Since the consequence results of individual sequences are 

not reported, the results of this sensitivity have been normalized. 

Table 60 Consequence Comparison – Land Contamination Sensitivity

Total Land Area Sensitivity to Dose/Activity Criteria

(Weather-Averaged; OCP3 Small Leak from high density SFP with unsuccessful deployment of 

50.54(hh)(2) equipment)

Dose

Protective Action Basis Dose Level1 Land Area 

EPA intermediate phase PAG2: 1st year 2 rem 21%

EPA intermediate phase PAGs2 (as 

commonly represented in previous studies) 4 rem / 5 years 32%

Pennsylvania dose limit to the public 500 mrem 100%

ICRP recommendation3

100 mrem 361%

10CFR Part20 Subpart D

10CFR Part20 Subpart E4 25 mrem 731%

Activity

Protective Action Basis

Activity Level

(Cs-137 Bq/m2) Land Area 

- 1.48E+06 73%

- 5.55E+05 181%

- 1.85E+05 346%

- 3.70E+04 557% 1

Annual doses, unless otherwise noted

2 EPA intermediate phase PAGs are: 2 rem in the first year, and 500mrem annually thereafter. 3 ICRP recommends using the lower portion of a band that spans 1-20 mSv as a reference level for protective 

measures, and past experience demonstrates 1mSv is typical. 4 10CFR Part20 Subpart E also includes ALARA, which is not considered here.

As seen from the table above, different dose or activity levels can significantly change the 

amount of land area that exceeds a given limit. In addition to the total land area, a range of 

different distances were also analyzed in the graph below. 
232

*Weather-averaged; OCP3 Small Leak from high density SFP with unsuccessful deployment of 50.54(hh)(2) 

equipment

Figure 136 Land Contamination Sensitivity to Dose Criteria

At shorter distances, the change in the land area that exceeds a given dose limit is not 

significant, while at far distances, the change can be more than a factor of 10. The distance 

where the amount of land contamination becomes sensitive to different dose criteria is expected 

to depend on the initial concentration and the deposition rate. For this release magnitude, most 

of the plume exceeds all of the dose criteria at close distances. However, irrespective of the 

release magnitude, the affected area will increase and the concentrations will decrease as the 

plume spreads. Therefore, for all releases, land contamination is more dependent on the dose 

criteria at far distances. 

9.8 Sensitivity to Time Truncation (MELCOR/MACCS2)

Project staff judged that a reasonable approach for the project is to consider radionuclide 

releases only if the fuel has become uncovered by 48 hours and to assume that any potential 

radiological release is stopped at 72 hours (Section 5.3). However, the use of a time truncation 

is uncertain, and is capable of significantly affecting the consequences. This assumption could 

be pessimistic since many resources are available at the State, regional, and national level that 

could be available to potentially truncate the accident more aggressively. On the other hand, 

this time truncation could be optimistic, as it assumes that an ongoing spent fuel pool release is 

capable of being truncated.

Given the uncertainty, this sensitivity considers the effects of both a more aggressive and a less 

aggressive time truncation. For simplicity, the sensitivity of longer time truncation was based on 

the weather-average results of a single accident sequence. This is unlike the results of the 

production analyses, which are frequency-weighted averages of all the release sequences. The 

sequence chosen was the OCP3 small leak from a high density SFP with unsuccessful 

deployment of 50.54(hh)(2) equipment, which has a cesium release fraction of 42% at 72 hours. 

Since the consequence results of individual sequences are not reported, the results of this 

sensitivity are reported as the fractional increase in consequences over the original results. The 

sensitivity of a shorter time truncation discusses in which sequences releases would be averted.

1%

10%

100%

1000%

0-10 10-50 50-100 100-200 200-500

Distance (miles)

Land Contamination* Sensitivity

to Dose Criteria

2rem

4rem / 5years

500mrem

100mrem

25mrem
233

A period of 96 hours was chosen to represent a less aggressive time truncation. MELCOR and 

MACCS2 calculations were extended to 96 hours from the original 72 hours. The effect on the 

release fractions and the relative effect on offsite consequence can also be seen in the Figure 

137 and Table 61 below.

Figure 137 Atmospheric release fractions for unmitigated high density small leak (OCP3)

with a 96 hour time truncation

Table 61 Consequence Comparison – Time Truncation Sensitivity

Time Truncation Sensitivity: 72 Hour vs. 96 Hour

(Weather-Averaged; OCP3 Small Leak from high density SFP with 

unsuccessful deployment of 50.54(hh)(2) equipment)

Type of Consequence Conditional* Consequences

Percent Increase

Individual Latent Cancer 

Fatality Risk** for 0-10 Miles 38%

Collective Dose (Person-Sv) 27%

Land Interdiction (mi2) 28%

Displaced Individuals (Persons) 27%

* Conditional on a release occurring (frequency of 1E-7 per year, or lower)

** Linear-No Threshold, Population-Weighted

A shorter time truncation, however, can also significantly affect the results. For the sequences 

involving a small leak from a high density SFP during OCP1 and OCP2 with unsuccessful 

deployment of 50.54(hh)(2) equipment, fuel uncovery occurs around 40 and 43 hours 

(compared to a baseline time truncation of 48 hours). Using a time truncation less than 40 and 

43 hours respectively, would avoid releases for these sequences. In other scenarios, fuel 

uncovery and release occur much sooner than the baseline time truncation.
234

These results highlight that some releases are expected to be prolonged and therefore a choice 

in a time truncation can affect offsite consequence predictions. 

9.9 Sensitivity to Reactor Building Leakage (MELCOR)

Four sensitivity calculations were performed to examine the impact of the reactor building

leakage on hydrogen combustion and accident progression. These covered the small leak 

scenarios in OCP2 and OCP3 without successful deployment of mitigation. Two larger leak 

sizes were considered, (1) an increase in the nominal leakage area by a factor of 10, and (2) an 

increase in the nominal leakage area corresponding to area of a blowout panel. In general, 

while an increase in area by a factor of 10 increases the leakage, any further increase in area 

has no effect since the building pressure adjusts to limit the leakage. The leakage area has no 

significant impact on accident progression, and since the hydrogen is produced over a relatively 

short time, the hydrogen mole fraction quickly reaches the 10% threshold for ignition. The Cs 

release fractions are not significantly altered. In OCP2, Cs release fraction is reduced by ~12% 

while it is increased by ~2% in OCP3 owing to slight variations in the course of the accident. 
235

10. ASSESSMENT OF PREVIOUS STUDIES OF SAFETY 

CONSEQUENCES ASSOCIATED WITH LOADING, TRANSFER, AND 

LONG-TERM DRY STORAGE

10.1 Introduction

Staff has performed an assessment to 1) identify previous studies of safety consequences of 

spent fuel accidents in both wet and dry storage, 2) determine the extent to which those 

previous studies are comparable to results from the SFPS, and 3) to the extent practicable, 

update the results of the previous studies to facilitate a comparative assessment. The SFPS 

discusses off-site consequences of a spent fuel pool accident in Chapter 7, and provides limited 

discussion of several similar previous spent fuel pool studies. This Chapter provides a more 

detailed comparison between off-site consequences calculated for the SFPS and those 

calculated from previous studies. This Chapter also provides a comparative assessment of 

SFPS results against previous studies of the safety consequences associated with loading, 

transfer, and long-term storage in dry cask storage systems (DCSS). In these assessments, 

staff limited its focus to offsite consequences of accidental releases at commercial nuclear 

power plants. Specifically, these assessments compare the direct impacts due to offsite 

radiological exposure and the indirect (e.g., economic or land use) impacts of protective 

measures taken to avert offsite radiological exposure, of the various studies considered. Offsite 

impacts from routine operations, doses to workers from routine or accidental exposures, or nonsafety related impacts such as costs of spent fuel management, were not considered. 

Furthermore, staff focused on studies associated with accidents, rather than studies of safety 

consequences associated with deliberate human actions such as sabotage or terrorism. 

10.2 Previous Spent Fuel Pool Studies

There have been several previous studies of the consequences of spent fuel pool accidents. 

These include those in support of Generic Safety Issue 82 and of consequences from spent fuel 

pool accidents at shutdown nuclear power plants:

- "Severe Accidents in Spent Fuel Pools in Support of Generic Safety Issue 82" (NUREG/CR4982, 1987)

- "Value/Impact Analyses of Accident Preventive and Mitigative Options for Spent Fuel Pools," 

(NUREG/CR-5281, 1989)

- "Regulatory Analysis for the Resolution of Generic Issue 82 'Beyond Design Basis Accidents 

in Spent Fuel Pools'" (NUREG-1353, 1989) 

- “A Safety and Regulatory Assessment of Generic BWR and PWR Permanently Shutdown 

Nuclear Power Plants” (NUREG/CR-6451, 1997)

- “Technical Study of Spent Fuel Pool Accident Risk at Decommissioning Nuclear Power 

Plants (NUREG-1738, 2001)

The studies conducted to evaluate beyond design basis accidents in spent fuel pools in the late 

1980’s (NUREG/CR-4982, NUREG/CR-5281, and NUREG-1353) report a variety of impacts 

related to both radiological doses (e.g., collective doses), as well as the potential impacts

associated with limiting radiological doses, such as costs and extent of land condemnation. 

NUREG-1353 reports collective doses within a 50 mile radius of 8 to 26 million person-rem per 

event based on MACCS calculations documented in NUREG/CR-5281. It also reports an 

interdiction area (“area with such a high level of radiation that it is assumed that it cannot be 

decontaminated”), based on CRAC2 calculations from NUREG/CR-4982, of 0 to 244 square 
236

miles (within a 50 mile radius) and offsite property damages of $3 to $26 billion in 1983 dollars 

(also within a 50 mile radius), based on MACCS calculations documented in NUREG/CR-5281. 

The more recent studies conducted to examine the risks from spent fuel pool accidents at 

shutdown nuclear power plants (NUREG/CR-6451 and NUREG-1738) also report a variety of 

both radiological and non-radiological impacts. NUREG-1738 reports radiological impacts of 0

to 200 early fatalities and potential latent cancer fatalities (out to 500 miles) in the hundreds of 

thousands. These studies used a variety of assumptions regarding pool inventory, release 

fraction, population density, and emergency response. The results of previous spent fuel pool 

studies are compared to the SFPS results for various consequence metrics in Table 62.

10.2.1 Quantitative Comparison of Spent Fuel Pool Analytical Results

The following table presents selected consequence results from previous studies of spent fuel 

pool accidents, and the SFPS.
237

Table 62 Comparison of consequence results from current and previous spent fuel pool an

Metric NUREG/CR4982, 

Table 4.7

NUREG/ CR5281, 

Table 3.2

NUREG/ CR6451, Tables 

4.1/4.2

NUREG-17381

Tables 

3.7-1/3.7-2

Early fatalities 

(0 to 500 miles) Not reported Not reported 0 to 101 0 to 200 

Individual LCF risk within 10 

miles (conditional) Not reported Not reported Not reported 7.7e-4 to 8.2e-2 

Collective dose within 50 

miles in Person-Sv

11,000 to 

26,0005 80,000 to 

256,0007

30,000 to 

810,000

37,000 to 

240,000

Collective dose within 500 

miles in Person-Sv 710,000 40,000 to 3,400,000

450,000 to 

600,000

Interdicted land (square 

miles) Not reported Not reported Not reported Not reported 

Condemned land (square 

miles) 4 to 2244,5,6 Not reported 1 to 2,800 Not reported 

1. Results presented in Section 3.7 are taken from there; otherwise values are from Appendix 4, Note that the upper end of these valu

high ruthenium source terms with late evacuation.

2. 0 to 500 mi results are actually 0 to 1000 mi results, which is likely analogous to past study modeling assumptions; uniform pattern r

this time; only LNT results are presented

3. The range of results is not bounding, as it does not represent ranges due to many uncertainties such as weather, operating cycle ph

states. Direct consideration of these uncertainties would increase the range, as it likely would for the previous studies as well. 

4. Note that the definition of interdicted land is not consistent with the definition used in the SFPS report. The text in NUREG/CR-4982 

reported is permanently uninhabitable land, which is analogous to condemned land.

5. This range is for fire scenarios. For the non-fire scenario, the values were 4 person-rem and 0.0 sq. mi interdiction area.

6. Note that this metric does not change between cases 1A (50 miles) and 1C (500 miles), indicated that there is no additional condem

in this analysis

7. The range at which this metric was computed is not specified in NUREG/CR-5281

237
238

10.2.2 Comparison of SFPS Results to previous Spent Fuel Pool Studies

Comparison of SFPS results to past spent fuel pool studies is not straight-forward, because 

those studies reported a variety of consequence metrics and used a range of assumptions 

regarding pool inventory, release fraction, population density, and emergency response. These 

ranges present a variety of approaches to represent uncertainties from select input parameters, 

depending on the study. For instance, the range of NUREG-1738 results represents a range 

due to evacuation times, ruthenium release modeling, time since reactor shutdown, and two 

competing seismic hazard models. The range of the SFPS, on the other hand, represents a 

range due to uncertainty in deployment of mitigation equipment and variations of potential pool 

loading density. In addition, the SFPS results are expected to be sensitive to uncertainties in 

hydrogen combustion ignition criteria and the time truncation value (and these uncertainties are 

not reflected in the range of results), as well as uncertainties in weather, decay power, and pool 

damage states (which are not explicit in the range of results since average results are 

presented). It is also important to remember that past studies generally used generic 

assumptions intended to envelope the situation, as opposed to the focus on site-specificity with 

the SFPS. Nevertheless, these ranges of consequence metrics are often cited by external 

stakeholders, and thus comparison is informative.

A comparison of the release characteristics from previous spent fuel pool studies demonstrates 

that releases of cesium are generally less in the current study than in previous studies, and the 

time from accident initiation to release to the offsite environment is generally longer:

Table 63 Comparison of Source Terms from Current and Previous SFP analyses

Resolution of GI-82 (NUREG1353, NUREG/CR-4982, 

NUREG/CR-5281):

NUREG-1738 SFPS (preliminary results):

• 10 to 100% Cs release 

(100% assumed for cases 

1 and 2)

• Release over 8 hours for a 

propagating SFP zirc fire 

(assumed)

• 0.25 (BWR) or 1.0 (PWR) 

conditional probability if 

fuel becomes uncovered

• 75% Cs release (assumed 

from NUREG-1465)

• Instantaneous draindown 

for large seismic

• 2 to 14 hour heatup 

depending on fuel age 

(see Table 1A-1)

• Cs release = < 1% to 49%

• Draindown to uncovery –

2.5 to 43 hours (when leak 

exists)

• Start of release = 8 hours 

to > 72 hours

The lack of any early fatalities attributable to acute radiation exposure in this study is consistent 

with results of some past SFP studies, and much lower than others (e.g., up to 200 early 

fatalities from NUREG-1738). The range of latent fatalities predicted in this study is consistent 

with the lower end of the range reported in past SFP studies. The conditional individual latent 

cancer fatality risk from 0 to 10 miles for the scenarios studied in this report is several orders of 

magnitude below that reported in NUREG-1738, which was the only other study to report this 

metric. Even when the early evacuation scenario from NUREG-1738 is used for comparison 

(average individual risk is in the range of 2.6E-3 to 4.8E-3), the results from the current SFPS 

study are significantly lower. The collective dose values predicted in this study are consistent 

with the lower end of the range reported in past SFP studies. The SFPS reports temporarily 

interdicted land (uninhabitable land during the first year following the postulated accident), in 

order to remove uncertainty in longer-term effects and policies related to weathering and de-
239

contamination decisions. Reporting interdicted land makes the results incomparable to the past 

SFP studies which have presented condemned land. The SFPS does not report other aspects 

of offsite property damage.

10.3 Previous Dry Cask Storage Studies

The number of studies of the consequences from dry cask handling and storage accidents are 

more limited than those for spent fuel pools. Safety analysis reports for dry cask storage 

systems, submitted in support of applications or renewals for site-specific independent spent 

fuel storage installation (ISFSI) licenses or for DCSS certificates, include some information on 

offsite consequences of potential accidents (e.g., tornado missile impacts, earthquakes, floods). 

However, such accidents are generally shown by analysis not to result in a release, and the 

likelihood of more severe accidents is sufficiently low that the consequences need not be 

explicitly evaluated. Staff identified one previous NRC analysis on the offsite safety 

consequences of accidents from dry cask storage systems. The report, “A Pilot Probabilistic 

Risk Assessment of a Dry Storage System at a Nuclear Power Plant” (NUREG-1864, 

ML071340012), documents a pilot PRA for a specific dry cask system (Holtec International HISTORM 100) at a specific boiling-water reactor (BWR) site. The study included an assessment 

of potential offsite consequences from the drop and failure of a cask. It provides estimates of 

the annual risk for one cask in terms of the individual probability of a latent cancer fatality within 

16 km (10 miles) of the site, and also reports that there are no prompt fatalities. The 

assessment was performed using MACCS2 for a representative site and is described in detail in 

Appendix E to NUREG-1864. Site-specific data important to modeling a HI-STORM dry cask 

30.5 meter (100-foot) drop accident scenario in the MACCS2 consequence calculation were 

collected and used. The important parameters/variables required to model the site are the 

population density/distribution and the site meteorology. The radionuclide inventory, source term 

(i.e., release fraction, release start time, and release duration), initial plume dimensions (related 

to the system geometry), and plume heat content were described. Other settings and models 

necessary for a MACCS2 calculation (e.g., food chain model) were taken from the NUREG1150 study MACCS2 input file prepared for the Surry Power Station. The input file is 

documented in Appendix C to the MACCS2 code manual and is referred to there as Sample 

Problem-A.

10.3.1 Supplemental Analyses

In order to provide quantitative estimates of safety consequences for accidents during dry cask 

handling and storage that are directly comparable with the results of the SFPS, and to provide 

additional output metrics for comparison, staff performed limited MACCS2 supplemental dry 

cask storage analyses. These supplemental analyses used the source term characteristics 

from NUREG-1864 coupled with the site-specific characteristics reflected in the MACCS input 

decks used in the SFPS analyses. The analyses conducted in NUREG-1864 were conducted at 

a different geographic location than the site selected for the SFPS and evaluated impacts only 

in terms of selected human health metrics (the individual probability of a prompt fatality within 

1.6 km (1 mile) and of latent cancer fatality within 10 miles, and the individual lifetime dose 

commitment). These metrics can be affected by site-specific characteristics such as 

meteorology and population distributions surrounding the site. To perform this analysis, staff 

modified the MACCS input files used for the analyses in the SFPS (described in detail in 

Chapter 7 of the SFPS) with the revisions discussed below.
240

Changes related to meteorology, site characteristics, and dosimetry:

No changes were made to the SFPS input deck related to meteorology, site demographic and 

economic characteristics, or dosimetry. Site data, including weather, population, and land 

values are therefore consistent with SFPS results. The dosimetry files used are consistent with 

FGR-13, whereas NUREG-1864 used the dose conversion factors used in NUREG-1150. This 

is a potential source of difference from the results reported in NUREG-1864.

Changes related to source term and release:

The radiological inventory was changed to be consistent with Table E.1 of NUREG-1864. In 

addition, a limited set of radionuclides present in the SFPS input deck that are expected to be in 

secular equilibrium with the nuclides listed in Table E.1 (Ba-137m, Pr-144, and Rh-106) were 

added with an activity equal to that of their parent radionuclide. However, a limited set of 

nuclides (Pm-147, Eu-154, Am-242m, Am-243, and Cm-243) reported in NUREG-1864 Table 

E.1 were not used in the SFPS MACCS2 input deck. Because the dosimetric data for these 

nuclides was not developed in the SFPS input deck, these radionuclides were not included in 

the modeled inventory for the supplemental analysis. Based on the much larger inventory of 

fission products such as Cs-137 and Sr-90, and of actinides such as Pu-241, the omission of 

these nuclides is not expected to significantly affect the results; however, this is a potential 

source of difference from the results reported in NUREG-1864. The number of chemical groups 

was changed to three to represent noble gases (NG), activation products (CRUD) and 

particulates (PART) to be consistent with the NUREG-1864 source term. Consistent with the 

NUREG-1864 source term, the only nuclide in the noble gas chemical group was Kr-85, and the 

only nuclide in the activation product chemical group was Co-60. For consistency with NUREG1864, all other nuclides were assigned to the particulate group in view of the fact that releases 

from dry casks are likely to result from impacts at a sufficiently low temperature that 

radionuclides would be released by mechanical means rather than because of different 

volatilities. The inventory modeled in this supplemental analysis is provided below:

Table 64 Modeled Inventory for Supplemental Reanalysis

Nuclide Bq Chemical Group

Co-60 1.61E+14 CRUD

Kr-85 2.77E+15 NG

Sr-90 3.40E+16 PART

Y-90 3.40E+16 PART

Ru-106 2.92E+14 PART

Rh-106 2.92E+14 PART

Cs-134 5.13E+15 PART

Cs-137 5.54E+16 PART

Ba-137m 5.54E+16** PART

Ce-144 5.08E+13 PART

Pr-144 5.08E+13** PART

Pm-147 0* (3.37E+15) PART

Eu-154 0* (4.15E+15) PART

Pu-238 3.98E+15 PART

Pu-239 1.87E+14 PART

Pu-240 3.47E+14 PART
241

Pu-241 5.23E+16 PART

Am-241 1.20E+15 PART

Am-242m 0* (1.97E+13) PART

Am-243 0* (3.07E+13) PART

Cm-243 0* (3.02E+13) PART

Cm-244 5.66E+15 PART

*These nuclides were not included in the supplemental analysis, as discussed above. The values from NUREG-1864 

are provided in parentheses to allow comparison of source terms.

**These short-lived progeny were not in Table E.1 of NUREG-1864 but are included in the SFPS input deck. These 

were included in this table to represent the fact that these are likely to be in secular equilibrium with their parent 

radionuclides.

The particle size distribution assumed for NUREG-1864 was not identified in Appendix E. For 

purposes of this supplemental analysis, the particle size distribution for the particulate and 

activation product chemical group was assigned to be equal to the particle size distribution of 

the lanthanide chemical group in the SFPS, as lanthanides are presumed to be released due to 

mechanical measures rather than by volatility. Although this distribution is not the most 

appropriate for a dry cask storage scenario, using a cask-specific distribution would likely not 

change the conclusions of this Chapter. Specifically, the larger particle sizes expected to be 

associated with such a scenario would result in more deposition closer to the site, resulting in 

fewer exposed individuals within ten miles. The values used are given below:

Table 65 Particle Size Information

Particle 

Size Group

Particle Size 

Distribution

Dry Deposition 

Velocity (m/s)

1 3.2% 0.0011

2 15% 0.001

3 29% 0.0014

4 21% 0.0023

6 10% 0.0045

6 3.0% 0.0092

7 1.50% 0.0177

8 0.60% 0.0291

9 0.20% 0.0367

10 16% 0.0367

The release height and release fractions were varied to be consistent with NUREG-1864, Table 

E.1, as given in Table 66 below. To simulate the short duration release modeled in NUREG1864, the number of plume segments was reduced to one with release starting at time zero, 

with a two minute (120 second) release duration. Reflecting the primarily mechanical rather 

than thermal nature of the release, the plume rise model was changed to a heat only option with 

a power of 18 kW to be consistent with NUREG-1864. However, parameters associated with 

building wake effects (e.g., building height, initial plume dimensions) were chosen to be 

consistent with SFPS values, as these values would be site specific. This represents another 

potential source of difference with NUREG-1864 values.

Changes related to emergency response and long-term protective actions

Consistent with the immediate release model and no evacuation assumption in NUREG-1864, 
242

the supplemental analysis eliminated all evacuating cohorts by changing the evacuation model 

to “No Evacuation”. However, sheltering and relocation parameters remained consistent with 

SFPS estimates. No changes were made to SFPS parameters for long-term protective actions 

such as decontamination levels and costs, as these were selected to be consistent with the 

SFPS site-specific values to allow for comparability. The application of SFPS emergency-phase 

sheltering, relocation, and long-term protective action parameters represent a source of 

difference between the results of NUREG-1864 and the supplemental analysis.

The results, and their comparability to the results provided in NUREG-1864, are provided in 

Table 66 and Table 67. Results are provided for a range of release fractions and release 

heights to facilitate comparison with the results reported in NUREG-1864. Staff considers the 

upper end of the release fraction for particulates in NUREG-1864 (0.12%) to represent a very 

conservative estimate of the potential respirable particulate release from a breached cask, as it 

assumes essentially complete fragmentation and entrainment of the high-burnup rim region and 

very limited filtration (10% released) within the cladding-fuel gap during entrainment flow. 

Reporting the full range of results, consistent with the results presented in Table E.1 of NUREG1864, allows a more informed comparison of results including the effects of potential 

conservatisms in the analyses.

Results are reported for a variety of output metrics. These include both direct measures of 

health impacts (doses and probabilities of early and latent fatalities) as well as indirect 

measures such as the amount of land that is either temporarily interdicted or permanently 

condemned) or the numbers of temporarily or permanently displaced individuals.
243

Table 66 Parameters and Results from NUREG-1864, Table E.3

Release Fraction

Noble 

Gases

Particles CRUD Release 

Height 

(m)

Ind. Risk 

of 

Prompt 

Fatality 

within 10 

mi

Ind. Risk 

of LCF 

within 10 

mi

Ind. 

Peak 

Dose at 

1.2-1.6 

km (Sv)

0.12 1.2E-03 1.5E-03 50 0 3.6E-04 1.85

0.12 1.2E-04 1.5E-04 50 0 5.2E-05 0.22

0.12 7.0E-06 1.5E-03 50 0 4.3E-06 2.6E-02

0.12 7.0E-07 1.5E-04 50 0 4.3E-07 2.7E-03

0.12 1.2E-03 1.5E-03 120 0 2.1E-04 0.14

0.12 7.0E-06 1.5E-03 120 0 2.6E-06 3.2E-03

Table 67 Supplemental Reanalysis with SFPS Input Deck

Release Fraction Reanalysis with SFPS Input Deck

NG PART CRUD Release 

Height 

(m)

Prompt 

Fatality 

within 10 

mi

Ind. Risk 

of LCF 

within 10 

mi

Ind. Peak 

Dose at 

1.2-1.6 km 

(Sv)

Collective 

Dose

(0-50 mi)

Person-Sv

Interdicted land 

in first year after 

accident 

(square miles) 

Condemn

land (squ

miles 

0.12 1.2E-03 1.5E-03 50 0 7.1E-05 0.33 740 20 1.6

0.12 1.2E-04 1.5E-04 50 0 8.9E-06 4.0E-02 86 1.4 

0.12 7.E-06 1.5E-03 50 0 7.3E-07 6.6E-03 5.7 1.2E-02 

0.12 7.E-07 1.5E-04 50 0 7.5E-08 6.8E-04 0.57 4.1E-06 

0.12 1.2E-03 1.5E-03 120 0 5.1E-05 7.4E-02 780 24 3.9

0.12 7.E-06 1.5E-03 120 0 7.0E-07 1.7E-03 6.2 3.2E-04 

243
244

10.3.2 Quantitative Comparison of Dry Cask Storage and SFPS Analytical Results

The following table presents selected consequence results from the previous study of dry cask 

storage accidents, the supplemental dry cask storage study described above, and the SFPS.

Table 68 Comparison of consequence results from SFPS, NUREG-1864, and DCSS 

supplemental analyses

Metric SFPS Results NUREG-1864 DCSS Suppl. 

Analyses

Early fatalities 

(0 to 500 miles)

0 0 0

Individual LCF risk 

within 10 miles 

(conditional) 

2.0e-4 to 

4.4e-4

4.3e-7 to 

3.6e-4

7.5e-8 to 

7.1e-5

Collective dose

within 50 miles in 

Person-Sv

7,400 to 

39,000

Not reported 0.6 to 780

Collective dose

within 500 miles in 

Person-Sv

27,000 to 

350,000

Not reported Not reported

Interdicted land 

(square miles)

170 to 9,400 Not reported <<1 to 24

Condemned land 

(square miles)

<1 to 83 Not reported <<1

10.3.3 Comparison of SFPS Results to Previous and Supplemental Cask Studies

Comparison of SFPS results to past dry cask studies is not straight-forward. This is because 

the type of information reported is different, the assumptions related to fuel and canister/cask 

damage are different, and the risks of dry cask handling, while low, are generally driven by 

design features that can vary significantly between different DCSS designs. For example, the 

NUREG-1864 study is based on a welded canister-in-overpack design, whereas the site 

selected for the SFPS study uses directly loaded bolted casks. 

Nevertheless, meaningful comparisons can be made. An examination of the conditional 

individual latent cancer fatality probability metric demonstrates the effectiveness of emergency 

response and long-term protective actions at mitigating dose, consistent with the observations 

made in previous studies such as NUREG/CR-4982 and NUREG/CR-6451. The maximum 

consequences (in terms of latent cancer fatality probability) for both a pool accident and a dry 

cask accident, although involving substantially different amount of released material, are both 

limited to a range of 1E-4 to 1E-3 per event. The contrast to the much higher conditional 

consequence reported in NUREG-1738 (8.2E-2) is due to the assumption of a late evacuation 

coupled with a high source term in this study. The difference between impacts from pool and 

cask accidents is more clearly highlighted in measures related to the areal extent of 

contamination rather than in measures of peak individual risk. Inspection of Table 67 and Table 

68 demonstrates that even in the case of very high release fractions from dry cask accidents, 

conditional results for metrics such as population dose or condemned or interdicted lands are 
245

several orders of magnitude lower than the low end of consequences of pool accidents. This 

comparison is significantly exaggerated if a less conservative estimate of the DCSS release 

fraction is used. The results suggest that a DCSS accident is unlikely to result in the need for 

extensive offsite protective action such as land interdiction or population displacement, in 

contrast to a pool accident that may require significant offsite protective action. Furthermore, for 

the risks (expressed as a frequency-weighted consequence) of a DCSS accident to be 

comparable to the risks of a pool accident, the frequency of a DCSS accident would have to be 

several orders of magnitude higher than that of a pool accident.

10.4 Summary of Assessment of Previous Studies

This assessment demonstrates that past SFP accident consequence estimates from large 

seismic events are similar to this study for most metrics. Comparison of this study to dry cask 

storage studies (NUREG-1864 and supplemental analyses from this Chapter), indicates that in 

some circumstances, the conditional individual LCF risk within 0 to 10 miles would be similar 

due primarily to the conservative upper bound estimate of the dry cask release as well as the 

expected effectiveness of protective actions in response to an SFP release. However, 

conditional results for metrics such as population dose or condemned or interdicted lands are 

several orders of magnitude lower for dry cask scenarios than the low end of consequences of 

pool accidents, due to the substantially smaller amount of released material.
246
247

11. REGULATORY ANALYSIS SCREENING SUMMARY

Based on past studies, the NRC has concluded that both spent fuel pools and dry casks provide 

adequate protection of public health and safety and the environment, and that the likelihood of 

an accident involving a radiological release from the spent fuel remains extremely small. While 

the staff believes that public health and safety is adequately protected for both spent fuel pool 

and dry cask storage, the Spent Fuel Pool Study (SFPS) provides one part of a technical 

analysis to confirm, using insights from Fukushima, that spent fuel pools continue to provide 

adequate protection. As indicated by its title, this study looks at the storage of spent nuclear 

fuel in spent fuel pools. The study also assesses whether any significant safety benefits (or 

detriments) would occur from expedited transfer of spent fuel to dry casks, and the potential 

costs associated with such expedited transfer.

The study establishes that both high and low density spent fuel pool arrangements at the 

reference plant provide reasonable assurance of adequate protection. The analysis in Appendix 

D, which is summarized here, assesses the benefits and costs of this action relative to the 

baseline of existing requirements, including current regulations and relevant orders.

11.1 Decision Rationale

11.1.1 Comparison to the Safety Goal Policy Statement

The Safety Goals for the Operation of Nuclear Power Plants: Policy Statement (51 FR 28044) 

(safety goal policy statement) was used to evaluate the impacts resulting from a severe spent 

fuel pool accident. The frequency of damage to the spent fuel pool is estimated to be 

approximately between 7.11x10-7 and 5.39x10-6 per year when considering all initiators that 

could challenge spent fuel pool cooling or integrity. This value, when compared to a target core 

damage frequency value of 1x10-4 per reactor-year in the Safety Goal Policy Statement,

represents 0.71 to 5.39% percent of the overall frequency of core damage.

As described in Appendix D it is difficult to compare the estimated 7.11x10-7 to 5.39x10-6 per 

reactor-year release frequencies for the postulated spent fuel pool accident when considering all 

initiators to a target value of 1x10-5 per reactor year for a large early release frequency (LERF). 

The spent fuel pool source term is not similar to the core damage (or melt) source term. The 

consequences of a spent fuel pool accident are predicted to have no early fatalities and public 

health risk is dominated by latent cancer risks resulting from long-term exposures. Because the 

analyzed spent fuel accident is a slow progression with at least eight hours before an 

environmental release occurs and the resultant release is not expected to result in any offsite 

early fatalities, the analysis suggests that the spent fuel pool release does not fall within the 

definition of a large early release. Although this analyzed accident is different from a reactor 

accident, the spent fuel pool estimated release frequencies of 7.11x10-7 to 5.39x10-6 per 

reactor-year meet the 1x10-5 LERF guidelines.

Collective risk is based on the statistically expected number of early and latent cancer fatalities. 

The safety goal policy statement defines the early fatality area calculation as that within one 

mile from the site boundary. A ten-mile radius is defined for calculating latent cancer fatalities. 

The quantitative objective of the Policy Statement is for the risk to the population in the vicinity 

of a nuclear power plant from an accident at a nuclear power plant to not exceed 0.1 percent of 

the sum of cancer fatality risks resulting from all other causes. Based on recent data, the total 
248

fatality rate from cancer in the U.S. is 580,350 per 315,747,500 persons 

(http://www.census.gov/popclock/) or a risk of 1.84x10-3 per year, which results in a safety goal 

of 1.84x10-6 per year. Using the bounding frequency of damage to the spent fuel pool of 

5.39x10-6 per year, which considers all initiators that could challenge spent fuel pool cooling or 

integrity, and the conditional individual latent cancer fatality risk within a 10-mile radius of 

4.4x10-4 yields a latent cancer fatality risk of 2.37x10-9 per year. This calculated value of 

2.37x10-9 latent cancer fatalities per reactor-year associated with a spent fuel pool accident 

represents a 0.13 percent fraction of the 1.84x10-6 per year societal risk goal.

Therefore, the risk and consequences of a spent fuel pool accident at the reference plant meet 

the Safety Goal Policy Statement public health objectives. They also meet the 1x10-5 per 

reactor-year LERF guideline. Therefore, the NRC concludes that a regulatory requirement for 

expedited transfer of spent fuel from the spent fuel pool to storage casks is not needed for the 

reference plant In order to meet the Safety Goals.

11.1.2 Cost-Benefit Analysis

The key findings of the analysis are as follows:

• Total Cost to the Reference Plant. The proposal to expeditiously move older 

spent fuel assemblies from pool storage to dry cask storage beginning in year 

2014 to achieve and maintain a low-density loading in the pool within five years 

will result in an estimated present value cost of $46.81 million (using a 7-percent 

discount rate) and $42.10 million (using a 3-percent discount rate) over the next 

26 years. The earlier upfront and incremental dry storage cask capital and 

loading costs dominated these incremental costs. The reference plant routine 

occupational health costs will result in an estimated present value cost of 

$27,000 (using a 7 percent discount rate) and $24,000 (using a 3-percent 

discount rate). Sensitivity analyses result in an estimated present value cost that 

ranged from $16.4 million to $46.9 million.

• Value of Benefits to the Reference Plant. The benefits for expeditious movement 

of spent fuel to dry cask storage will result in an estimated present value benefit 

of $493,000 (using a 7-percent discount rate) and $711,000 (using a 3-percent 

discount rate). These benefits result from the monetized value for averted public 

and occupational radiation exposure and averted onsite impacts and offsite 

property damage. Sensitivity analyses result in an estimated present value 

benefit that ranged from $0.5 million to $27.7 million.

• Costs to NRC. The NRC costs to require the expeditious movement of spent fuel 

to dry cask storage were conservatively ignored to calculate the maximum 

potential benefit. Even though the NRC is not expected to incur substantial 

implementation or annual costs for this alternative, these costs would further 

reduce the calculated net benefit for the proposed expeditious movement of older 

spent fuel assemblies from pool storage to dry cask storage for the reference 

plant.

There are uncertainties in estimating the frequency of events for natural phenomena that are 

postulated to challenge spent fuel pool cooling or integrity. There are also uncertainties in the 

calculation of event consequences in terms of the dispersion and disposition of radioactive 

material into the site environs. This is due in part to uncertainties regarding the degree to which 
249

topographical features and other phenomena are modeled at distances away from the reference 

plant. Estimating economic consequences also includes large uncertainties, as it is difficult to 

model the impact of disruptions to many different aspects of local economies and the loss of 

infrastructure on the general U.S. economy. An example of this is the supply chain disruptions 

that followed the 2011 Tohoku earthquake and subsequent tsunami in Japan or the 2004 Indian 

Ocean earthquake and tsunami in Thailand.

The NRC recognizes that there are also costs and risks associated with the handling and 

movement of spent fuel casks in the reactor building. These impacts, if included in this analysis, 

would further reduce the overall net benefit in relation to the regulatory baseline. These effects 

(e.g., the added risks of handling and moving casks) were conservatively ignored in order to 

calculate the maximum potential benefit by only comparing the safety of high-density fuel pool 

storage relative to low-density fuel pool storage and its implementation costs without 

consideration of cask movement risk.

The release of fission products to the environment resulting from other events that cause the 

loss of spent fuel pool cooling or integrity (i.e., missiles, heavy load drops, loss of cooling or 

make-up, inadvertent drainage or siphoning and pneumatic seal failures) are estimated to occur 

approximately once in 2.7 million years or 3.7x10-7 per reactor-year. Operator diagnosis and 

recovery are important factors considered in the development of the event frequencies for these 

events and portions of this evaluation are premised on licensees having taken appropriate 

actions to understand the potential consequences of spent fuel pool accident events and 

develop appropriate procedures and mitigating strategies to respond and mitigate the 

consequences.

In section 9.2 of the SFPS, a sensitivity analysis is provided in which a more favorable fuel 

pattern is considered in which eight cold assemblies surround each hot assembly (i.e., 1x8 fuel 

assembly pattern). Although only a few sensitivity analyses were performed using this 

configuration, the results looked promising for inhibiting spent fuel pool releases. The sensitivity 

calculations for the high-density 1x8 fuel pattern showed a shorter time to air coolability (i.e. no 

releases in OCP3). Even for the cases that led to the release of radioactive materials in OCP2, 

the release magnitude was much smaller than for the 1x4 fuel pattern, and comparable to the 

low density cases. The fuel thermal response has a slower heatup when compared to a fuel 

pattern in which four cold assemblies surround each hot assembly (i.e., 1x4 fuel assembly 

pattern) because there is more mass to absorb heat. Furthermore, the loading configuration 

may result in similar reductions in risk to the low-density storage option evaluated without the 

significant capital costs for implementation. Further evaluation of this alternative and possibly 

other loading configurations for all operating cycle phases is recommended as part of the 

regulatory analysis for expedited fuel movement as part of the program plan described in 

SECY-12-0095 to evaluate the transfer of spent fuel to dry cask storage.

Sensitivity analyses that extend the analyses beyond 50 miles at the reference plant show that 

the low-density spent fuel storage alternative was not cost-justified for any of the discounted 

sensitivity cases, and was only marginally justified if discounting was not applied. Therefore, 

the expedited transfer of spent fuel from pools to dry cask storage containers at the reference 

plant does not meet the cost-justified substantial safety enhancement criterion.

11.2 Further Actions

The NRC plans to use the insights from this study along with other analyses to inform a broader 

regulatory analysis, which will help decisionmakers determine whether operating or future 
250

nuclear power reactor licensees should be required to maintain a low-density configuration in 

their spent fuel pools.

The analysis for the reference plant and the longer-term generic regulatory analysis address the 

questions of what can go wrong; how likely is it; and what are the consequences. Although this 

approach is well established at the NRC and other government agencies, it is often difficult to 

explain following rare disasters such as the accident at Fukushima Dai-ichi, or in presenting the 

results of studies such as this one. It is not enough to look at only the estimates of the low 

probabilities for failing spent fuel pools or only at the worst-case consequences in the unlikely 

event of failures of spent fuel pool integrity and existing mitigating systems. One needs to look 

at the totality of information presented in this report, previous studies, operating experience, and 

assess both the potential advantages and disadvantages of regulatory actions regarding the 

movement of spent fuel from storage pools to dry cask storage containers.
251

12. SUMMARY AND CONCLUSIONS

12.1 Summary

This study sought to investigate the relative consequences between low and high-density 

loading situations for a selected site following a seismic event greater than the maximum 

earthquake reasonably expected to occur at the reference plant location. The NRC expects that 

the ground motion used in this study is more challenging for the spent fuel pool structure than 

that experienced at the Fukushima Daiichi nuclear power plant from the earthquake that 

occurred off the coast of Japan on March 11, 2011. That earthquake did not result in any spent 

fuel pool leaks. Chapter 1 discussed some of the considerations that are raised by stakeholders 

with respect to these differences. These are re-visited here to set the stage for presenting the 

study’s findings.

• Expedited movement of fuel from the SFP to dry storage will decrease the inventory of 

longer-lived radionuclides such as cesium-137

OCP High density (MCi) Low density (MCi) Ratio (low/high)

OCP1 54 17 0.31

OCP2 59 22 0.37

OCP3 59 22 0.37

• As a result of the above, less radioactive material would be present if a radioactive 

release occurred, which would be expected to reduce potential health effects, potential 

land contamination, and economic impacts

This point is covered in the findings below.

• Removal of older fuel slightly reduces the overall heat load in the pool, which can have 

the effect of delaying the start of a radioactive release (and thus increasing the time 

available to take mitigative action) for many types of accidents

OCP High density (kW) Low density (kW) Ratio (low/high)

OCP1 2,951 2,526 0.86

OCP2 3,567 3,143 0.88

OCP3 2,571 2,149 0.84

• Removal of older fuel will increase the volume available for cooling water 

As mentioned before, this is mathematically a small effect with the older fuel comprising on the 

order of 5% of the total pool volume (recall that most of the pool is occupied by water, not fuel). 

In the scenarios studied here, a 5% difference in the initial water inventory generally would not

have affected the course of the accident and the offsite consequences.

The results of the study are as follows:

1. A beyond design basis event with a frequency of occurrence of 1 in 60,000 per year was 

used in this study, and more likely earthquakes are not expected to challenge the SFP 

structure.
252

2. Past studies have indicated that large seismic events could lead to the loss of structural 

integrity of the spent fuel pool liner. This study’s results confirm that such a condition is 

unlikely. For the low probability seismic event described above, the study estimated a 

conditional probability of failure of 0.1. The specific conditions under which a failure 

might occur are site-specific.

3. NUREG-1353 (1989) predicted generic seismically-induced SFP liner failure likelihoods 

of 2x10-6 to 6x10-6 per year, generally associated with events greater than 0.5g peak 

ground acceleration. NUREG-1738 (2001) predicted generic seismically-induced SFP 

liner failure likelihoods of 2x10-7 to 2x10-6 per year, generally associated with events 

around 1.2 g. The current study looks at a seismic event in the range of 0.5 to 1 g, and 

estimates a site-specific SFP liner failure likelihood of 2x10-6 per year (based on the 

informed expectation that this seismic range has the greatest contribution to frequencyweighted consequences). Since the updated initiating event frequency estimate (based

on the 2008 U.S. Geological Survey model) for the reference plant for events greater 

than 1 g is 6x10-6 per year, this portion of the seismic hazard (i.e., > 1 g) may contribute 

more significantly to the overall frequency-weighted consequences for the reference 

plant than previously anticipated, depending on the conditional structural SFP liner 

failure probability associated with these larger events. The effect of this scope limitation 

may be offset by potential conservatisms in the structural analysis described in Section 4 

of this report.

4. In this study, no set of conditions short of a liner failure led to a radiological release in 

less than 3 days, which is consistent with past studies. In most cases, the available time 

to prevent a radiological release was much greater than 3 days.

5. In this study, without mitigative action, fuel is estimated to be air coolable for at least 72 

hours for all but roughly 10% of the operating cycle45. Past studies estimated this time 

to be a greater fraction of the operating cycle, when hotter fuel was contiguously stored.

In other words, use of the 1x4 pattern has a positive effect in promoting natural 

circulation air coolability and reducing the likelihood of a release should the SFP become 

completely drained. An even shorter time was predicted for the 1x8 pattern currently 

employed at PBAPS. While variability in SFP loading configurations was not a focus of 

this study, this report consistently shows the advantages associated with dispersed fuel 

loading patterns.

6. In the cases studied, which in general did not account for multiple or concurrent reactor 

and SFP accidents, the precise time to diagnose the need for SFP mitigation did not 

have an effect on the course of most scenarios.

Nevertheless, the improved reliable and available SFP indication required by the 

NRC Order of March 12, 2012 (EA-12-051) is important to ensure that plant 

personnel can effectively prioritize emergency actions. The availability of such 

instrumentation may have changed the mitigation mode (makeup versus sprays) 

deployed to mitigate events that resulted in a release.

 45 The actual time is between 37 days (not air coolable) and 107 days (air coolable), with 60 days representing 

the demarcation point between these two Operating Cycle Phases. The citation of 60 days as a 

representative value is reasonable based on other separate effects analyses not documented in this report. 

The actual time to air coolability could be more or less, depending on specific conditions.
253

7. This study considered variations in both pool loading and the effective deployment (or 

lack thereof) of 10 CFR 50.54(hh)(2) mitigation capabilities (i.e., water makeup or spray 

using portable equipment). Of these, effective deployment of mitigation had the largest 

impact on preventing a release of radioactive material, reducing the release frequency 

by a factor of about twenty (from 1x10-7

/yr to 6x10-9/yr).

Note that ongoing regulatory actions under Order EA-12-049 dated March 12, 

2012 (and related correction dated March 13, 2012) increase the capability of 

operating nuclear power plants to mitigate beyond-design-basis external events, 

such as the seismic event studied here. 

8. The difference between high-density and low-density loading situations were as follows:

• In terms of the likelihood of release within 3 days, no difference was seen.

• In terms of consequences, the low density cases resulted in a smaller release 

due to the smaller inventory of radioactive material and the lower potential for 

hydrogen combustion. For high-density loading, the rapid draindown cases in 

general had smaller releases mainly because the reactor building remained intact 

(hydrogen combustions not predicted). For slow draindown events, longer times 

are available for deployment of mitigation. Without successful deployment of 

mitigation, the releases could be up to two orders of magnitude larger (these 

cases are associated with hydrogen combustion events).

9. For all scenarios, no offsite early fatalities attributable to acute radiation exposure are 

predicted to occur. Due to radioactive decay, spent fuel pools tend to have significantly 

less shorter-lived radionuclides (e.g. I-131) than reactors. Partly because of this, the 

release is not predicted to be fast and large enough to significantly exceed offsite dose 

levels necessary to induce early fatalities. . When necessary, emergency response as 

treated in this study effectively prevents early fatalities from acute radiation exposure. 

10. In both high and low density loading without successful deployment of mitigation, the 

individual latent cancer fatality risk within 10 miles for the studied scenarios is predicted 

to be on the order of 10-10 to 10-11 per year, based on the linear no threshold dose 

response model. While this risk is scenario-specific and related to a single spent fuel 

pool, it is several orders of magnitude lower than the 2e-6 per year individual latent 

cancer fatality risk corresponding to the quantitative health objective for latent cancer 

fatalities and therefore unlikely to contribute significantly to a risk that would challenge 

the Commission’s safety goal policy (NRC 1986). In addition, there is uncertainty in the 

risk calculations because it is dominated by low doses. As a perspective on uncertainty, 

excluding the uncertain effects of low doses significantly reduced the quantified 

individual latent cancer fatality risk within 10 miles. Average individual latent cancer 

fatality risk is low because of low release frequencies and the expected protective 

actions.

11. Average individual latent cancer fatality risk is low and decreases slowly as a function of 

distance from the plant. For scenarios with large releases, significant collective doses 

are estimated; however, risk of cancer fatalities from these doses would be a small 

fraction of the risk of cancer fatalities from all causes. Additionally, these individual risks 

are dominated by long-term exposures to very lightly contaminated areas for which 

doses are small enough to be considered habitable. In comparing pool configurations, 
254

collective dose (and latent cancer fatalities) for the studied scenarios could be an order 

of magnitude higher for the high density loading situation as compared to the low density 

loading situation 

12. The amount of land interdiction for the studied scenarios could be up to two orders of 

magnitude greater for certain high density loading situations as compared to the low 

density loading situations. Also, like releases in the low density loading situation, 

successfully deployed mitigation in the high density loading situation is predicted to 

reduce the amount of land interdiction to a similar extent. For both situations, the major 

difference is driven by hydrogen combustion events and associated large releases, 

which are only predicted to occur in scenarios with unsuccessful deployment of 

mitigation.

13. While the likelihood of release is very low, offsite protective measures in the form of 

population relocation and land interdication may be extensive. High-density loading 

releases without 10 CFR 50.54(hh)(2) mitigation measures are calculated to result in 

release frequency-weighted land interdiction values of 0.001 mi2 per year and 0.5 

displaced individuals per year which are arrived at by multiplying the estimated 

frequency and the estimated consequence. While the amount of land interdiction can be 

large, the fraction expected to be permanently interdicted is small if a release were to 

occur. For low-density loading or with successful deployment of 10 CFR 50.54(hh)(2) 

mitigation measures, considerably less land interdiction and displaced individuals are 

predicted. 

14. A comparison of the risks of different fuel handling strategies, such as current practice 

and expedited transfer, depends on several factors including the relative, site-specific 

risks, and the time spent in each stage of spent fuel storage. Other risks, such as the 

risk from cask drop events damaging fuel in the cask or the SFP, may at least partially 

offset the benefit of lower spent fuel pool risk from low density loading.

15. The human reliability study shows that in most situations SFP mitigation can be 

deployed in time to prevent release given the assumptions that sufficient plant staff and 

equipment is available for SFP mitigation and the work area is accessible to perform 

mitigation. There are two exceptions where mitigation will be ineffective under the 

moderate leak scenarios: (1) the earthquake occurs at the beginning of a refueling 

outage when the spent fuel is too hot for the assumed mitigation; and (2) the earthquake 

occurs when spent fuel is relatively hot and the reactor and spent fuel pool are 

hydraulically disconnected resulting in insufficient time to deploy mitigation and natural 

cooling mechanisms cannot prevent fuel damage. This study identified that possible 

improvements in mitigation flow and nozzle placement in low-dose locations could 

improve mitigation success likelihood, but this would require further verification.

16. This study demonstrates that past SFP risk estimates from large seismic events are 

similar to this study for most consequence metrics (see Chapter 10). Comparison of this 

study to dry cask storage studies (NUREG-1864 and supplemental analyses from 

Chapter 10) indicates that in some circumstances, the conditional individual LCF risk 

within 10 miles would be similar due primarily to the conservative upper bound estimate 

of the dry cask release as well as the expected effectiveness of protective actions in 

response to an SFP release. However, conditional results for metrics such as temporary 

or permanently interdicted land or population dose are several orders of magnitude 
255

lower for dry cask scenarios than the low end of consequences of pool accidents, due to 

the substantially smaller amount of released material.

17. Applying the NRC’s regulatory analysis guidelines to analyse the results of the SFP 

Study indicates that requiring the low-density spent fuel pool storage alternative is not 

justified for the reference plant. This conclusion is subject to the analysis model, data, 

inputs and assumptions in Section D.3 of Appendix D. The risk due to beyond design 

basis accidents in the spent fuel pool analyzed in this study is sufficiently low that the 

added costs involved with expediting the movement of spent fuel from the pool to 

achieve the low-density fuel pool storage alternative are not warranted. Sensitivity 

analyses that extend the analyses beyond the primary area considered also show that 

the low-density spent fuel storage alternative was not cost justified for any of the 

discounted sensitivity cases.

12.2 Conclusions

In conclusion, past SFP risk studies have shown that storage of spent fuel in a high-density 

configuration is safe and risk is low. This study is consistent with earlier research conclusions 

that spent fuel pools are robust structures that are likely to withstand severe earthquakes 

without leaking. The study estimated that the likelihood of a radiological release from the spent 

fuel pool resulting from the selected severe seismic event analyzed in this study is on the order 

of one time in 10 million years or lower. The factors leading to this low likelihood, as discussed 

in Section 12.1, are summarized in Figure 138.

Figure 138 Affecting Likelihood of SFP Release from a Severe Seismic Event

Event

For the hypothetical releases studied, no early fatalities attributable to acute radiation exposure 

were predicted and individual latent cancer fatality risks are projected to be low, but extensive 
256

protective actions may be needed. Comparisons of the calculated individual latent cancer 

fatality (LCF) risk within 10 miles to the NRC Safety Goal are provided in Figure 139 to give 

context that may help the reader to understand the contribution to cancer risks from the accident 

scenarios that were studied. The NRC Safety Goal for latent cancer fatality risk from nuclear 

power plant operation (i.e., 2x10-6 or two in one million per year) is set 1,000 times lower than 

the sum of cancer fatality risks resulting from all other causes (i.e., ~2x10-3 or two in one 

thousand per year). Comparing the study results to the NRC Safety Goal does involve 

important limitations. First, the safety goal is intended to encompass all accident scenarios on a 

nuclear power plant site, including both reactors and spent fuel. This study does not examine 

all scenarios that would need to be considered in a probabilistic risk assessment for a spent fuel 

pool, although seismic contributors are considered the most important contributors to spent fuel 

pool risk. Also, this study represents a mix of limited probabilistic considerations with a 

deterministic treatment of mitigating features. All analytical techniques, both deterministic and 

probabilistic, have inherent limitations of scope and method and also have uncertainty of varying 

degrees and types. As a result, comparison of the scenario-specific calculated individual LCF 

risk to the NRC Safety Goal is incomplete. However, it is intended to show how multiple spent 

fuel pool scenarios’ risk results in the one in a trillion (10-12) to one in 10 billion (10-10) per year 

LCF range) are low. While the results of this study are scenario-specific and related to a single 

spent fuel pool, staff concludes that since these risks are several orders of magnitude smaller

than the 2x10-6 (two in one million) individual LCF risk that corresponds to the safety goal for 

latent cancer fatalities, it is unlikely that the results here would contribute significantly to a risk 

that would challenge the Commission’s safety goal policy (NRC, 1986).

Figure 139 Comparison of Population-Weighted Average Individual Latent Cancer Fatality Risk Results for 

this Study to the NRC Safety Goal (plotted on logarithmic scale)
257

The study results demonstrated that in a high-density loading configuration, a more favorable 

fuel pattern or successful mitigation generally prevented or reduced the size of potential 

releases. Low-density loading reduced the size of potential releases but did not affect the 

likelihood of a release. When a release is predicted to occur, individual early and latent fatality 

risks for individuals within 10 miles do not vary significantly between the scenarios studied 

because protective actions, including relocation of the public and land interdiction, were 

modeled to be effective in limiting exposure. The beneficial effects in the reduction of offsite 

consequences between a high-density loading scenario and a low-density loading scenario are 

primarily associated with the reduction in the potential extent of land contamination and 

associated protective actions. 

The results of the SFP Study show that the overall level of safety with respect to spent fuel 

storage in a spent fuel pool currently achieved at the reference plant is high and that the level of 

risk at the reference plant is very low. Applying the NRC’s regulatory analysis guidelines to 

analyze the results of the SFP Study with respect to the quantitative benefits attributable to 

expedited transfer of spent fuel at the reference plant, and the risk reduction attributable to 

expedited transfer against the NRC’s Safety Goals, the NRC concludes the incremental safety 

(including risk) reduction associated with expedited transfer of spent fuel at the reference plant 

is not warranted in light of the added costs involved with expediting the movement of spent fuel 

from the pool to achieve low-density fuel pool storage. Therefore, an NRC requirement 

mandating expedited transfer of spent fuel from pools to dry cask storage containers at the 

reference plant does not appear to be justified. The NRC plans to use the insights from this 

analysis to inform a broader regulatory analysis of the spent fuel pools at US nuclear reactors. 
258
259

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A-1

APPENDIX A: DETAILED EMERGENCY RESPONSE MODELS

The detailed evacuation timing and speeds for each cohort developed using the information and 

approach described in Section 7.1.4 are described in this appendix. Selected input parameters 

for WinMACCS are described below:

• Delay to shelter (DLTSHL) represents a delay from the time of the start of the accident 

until cohorts enter the shelter. 

• Delay to evacuation (DLTEVA) represents the length of the sheltering period from the 

time a cohort enters the shelter until the point at which it begins to evacuate. 

• The speed (ESPEED) is assigned for each of the three phases used in WinMACCS,

which are the beginning, middle, and late phases. Average evacuation speeds were 

derived from the reference plant’s ETE report. Speed adjustment factors are used in the 

WinMACCS application to represent free flow in rural areas and congested flow in urban 

areas.

• Duration of beginning phase (DURBEG) is the duration assigned to the beginning phase 

of the evacuation and may be assigned uniquely for each cohort.

• Duration of middle phase (DURMID) is the duration assigned to the middle phase of the 

evacuation and may also be assigned uniquely for each cohort. The remainder of the 

evacuation, following period defined by DURMID, is the late phase.

A.1 Evacuation Model 1: WinMACCS response parameters for sequences 

where PAGs are not exceeded beyond the EPZ.

The following cohorts were established for this evacuation model: 

0 to 10 Miles, Early Evacuees: This population begins to evacuate before receiving an 

evacuation order. Focus group work conducted to support NUREG/CR-6953, Volume 2 (NRC, 

2008c) suggested that some residents are prepared and ready to evacuate at the first indication 

of an accident at the nuclear power plant. Results of the telephone survey conducted with 

NUREG/CR-6953 showed that on a national level, 20 percent of residents of EPZs have packed 

a ”go-bag” and are ready to leave. Because the accident is initiated by a severe earthquake, it 

is assumed 30 percent of the public evacuate. 

10 to 20 Miles, Shadow: These residents evacuate from areas that are not under an official 

evacuation order. The distribution of the shadow evacuation would likely include a larger 

percentage of the public near the boundary of the EPZ, and the percent would decrease 

proportional to the distance away from the EPZ. For this analysis, a uniform fraction of the 

population is assumed to evacuate within the 10- to 20-mile region. Because the accident is 

initiated by a severe earthquake, it is assumed 30 percent of the public evacuate. This cohort 

will begin evacuating as they hear of the evacuation orders and observe EPZ evacuees 

traveling through the area. 

0 to 10 Miles, Public: This population group evacuates over a period of time, with some

residents leaving promptly and others leaving later. For this analysis, the bulk of this group is 
A-2

modeled as a single cohort, while the rest of the group is captured as different cohorts, such as 

the tail.

0 to 10 Miles, Special Facilities: This is a small but unique population group within the EPZ. 

There is no delay to shelter because these residents are assumed to be in a robust facility when 

the accident begins. Specialized vehicles to evacuate these facilities take time to mobilize. 

0 to 10 Miles, Tail: The tail represents the last 10 percent of the EPZ population who typically 

take a longer time to begin to evacuate. 

0 to 10 Miles, Schools: This cohort includes elementary, middle, and high school student 

populations within the EPZ. Schools receive early and direct warning from OROs and have 

response plans in place to support busing of students out of the EPZ. 

Nonevacuating Public: A portion of the public does not follow protective action orders. It is 

assumed that 0.5 percent of the general public within the EPZ refuse to evacuate. 

Table 69 Evacuation Model 1: EPZ Evacuation

Population Response Delays (hours) Phase Duration (hr) Evacuation Travel Speeds (mph)

Cohort Population 

Fraction

Siren 

(OALAR

M)

Delay to 

Shelter

Delay to 

Evacuation

Total 

(Depart 

time)

Early 

(DURB

EG)

Middle 

(DURMID)

Early 

(ESPEED)

Middle 

(ESPEED)

Late

(ESPEED)

1

0 to 10 miles 

Early Evacuees 

0.3 1

0 0 1

1 0.5 20 15 5

10 to 20 miles 

Shadow 2 1 4

2 0 to 10 miles

General Public 0.417 1 1 1 3 0.25 3 5 2 20

3

0 to 10 miles 

Special 

Facilities 

0.006 1 0 4 5 0.5 0.5 2 15 20

4 0 to 10 miles

Evacuation Tail 0.1 1 2 3 6 0.5 0.5 2 15 20

5 0 to 10 miles

Schools 0.172 1 0 0.5 1.5 1 0.5 20 15 20

6

0 to 10 miles

Nonevacuating 

Public

0.005 1 - - - - - - - -

For this sequence, hotspot relocation is 5 rem at 4 hours and normal relocation is 1 rem at 8 

hours. The values were established specific for this evacuation model developed for sequences 

with relatively small releases.
A-3

A.2 Evacuation Model 2: WinMACCS response parameters for late release 

sequences where the PAG is exceeded beyond the EPZ.

Preliminary results suggest that emergency-phase doses of 1 rem may extend 30 to 40 miles 

from the plant for some of the larger postulated releases. The EPA PAG suggests evacuation 

to these distances. In this analysis, it is assumed that evacuation to 30 miles is completed and 

SIP is implemented in the 30- to 40-mile area, which reduces the dose to the public below the 

PAG.

The population within a 30-mile radius of the reference plant is approximately 1.4 million. The 

population within the 40-mile radius is approximately 3.4 million. Because of larger populations 

at longer distances, it is important to better understand the potential directions that the plume 

would travel. The reference plant’s wind rose in the figure below suggests that the predominant 

wind direction is to the south and east, which is generally toward lower population areas. A 

secondary direction in terms of likelihood is to the northwest to north. This region is also low in 

population. Thus, if a release were to occur, it is more likely that a relatively small population 

would be affected than if the release occurred at a facility near a major city. 

Figure 140 The Reference Plant’s Wind Rose

It is assumed in this evacuation model that ORO’s begin to order evacuations beyond the EPZ

24 hours after the start of the accident. This is based on preliminary results that indicates a 

large release beginning at 48 hours. For this sequence, the population within a 30-mile radius is 

evacuated after the EPZ has evacuated. The overall evacuation would be implemented as a 

staged evacuation, which is common for plume-related emergency response. In addition, a SIP 

is assumed to be ordered for the 30- to 40-mile radius area. 

To develop an ETE and corresponding speeds for the areas beyond the EPZ, it was assumed 

90 percent of the general public who reside between 10- and 30-miles from the plant can be 

evacuated 24 hours after ordered to evacuate. This is consistent with the lengthy travel times

observed in hurricane evacuations of similar populations. The last 10 percent (evacuation tail) 

is estimated to take an additional 12 hours. Because of the lengthy time for this release to the 

atmosphere, this evacuation model effectively includes two separate evacuations, the first being 

within EPZ followed later by the 10- to 30-mile area.

 

0.00%

2.00%

4.00%

6.00%

8.00%

10.00%

12.00%

N

NNE

NE

ENE

E

ESE

SE

SSE

S

SSW

SW

WSW

W

WNW

NW

NNW
A-4

The following cohorts were established for this evacuation model: 

0 to 10 Miles, Schools: This cohort includes elementary, middle, and high school student 

populations within the EPZ. Schools receive early and direct warning from OROs and have 

response plans in place to support busing of students out of the EPZ. 

0 to 10 Miles, Early Evacuees: This population begins to evacuate before receiving an 

evacuation order. Focus group work suggested that some residents are prepared and ready to

evacuate at the first indication of an accident at the nuclear power plant (NRC, 2008c).

Because the accident is initiated by a severe earthquake, it is assumed 30 percent of the public 

evacuate.

0 to 10 Miles, Public: This population group would evacuate over a period of time, with some 

residents leaving promptly and others leaving later. For this analysis, the bulk of this group is 

modeled as a single cohort, while the rest of the group is captured as different cohorts, such as 

the tail.

10 to 20 Miles, Shadow: These residents evacuate from areas that are not under an official 

evacuation order. The distribution of the shadow evacuation would likely include a larger 

percentage of the public near the boundary of the EPZ, and the percent would decrease 

proportional to the distance away from the EPZ. For this analysis, it is assumed that 30 percent 

of the general public from the 10 to 20 mile area shadow evacuate. For simplicity, this cohort is 

assumed to be distributed uniformly over the 10- to 20-mile area. This cohort begins evacuating 

as they observe EPZ evacuees traveling through the area.

0 to 10 Miles, Special Facilities: This is a small but unique population group within this EPZ. 

There is no delay to shelter because these residents are assumed to be in a robust facility when 

the accident begins. Specialized vehicles to evacuate these facilities take time to mobilize. 

0 to 10 Miles, Tail: The tail represents the last 10 percent of the EPZ population who typically 

take a longer time to begin to evacuate. 

10 to 30 Miles, Public: This population group evacuates over a period of time, with some 

residents leaving promptly and others leaving later. For this analysis, the bulk of this group is 

modeled as a single cohort that enters the roadway network while EPZ evacuees are travelling 

through. The rest of the group is captured as different cohorts, such as the tail. 

10 to 30 Miles, Special Facilities: Special vehicles needed to evacuate these facilities require 

additional time to mobilize and support the evacuation. 

30 to 40 Miles, Shadow: A shadow evacuation may be expected in the area beyond the 

evacuation area. For this analysis, it is assumed that 20 percent of the general public from the 

30- to 40-mile area evacuate. This cohort begins evacuating as they hear the order to evacuate

for the 10- to 30-mile area, or observe evacuees traveling through the area.

10 to 30 Miles, Tail: The tail represents the last 10 percent of the population of this area who 

typically take a longer time to begin to evacuate.

30 to 40 Miles, Shelter in Place (SIP): For this evacuation model, it is assumed that 80 percent

of the public remaining after the shadow evacuation complies with the SIP order.
A-5

Nonevacuating Public: A small portion of the public does not follow protective action orders. It 

is assumed that 0.5 percent of the general public within the 0 to 40 mile area refuse to 

evacuate. This group, however, is subject to relocation and a portion of this cohort is relocated 

according to the relocation parameters discussed above.

Table 70 Evacuation Model 2: Evacuation for PAGs exceeded beyond the EPZ (SFP 

release after 40 hours.)

Population Response Delays (hours) Phase Duration (hr) Evacuation Travel Speeds (mph)

Cohort Population 

Fraction

Delay to 

Shelter*

Delay to 

Evacuation

Total 

(Depart 

time)

Early 

(DURBEG)

Middle 

(DURMID)

Early 

(ESPEED)

Middle 

(ESPEED)

Late

(ESPEED)

1

0 to 10 miles 

Schools .172 0.25 1.25 1.5 0.25 2 20 15 20

2

0 to 10 miles 

Early Evacuees .2 0.5 0.5 1 1 2 20 10 20

3

0 to 10 miles 

General Public .517 1 2 3 0.25 3 5 2 20

4

10 to 20 miles 

Shadow .3 2 2 4 0.25 6 20 15 20

5

0 to 10 miles 

Special 

Facilities .006 0 5 5 3 2 2 5 20

6

0 to 10 miles 

Evacuation Tail .1 3 3 6 2 2 2 5 20

7

10 to 20 miles 

General Public .552

24 4 28 2 18 2 1 20

20 to 30 miles 

General Public .852

8

10 to 30 miles 

Special 

Facilities .043 15 15 30 1 10 1 1 20

9

30 to 40 miles 

Shadow .2 24 8 32 1 6 15 5 20

10

10 to 30 miles 

Evacuation Tail .1 24 16 40 10 2 1 10 20

11

30 to 40 miles 

Shelter in Place .795 NA NA NA NA NA NA NA NA

12

0-40 miles 

Non-evacuating 

Public .005 NA NA NA NA NA NA NA NA

*Delay to shelter is from the start of the accident (i.e. OALARM set to zero) 

For this sequence, hotspot relocation is 5 rem at 4 hours and normal relocation is 1 rem at 16 

hours. The releases for these sequences do not begin until about 40 hours or thereafter, and 

hotspot relocation does not begin until 4 hours after the plume reaches the location. OROs 

would be able to assemble considerable resources to monitor radiological conditions and could 

be expected to relocate people relatively rapidly should it be necessary. 
A-6

A.3 Evacuation Model 3: WinMACCS response parameters for early release 

sequences where the PAG is exceeded beyond the EPZ. 

This evacuation is similar to Evacuation Model 2. Preliminary results suggest that certain 

sequences that have large releases that begin between 8 and 18 hours and are capable of 

emergency-phase doses that exceed the PAGs beyond the EPZ. It is expected that dose 

projections would indicate protective actions beyond the EPZ are necessary. It is assumed that 

evacuation of the area beyond the EPZ would begin at 10 hours after the start of the accident.

Because the evacuation of the 10- to 30-mile area begins at 10 hours, this response is typical of 

a staged evacuation that would be employed in the case of a chemical release. The EPZ 

evacuation impacts the evacuation speeds of the 10- to 30-mile area. The following cohorts 

were established for this evacuation model: 

0 to 10 Miles, Schools: This cohort includes elementary, middle, and high school student 

populations within the EPZ. Schools receive early and direct warning from OROs and have 

response plans in place to support busing of students out of the EPZ. 

0 to 10 Miles, Early Evacuees: This population begins to evacuate before receiving an 

evacuation order. Focus group work suggested that some residents are prepared and ready to 

evacuate at the first indication of an accident at the nuclear power plant (NRC, 2008c).

Because the accident is initiated by a severe earthquake, it is assumed 30 percent of the public 

evacuate.

0 to 10 Miles, Public: This population group would evacuate over a period of time, with some 

residents leaving promptly and others leaving later. For this analysis, the bulk of this group is 

modeled as a single cohort, while the rest of the group is captured as different cohorts, such as 

the tail.

10 to 20 Miles, Shadow: These residents evacuate from areas that are not under an official 

evacuation order. The distribution of the shadow evacuation would likely include a larger 

percentage of the public near the boundary of the EPZ, and the percent would decrease 

proportional to the distance away from the EPZ. For this analysis, it is assumed that 30 percent 

of the general public from the 10 to 20 mile area shadow evacuate. For simplicity, this cohort is 

assumed to be distributed uniformly over the 10- to 20-mile area. This cohort begins evacuating 

as they observe EPZ evacuees traveling through the area. 

0 to 10 Miles, Special Facilities: This is a small but unique population group within this EPZ. 

There is no delay to shelter because these residents are assumed to be in a robust facility when 

the accident begins. Specialized vehicles to evacuate these facilities take time to mobilize. 

0 to 10 Miles, Tail: The tail represents the last 10 percent of the EPZ population who typically 

take a longer time to begin to evacuate. 

10 to 30 Miles, Public: This population group evacuates over a period of time, with some 

residents leaving promptly and others leaving later. For this analysis, the bulk of this group is 

modeled as a single cohort that enters the roadway network while EPZ evacuees are travelling 

through. The rest of the group is captured as different cohorts, such as the tail. 

10 to 30 Miles, Special Facilities: Special vehicles needed to evacuate these facilities require 

additional time to mobilize and support the evacuation. 
A-7

30 to 40 Miles, Shadow: A shadow evacuation may be expected in the area beyond the 

evacuation area. For this analysis, it is assumed that 20 percent of the general public from the 

30- to 40-mile area evacuate. This cohort begins evacuating as they hear the order to evacuate

for the 10- to 30-mile area, or observe evacuees traveling through the area.

10 to 30 Miles, Tail: The tail represents the last 10 percent of the population of this area who 

typically take a longer time to begin to evacuate.

30 to 40 Miles, Shelter in Place (SIP): For this evacuation model, it is assumed that 80 percent

of the public remaining after the shadow evacuation complies with the SIP order.

Nonevacuating Public: A small portion of the public does not follow protective action orders. It 

is assumed that 0.5 percent of the general public within the 0 to 40 mile area refuse to 

evacuate. This group, however, is subject to relocation and a portion of this cohort is relocated 

according to the relocation parameters discussed above.

Table 71 Evacuation Model 3: Evacuation for PAGs exceeded beyond the EPZ (SFP 

release after 8 hours).

Population Response Delays (hours) Phase Duration (hr) Evacuation Travel Speeds (mph)

Cohort Population 

Fraction

Delay to 

Shelter*

Delay to 

Evacuation

Total 

(Depart 

time)

Early 

(DURBEG)

Middle 

(DURMID)

Early 

(ESPEED)

Middle 

(ESPEED)

Late

(ESPEED)

1

0 to 10 miles 

Schools .172 0.25 1.25 1.5 0.25 2 20 15 20

2

0 to 10 miles 

Early Evacuees .2 0.5 0.5 1 1 2 20 15 20

3

0 to 10 miles 

General Public .517 1 2 3 0.25 3 5 2 20

4

10 to 20 miles 

Shadow .3 2 2 4 0.25 6 20 15 20

5

0 to 10 miles 

Special 

Facilities .006 0 5 5 3 2 2 5 20

6

0 to 10 miles 

Evacuation Tail .1 3 3 6 2 2 2 5 20

7

10 to 20 miles 

General Public .552

6 4 10 2 18 2 1 20

20 to 30 miles 

General Public .852

8

10 to 30 miles 

Special 

Facilities .043 0 20 20 1 10 1 1 20

9

30 to 40 miles 

Shadow .2 6 6 12 1 6 15 5 20

10

10 to 30 miles 

Evacuation Tail .1 10 20 30 10 2 1 10 20

11

30 to 40 miles 

Shelter in Place .795 NA NA NA NA NA NA NA NA

12

0-40 miles 

Non-evacuating 

Public .005 NA NA NA NA NA NA NA NA

*Delay to shelter is from the start of the accident (i.e. OALARM set to zero) 

For this evacuation model, the hotspot relocation is 5 rem at 26 hours and normal relocation is 1 

rem at 38 hours. The plume’s initial release for these sequences begins between 8 and 18 

hours after accident initiation. The assumed inability of the licensee to halt this release is the 

basis for expecting decision makers to expand the evacuation. Sequences that develop rapidly
A-8

could challenge ORO resources to assess radiological conditions beyond the evacuated areas 

and delay relocation of affected people. In this evacuation model, hotspot relocation does not 

begin until 26 hours after the release arrives, in order to account for the relatively earlier release 

and the evacuation of the public from the 10-30 mile area (which is expected to take as long as 

24 hours). 
B-1

APPENDIX B: A QUALITATIVE RISK COMPARISON OF SPENT FUEL 

STORAGE STRATEGIES

B.1 Introduction

In Staff Requirements Memorandum (SRM) M120607C, dated July 16, 2012, the Commission

directed the staff to conduct a comparative assessment of the results of the Spent Fuel Pool 

Study (SFPS) against previous studies of the safety consequences associated with loading, 

transfer, and long-term storage in dry cask storage systems (DCSS). Since the SFPS only 

includes a consequence study of certain seismic events, it is necessary to create a step-by-step 

model that can be used to compare safety consequences associated with the various stages of 

onsite spent fuel management. As part of the response to this SRM, this analysis (1) defines 

several fuel storage strategies to be compared, (2) develops a structure for calculating the 

difference in risks between these strategies, (3) identifies what relevant information exists, and 

(4) identifies what new information may be needed.

B.2 Spent Fuel Storage Strategies

For the purpose of studying this issue, two distinct spent fuel storage strategies commonly 

considered are defined: (1) current practice and (2) expedited transfer of spent fuel into dry 

storage. A large amount of variation exists in current spent fuel storage practices at various 

sites. Expedited transfer strategies, if implemented, would also be expected to vary 

considerably from site to site. Rather than attempting to bound all of the practices that are or 

may be implemented at various sites, this appendix will focus on the key elements of spent fuel 

storage strategies covered by existing risk analyses. Current practice generally consists of 

loading casks only when the pool, in a high density configuration, is nearly full. Just enough 

casks are loaded to maintain the capability to unload one full core into the pool. Expedited 

transfer of spent fuel into dry storage involves loading casks at a faster rate for a period of time 

to achieve a low density configuration in the spent fuel pool (SFP). The expedited process 

maintains a low density pool by moving all fuel cooled longer than 5 years out of the pool.

B.3 Spent Fuel Storage Stages

The risks associated with spent fuel storage will vary throughout the lifetime of a plant site and 

will depend on how the fuel is stored, and in what quantities. To analyze the lifecycle risk of 

spent fuel storage at a plant site, this appendix defines five fuel storage stages, beginning with a 

low-density pool approaching high density and ending with the final core offload being loaded 

into casks. The current practice strategy will not include the expedited transfer stage defined 

below.

Stage 1 consists of the fuel being offloaded into a low-density pool that eventually reaches high 

density. It is assumed that no casks are loaded during this stage as is generally industry 

practice. Stage 2 is when the pool is full in a high-density configuration and only as many casks 

are loaded as necessary. Stage 3 is during the expedited transfer period when the amount of 

cask loading is increased so as to decrease the inventory in the SFP. Stage 4 begins when 

expedited loading has been completed and the pool has returned to a low-density configuration 

and a lower rate of cask loading. Stage 5 begins when the reactor is permanently shut down 

and the last core is offloaded to the SFP. Stage 5 ends when all fuel has been placed in dry 

cask storage.
B-2

B.4 Risk of Spent Fuel Storage

This section presents generalized equations for the risk of spent fuel storage. These equations 

will serve as a guide to a subsequent discussion of the relative risk between storage stages and 

what drives the changes in risk.

The total annual risk of storing spent fuel during any stage can be expressed as the sum of the 

risk from the SFP and the dry casks. This can be expressed as,

R = Rcasks + Rsfp

where: R = annual risk of spent fuel

Rcasks = annual risk of loading and storing fuel in dry casks

Rsfp = annual risk of the spent fuel pool

The risk from loading and storing each dry cask is assumed to be constant and only dependent 

on the number of casks loaded or stored. The total risk of loading and storing casks is given by,

Rcasks = rcask,load * Nload + rcask,store * Nstore

where: rcask,load = risk per cask loaded

Nload = number of casks loaded per year

rcask,store = risk per cask in storage

Nstore = number of casks being stored

Section 1.5 of the SFPS report provides an overview of contributors to SFP risk. The majority of 

SFP risk is thought to emanate from a loss of water from a leak or a boiloff. The risk from the 

SFP can then be characterized as the frequency of fuel uncovery multiplied by the 

consequences of the accident. The uncovery frequency is the sum of the frequency of uncovery 

from cask drops, seismic events, and other initiators. The frequency of a cask drop damaging 

the pool and leading to uncovery is the product of the number of casks loaded, the probability of 

a drop, and the probability of pool damage and uncovery given a drop. This value is given by,

Rsfp = (Nload * Pdrop * Pdamage + Fseismic + Fother) * Cuncovery

where: Pdrop = probability of a cask drop per cask loaded

Pdamage = probability of a dropped cask leading to fuel uncovery

Fseismic = frequency of uncovery from seismic events

Fother = frequency of uncovery from sources other than cask drops 

and seismic

Cuncovery = consequences of fuel uncovery

The SFPS provided a detailed analysis of the consequences, Cuncovery, for a particular site and a 

calculation of Fseismic for seismic bin 3. To fully calculate Fseismic, seismic bin 4 would need to be 

analyzed as well. The SFPS did not analyze other initiators for pool accidents that contribute to 

SFP risk.
B-3

B.5 Risk during Each Stage

Figure 139 is an illustration of the spent fuel risks during each stage for both the current practice 

and expedited transfer strategies. Though the “current practice” strategy does not include 

expedited loading, it is divided into the same stages (time periods) for comparison purposes. 

The figure depicts the SFP risk, dry cask loading risk and dry cask storage risk. The SFP risk 

includes the risk to the pool from dropped casks.

Figure 141 includes the following major assumptions and limitations:

• The figure is only intended to show trends, not absolute differences in risk. No specific 

numbers were used to generate the figure.

• The type of risk used will significantly affect the relative values of different portions of the 

figure. Table 37 gives the ratio of consequences between a high- and low- density pool 

for several types of risk, with the risk reduction from a low-density pool varying from a 

factor of 2.1 for individual latent cancer fatality risk for 0–10 miles to 56 for land 

interdiction.

• The amount of time spent in each stage will affect a calculation of the total risk.

• Changes in Nload, Nstore, and Cuncovery are the drivers for the change in risk between 

stages. Other terms in the above equations are assumed to be constant.

• The specific shape of the figure will depend on site specific parameters such as the 

pool’s susceptibility to be drained from a cask drop event (Pdamage).

• Risks are averaged over the operating cycle to demonstrate general trends rather than 

short-term changes in risk.
B-4

Figure 141 Graphical representation of spent fuel risks

Stage1 Stage 2 Stage 3 Stage 4 Stage 5

SFP Risk

Current Practice

Expedited

Stage1 Stage 2 Stage 3 Stage 4 Stage 5

Dry Cask Loading Risk

Current Practice

Expedited

Stage1 Stage 2 Stage 3 Stage 4 Stage 5

Dry Cask Storage Risk

Current Practice

Expedited
B-5

During Stage 1, for both the current practice and expedited transfer scenarios, the amount of 

spent fuel being stored in the SFP is increasing. As the pool moves from low-density loading to 

high-density loading, the consequences of fuel uncovery, Cuncovery, and thus SFP risk, Rsfp, 

increase. Dry cask loading and storage risk is zero since no casks are loaded during this stage 

(Nload and Nstore = 0).

At the beginning of Stage 2, the pool reaches a high-density configuration and cask loading 

begins. SFP risk is greater than at the end of Stage 1 because of the possibility of cask drops 

(Nload > 0). It is assumed that the rate of dry cask loading is constant throughout this stage,

leading to a constant loading risk and a gradually increasing storage risk as more casks are 

stored (Nstore is increasing). For the current practice spent fuel storage strategy, the same 

loading rate is maintained in Stages 2, 3 and 4 and the pool is maintained at maximum loading.

For the expedited transfer strategy, Stage 3 is the beginning of an increased cask loading rate 

(Nload increases). The SFP risk undergoes another step increase (from the increased frequency 

of cask drop events) and then declines as the pool approaches a low-density configuration and 

the consequence of fuel uncover, Cuncovery, decreases. Cask loading risk increases as the rate 

of loading, Nload, increases. Storage risk increases at a faster rate while more casks are being 

loaded.

Stage 4 marks the end of the expedited transfer phase when the pool has reached a low-density 

configuration. The cask loading rate and risk decrease to levels slightly higher than in Stage 2. 

The hotter fuel being loaded requires more lower capacity casks. The decrease in cask loading 

rate, Nload, and consequences of uncovery, Cuncovery, decrease the SFP risk which remains at a 

constant, lower level for the rest of the stage. Cask storage risk continues increasing at a 

slower rate.

At the beginning of Stage 5, the reactor ends its final operating cycle and fuel in the reactor core 

is offloaded to the SFP. After several months, the fuel in the SFP is generally capable of being 

air cooled, and the risk decreases for both the current practice and expedited transfer 

strategies. The risk is nonzero because of the possibility of events which may impede air 

cooling of the fuel. It is assumed that casks continue being loaded at a constant rate until the 

pool is empty. The SFP risk continues decreasing gradually as the fuel cools and is removed 

from the pool. When the cask loading is complete, the pool risk and the cask loading risk go to 

zero, and the cask storage risk stabilizes. The low-density pool in the expedited transfer case

contains less fuel and is emptied sooner since much of the fuel was removed in Stage 3.

B.6 Total Risk over Time

Two components compromise the total risk over a period of time, (1) the amount of time spent in 

each stage and (2) the risk in each stage. The time spent in each stage will vary depending on 

how soon expedited transfer is initiated (if at all), how long it takes, and how long the reactor 

continues to operate with a low-density pool. The risks during each stage will depend on the 

relative hazards at each site.

The recent EPRI report analyzing spent fuel transfer (EPRI, 2012) estimates that expedited 

transfer for fuel cooled longer than five years would take between 8-15 years at most sites. The 

amount of fuel in the SFP, whether the site has multiple units sharing a single cask handling 

crane, and the availability of trained personnel, and equipment affect this estimate. The amount 

of time then spent in a low-density configuration depends on how much longer the reactor is 
B-6

operated. It is expected that most reactors will apply for and receive extensions of their 

operating licenses to 60 years.

The site-specific risks during each stage will drive whether expedited transfer decreases risk, 

and over what timeframe. An accounting of the risk increases and decreases of expedited 

transfer compared to current practice will illustrate this point.

For expedited transfer, risk increases relative to current practice are seen in the following 

stages:

• SFP risk in the beginning of Stage 3 from cask drop events,

• Cask loading risk during Stage 3, and to a lesser extent in Stages 4 and 5,

• Storage risk in Stages 3 and 4 and the beginning of Stage 5

Risk decreases occur in:

• SFP risk later in Stage 3 and in Stage 4,

• cask loading risk in Stage 5

Since the total number of casks loaded is likely to be only slightly higher for the expedited 

strategy, the increase in cask loading risk during Stage 3 is expected to be mostly offset by the 

decrease in risk in Stage 5. Furthermore, previous studies such as NUREG-1864, “A Pilot 

Probabilistic Risk Assessment of a Dry Cask Storage System At a Nuclear Power Plant,” issued 

March 2007, and an EPRI report entitled, “Probabilistic Risk Assessment (PRA) of Bolted 

Storage Casks,” indicate that dry cask storage risk is likely lower than SFP risk. Hence, this 

leaves a comparison of the increase in SFP risk from cask drops to the decrease in risk from a 

low-density pool.

The risk increase will depend on the pool’s susceptibility to drops (discussed below). For 

example, if the SFP risk at a particular site were dominated by cask drop risk, it could take many 

years of operating with a low-density pool to “pay back” the temporary risk increase seen at the 

beginning of Stage 3. This increase in risk could be mitigated by only loading casks during 

operating cycle phases 4 or 5 when the SFP is typically air coolable for at least 72 hours, for a 

complete draindown. In contrast, a pool with low susceptibility to cask drops and high seismic 

risk will see a greater risk benefit sooner.

One aspect not included in the above figure is the potential need to repackage casks that have 

already been loaded, before interim storage or permanent disposal. Given the uncertainty in the 

national strategy for spent fuel, the specifications for disposal at a long-term repository or 

interim storage site are currently unclear but may be developed in the future. Earlier movement 

of fuel into current cask designs increases the probability that fuel will have to be repackaged to 

meet these specifications.

B.7 Availability of Information

The equations defined earlier identify the variables needed to calculate the risk of spent fuel 

storage. The general shape of Figure 141 is believed to generally apply to operating reactors. 

However, no analysis has attempted to quantify the horizontal or vertical axes. The discussion 

below points to existing information that could be useful in quantifying these variables as well as 

what further information could be useful.
B-7

B.7.1 Cask Risks (rcask,load and rcask,store)

Two major studies have addressed the risk of dry cask loading and storage, NUREG-1864 and 

EPRI’s probabilistic risk assessment of bolted storage casks. NUREG-1864 analyzed a welded 

cask at a particular boiling-water reactor site. In a complementary effort, the EPRI study 

analyzed a bolted cask at a generic pressurized-water reactor site. NUREG-1864 identified 

cask drops and aircraft strikes as the major contributors to risk during cask loading and cask 

storage, respectively. The EPRI study found the major contributors to risk during cask loading 

to be drops, failure of the refueling building, and high temperature fires. During storage, major 

risk contributors were high temperature and forces (e.g., explosions) or heavy loads (e.g., high 

winds) leading to cask tipover. The difference in major contributors to risk is likely because of 

differences in the methods used in the analysis as well as differences in the analyzed systems. 

Regardless, both studies found the risk of dry cask loading and storage to be extremely small. 

Key factors contributing to this result include the robustness of the analyzed casks and the 

availability of the refueling building ventilation system, which is capable of significantly 

decreasing the source term for many accident sequences that result in a cask release.

Several additional factors may affect a calculation of dry cask risk. Considerable uncertainty 

exists in the source term expected from cask accident sequences resulting in a significant range 

in consequences, as discussed in Chapter 10. Different cask designs will vary in their ability to 

resist hazards and may have failure modes not considered in previous studies. Since no 

standard for performing a dry cask PRA exists, these issues will have to be addressed on a 

case-by-case basis. The applicability of the assumptions and limitations of previous studies to 

any future analysis will have to be carefully considered.

B.7.2 Number of Casks (Nload and Nstore)

The number of casks loaded, Nload, and stored, Nstore, affects the total cask risk, Rcask. The 

number of casks loaded also affects the SFP risk, Rsfp, because of the potential for cask drops. 

As discussed above, cask loading is assumed to begin in Stage 2, increase during Stage 3 

(expedited transfer), and, in Stage 4, return to a lower level necessary to maintain a low-density 

configuration in the pool.

NUREG-1353 (NRC, 1989a) and the EPRI report on spent fuel transfer (EPRI, 2012) include 

estimates for the number of casks loaded. NUREG-1353 initially assumes two casks are loaded 

per week resulting in 104 loads per reactor year. Using assumptions based on more updated 

information regarding the number of assemblies discharged per reload, the length of the fuel 

cycle, and the capacity of storage casks in use at the time, the analysis revised this estimate 

downward by a factor of 10. The EPRI report contains a more detailed analysis considering 

multi-unit sites and possible expedited loading scenarios. Dependent on these factors, the 

number of casks loaded annually was estimated to average 3 to 4 annually for current sites and 

up to 15 to 19 annually for some sites during expedited loading.

At the end of 2011, more than 1,500 casks had been loaded (EPRI). For a comparative risk 

calculation, site-specific information would have to be collected or estimated. Given the 

uncertainties in the calculation of risk per cask, and the fact that the risk of loading and storing 

casks has been estimated to be lower than the risk associated with SFPs, a precise number of 

casks loaded and stored is not expected to drive the results.
B-8

As a first approximation, one might assume that the total number of casks loaded from the SFP 

would be the same no matter the fuel management strategy. However, expedited fuel transfer 

requires loading fuel with a higher heat load into casks. Therefore, expedited fuel transfer may 

result in more casks being loaded with different accident consequences than the current 

package. The EPRI report estimates the increased number of casks required.

B.7.3 Pool Uncovery Frequency from Cask Drop Events (Nload, Pdrop and Pdamage)

Heavy load drops have the potential to damage the SFP, possibly leading to uncovery of the 

fuel. In general, casks are considered to be the only loads handled over the pool heavy enough 

to have the potential to cause structural damage. Other heavy loads are usually prevented from 

traveling directly over the pool.

SFPs can have a variety of configurations affecting their susceptibility to cask drop events. 

Some pools contain cask loading pits with floors at a higher elevation than the bottom of the 

pool. Damage from a cask drop event would only drain the pool to a certain level, potentially 

giving operators sufficient time to align a makeup water source and continue keeping the fuel 

covered. The cask loading pit may be separated from the pool by a gate. In other pools, casks 

are loaded directly in the SFPs in a section which may or may not be reinforced to reduce the 

risk of damage in a cask drop event.

The total frequency of uncovery will be a function of how many casks are loaded, the estimated 

probability of a drop per loading, and the probability of damage given a drop. Expedited transfer 

of spent fuel will lead to increased cask loading for a number of years, increasing the risk of a 

dropped cask damaging the pool. The number of casks loaded is discussed above.

Several studies have addressed the issue of heavy load drops and the anticipated effect on the 

SFP. NUREG-1353 estimated a drop rate of 3.1×10-4 per reactor year assuming two lifts per 

week without consideration of Generic Safety Issue (GSI) A-36, “Control of Heavy Loads Near 

Spent Fuel.” The reduction in the probability of a cask drop for a plant which complies with the 

resolution of GSI A-36 was estimated to be a factor of 0.001 for a revised probability of 3.1x10-7

per reactor year. A LLNL analysis reported in NUREG/CR-5176 in support of NUREG-1353 

considered worst-case cask drops on the pool wall for a boiling-water reactor and a pressurizedwater reactor (Prassinos, 1989). The analysis concluded that it is likely that the liner would be 

severely damaged, so a value of 1 was used for Pdamage. Based on updated information, 

NUREG-1353 judged two lifts per week (104 per year) to be an overestimate by about a factor 

of 10. The final estimate of the frequency of a cask being dropped and damaging the SFP is 

3.1×10-8 with an upper bound estimate of 3.1×10-7

. This analysis only considered casks 

dropped on the SFP wall.

NUREG-1738 considered drops that would catastrophically fail the pool, leading to a rapid 

draindown and failure modes other than drops onto the SFP wall (NRC, 2001). The analysis 

assumed that only casks are heavy enough to cause catastrophic damage to the pool. Data 

from NUREG-0612, “Control of Heavy Loads at Nuclear Power Plants: Resolution of Generic 

Technical Activity A-36,” issued July 1980, and NUREG/CR-5176 were combined with a 

calculation of the fraction of the load path spent over the pool and the fraction of the total path 

the load is lifted high enough to damage the pool to estimate the probability of a drop that 

damages the pool. For a nonsingle-failure-proof crane, the drop frequency was estimated, 

based on NUREG-0612 information, to have a mean value of 2.1×10-5 per year (using 100 lifts 

per year). For single-failure-proof cranes or plants that conform to the NUREG-0612 guidelines, 

the drop frequency was estimated to have a mean value of 2×10-7 per year (for 100 lifts per 
B-9

year). The analysis assumed that licensees with a non-single-failure-proof crane took 

appropriate mitigative actions to reduce the expected frequency of catastrophic damage to the 

same range as for facilities with a single-failure-proof crane.

NUREG-1864 used empirical drop data reported in NUREG-1774, “A Survey of Crane 

Operating Experience at U.S. Nuclear Power Plants from 1968 through 2002,” issued July 2003, 

to estimate the probability of dropping a cask. Three load drop events were identified from an 

estimated 54,000 lifts in the 1968–2002 time period, giving a probability of 5.6×10-5 per lift. This 

probability was considered conservative given that, of the three events, only one was a freefall 

while the other two were uncontrolled descents. The probability of pool damage was not 

estimated.

The EPRI dry cask PRA constructed a fault tree of a crane to address a range of factors and to 

account for specific crane features. Data from NUREG-0612 and other sources were used to 

estimate failure probabilities for basic components as well as human error probabilities. The 

final probability of a cask drop given a lift was estimated to be 5.3×10-6

. The probability of pool 

damage was not estimated.

Data cataloguing the susceptibility of SFPs to cask drops for the reactor fleet is not readily 

available. To address this issue, a risk analysis would need to either perform a site-specific 

analysis of cask drops, or conservatively assume that most (if not all) drops will damage the 

pool.

B.7.4 Pool Uncovery Frequency from Seismic Events (Fseismic)

The frequency of seismic events damaging the pool liner and leading to fuel uncovery depends 

on both the seismic hazard (i.e., the frequency of the initiating event) and the fragility of the SFP 

(i.e., the probability that the liner fails given that the event occurs). Chapter 3 of the SFPS 

report discusses the availability of seismic hazard information.

In contrast, seismic fragility data has not been characterized for most SFPs. NUREG/CR-5176 

used a fragility analysis approach involving seismic loads derived from floor response spectra 

for the reactor building based on design response spectra. These loads were then combined 

with analytical methods for the calculation of the fundamental period of vibration of SFP floors 

and walls, as well as approximate methods for the calculation of the strength of these floors and 

walls. The approach used to derive the SFP fragility was generally consistent with methods 

used for seismic margin assessments at the time of that study. Since NUREG-1738 was not a 

site-specific analysis, an attempt was made to generalize this information. NUREG-1738 

convolved a generic fragility (roughly corresponding to the fragilities calculated in 

NUREG/CR-5176) with EPRI and LLNL seismic hazard estimates to estimate the seismic risk. 

The study also developed a screening checklist such that a plant passing the checklist would 

have confidence of having a pool fragility of at least the assumed amount. 

Finally, the SFPS performed a detailed analysis for the reference plant employing a combination 

of the approach used in NUREG/CR-5176 to estimate seismic loads with finite element 

analyses of the SFP structure to calculate hydrodynamic impulsive loads, nonlinear response 

mechanisms and strain concentrations in the liner. Chapter 4 of the SFPS report describes the 

structural analysis and estimated SFP performance. Chapter 10 provides a comparative 

assessment of the estimated performance for the SFP considered in the study with the 

performance of SFPs in recent major earthquakes in Japan including the SFPs at the 

Fukushima Daiichi nuclear power plant under the Tohoku earthquake of March 11, 2011. 
B-10

The most robust way to estimate the seismic risk would be to utilize existing hazard estimates, 

and perform a site-specific fragility analysis. For some analyses, particularly those considering 

multiple sites, this may be time and cost prohibitive since the staff and licensees do not 

generally have fragility analyses of the pools. A generalized analysis for all plant sites would 

have to address the uncertainty in the variation of SFP responses to seismic events. One 

approach would be to use a conservative fragility estimate and to develop a checklist to ensure 

that the estimate is appropriate for the pools being considered.

B.7.5 Pool Uncovery Frequency from Other Events (Fother)

As discussed in Section 1.5 of the SFPS report, the majority of SFP risk is believed to emanate 

from pool leakage events such as cask drops and seismic events discussed above, as well as

events that preclude water injection for a long period of time (e.g. days). Table 1 in the SFPS 

report shows the frequency of fuel uncovery from various contributors calculated in 

NUREG-1353 and NUREG-1738.

B.7.6 Pool Consequences (Cuncovery)

The SFPS is the most detailed analysis to date of SFP consequences. As discussed in the 

SFPS report, the study was performed for a specific site for a specific initiating event. However, 

the consequence results will largely hold for other initiating events and may offer insights 

applicable to other sites. When applying the consequence results to one or several other sites, 

the assumptions used in the study, discussed in Chapter 2 of this report, must be considered 

along with which factors drove the results, discussed in Chapter 10 of this report. Some of 

these drivers and how they are expected to vary between plants are discussed below.

Once fuel in the pool has become uncovered, it may still be coolable from natural circulation of 

air, depending on the amount of decay heat and the amount of cooling. In the SFPS, the fuel is 

estimated to not be air coolable for 10 percent of the operating cycle. Factors affecting this 

include the amount of fuel in the pool, its configuration, burnup, geometry of the fuel racks, etc. 

A partial draindown event with channeled fuel could impede airflow and increase the time to 

coolability.

A significant driver of the amount of radioactivity released is whether a hydrogen combustion 

event occurs. The SFPS results predict these events in some high density loading situations, 

but not in any low density loading scenarios. It’s not clear whether this result will hold true for 

other reactor sites and what level of pool loading is sufficient to achieve this result. 

Furthermore, the SFPS did not consider hydrogen events from hydrogen originating from a 

concurrent reactor accident.

The consequence metric used will significantly affect the outcome of any comparative risk 

calculation. Comparisons of results using different consequence metrics are seen in Table 38

of the SFPS for high-density versus low-density fuel loading and are discussed in Chapter 10 for 

previous SFP and dry cask risk studies. These results demonstrate that the individual latent 

cancer fatality risk metric is relatively insensitive to changes in release magnitude for spent fuel 

accidents. Other metrics, such as land contamination, are much more affected by the amount 

of radioactivity released. Specific reasons for this are discussed in more detail in Sections 7.6 

and Chapter 10.
B-11

Other site-specific factors that may affect the consequences of pool uncovery include SFP 

inventory, mitigative measures, and the surrounding population density and land value. The 

SFPS analysis for these aspects may have varying levels of applicability to other sites.

B.7.7 Other Spent Fuel Risk Considerations

Several additional events are not believed to significantly contribute to spent fuel risk. Dropping 

a single assembly is not expected to challenge the integrity of the pool, but may release some 

radiation. NUREG-1864 analyzes this event. Criticality events, which NUREG-1353 assumes 

not to be an issue, are considered in Section 3.6 of NUREG-1738. Although this report does 

not explicitly evaluate criticality events, Section 2.3 does discuss them.

B.8 Conclusions

This appendix discusses some of the information needed to perform a risk comparison of spent 

fuel storage strategies. NUREGs 1353, 1738 and 1864 provide much of the information for 

certain plants, and could be supplemented by the site-specific analyses described in the SFPS 

report. A complete comparison depends on several factors, including the relative, site-specific

risks, and the time spent in each stage of fuel storage. The benefit of lower SFP risk from low 

density loading may be offset by increases in other risks, such as the risk from cask drop events 

damaging fuel in the cask or the SFP. However, the magnitude of that offset has not been 

completely calculated for any single plant. Additional work would be necessary to evaluate the 

applicability of existing information to a particular site or a group of sites.

C-1

APPENDIX C: CORRESPONDENCE

The following is a list of correspondences that are related to the development and finalization of 

this report. They are laid out in chronological order.

1. “Spent Fuel Pool Scoping Study” (ADAMS Accession No. ML12108A216):

Letter from the Advisory Committee on Reactor Safeguards (ACRS) on April 25th

2012 to the Honorable Chairman Jazko. In this letter the ACRS describes the 

results of the meeting between the ACRS and the Office of Nuclear Regulatory 

Research (RES) with respect to the spent fuel pool study during the 593rd

meeting of the ACRS.

2. “Response to the Advisory Committee on Reactor Safeguards Letter, Dated April 25, 

2012, On The Spent Fuel Pool Scoping Study” (ADAMS Accession No. ML12137A343): 

Letter from the NRC Staff to the ACRS Chairman, Dr. J. Sam Armijo, dated May

23rd, 2012. This letter was in response to the ACRS Chairman’s April 25th letter to 

the Commission.

3. “Staff Requirements – Meeting With The Advisory Committee on Reactor Safeguards, 

9:30 A.M., Thursday, June 7, 2012, Commissioners’ Conference Room, One White Flint 

North, Rockville, Maryland (Open to Public Attendance)” (ADAMS Accession No. 

ML121980043): 

Staff Requirements Memo of July 16th, 2012. SRM directing staff to include 

additional scope to the Spent Fuel Pool Scoping Study report as recommended 

in the ACRS letter to the Chairman on April 25th, 2012.

4. “Spent Fuel Pool Study” (ADAMS Accession No. ML13198A433):

Letter from the ACRS on July 18th, 2013 to the Honorable Chairman Allison M. 

Macfarlane, from Dr J. Sam Armijo. In this letter the ACRS describes the results 

of a meeting between the ACRS and RES during the 606th meeting of the ACRS.

5. “Response To The Advisory Committee On Reactor Safeguards Letter, Dated July 18, 

2013, On The Spent Fuel Pool Study” (ADAMS Accession No: ML13212A087): 

Letter from the NRC Staff to the to the ACRS Chairman, Dr. J. Sam Armijo, dated 

August 14, 2013. The letter was in response to the ACRS Chairman’s July 18th

letter to the Commission.

6. “NRC Seeks Public Comment on Spent Fuel Pool Study” (ADAMS Accession No.: 

ML13175A104): 

NRC staff provided press release requesting public comments, dated June 24, 

2013. Also see Federal Register Notice dated July 2, 2013 (FR Doc No.: 2013-

15840, Pg 39781-39782).

7. “Consequence Study Of A Beyond-Design-Basis Earthquake Affective The Spent Fuel 

Pool For A U.S. Mark 1 Boiling-Water Reactor” (ADAMS Accession No.: ML13256A339):

NRC SECY to the Commission providing the final report, dated October 9, 2013
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1: “Spent Fuel Pool Scoping Study” (ADAMS Accession No. ML12108A216)
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C-4
C-5
C-6
C-7
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2: “Response to the Advisory Committee on Reactor Safeguards Letter, Dated April 25, 

 2012, On The Spent Fuel Pool Scoping Study” (ADAMS Accession No. ML12137A343).
C-9
C-10
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3: Staff Requirements – Meeting With The Advisory Committee on Reactor Safeguards, 

 9:30 A.M., Thursday, June 7, 2012, Commissioners’ Conference Room, One White Flint 

 North, Rockville, Maryland (Open to Public Attendance)” (ADAMS Accession No. 

 ML121980043)
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4: “Spent Fuel Pool Study” (ADAMS Accession No. ML13198A433):
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5: “Response To The Advisory Committee On Reactor Safeguards Letter, Dated July 18, 

 2013, On The Spent Fuel Pool Study” (ADAMS Accession No: ML13212A087)
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6: “NRC Seeks Public Comment on Spent Fuel Pool Study” (ADAMS Accession No.: 

 ML13175A104)
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7: “Consequence Study Of A Beyond-Design-Basis Earthquake Affective The Spent Fuel 

 Pool For A U.S. Mark 1 Boiling-Water Reactor” (ADAMS Accession No.: ML13256A339)
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D-1

APPENDIX D: REGULATORY ANALYSIS AND BACKFITTING 

DISCUSSION TO DETERMINE THE SAFETY BENEFIT OF 

EXPEDITED TRANSFER OF SPENT FUEL AT A REFERENCE 

PLANT

U.S. Nuclear Regulatory Commission

Office of Nuclear Reactor Regulation
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ABBREVIATIONS AND ACRONYMS

ADAMS Agencywide Documents Access and Management System

Bq Becquerel

BLS Bureau of Labor Statistics

BWR boiling-water reactor

CDF Core Damage Frequency

CFR Code of Federal Regulations

CoC certificate of compliance

Cs cesium

DOE U.S. Department of Energy

DSC dry storage cask systems

EPA U.S. Environmental Protection Agency

EPRI Electric Power Research Institute

FR Federal Register

FTE Full-Time Equivalent

ISFSI independent spent fuel storage installation

LCF latent cancer fatality

LERF Large Early Release Frequency

LNT linear no-threshold

LOOP loss of offsite power

MACCS2 MELCOR Accident Consequence Code System, Version 2

MELCOR not an acronym

NPV Net Present Value

NRC Nuclear Regulatory Commission

NTTF Near-Term Task Force

OCP operating cycle phase

OMB Office of Management and Budget

PAG protective action guides

PGA peak ground acceleration

PRM petition for rulemaking

RA Regulatory Analysis

SFP spent fuel pool

SRM Staff Requirements Memorandum

USGS United States Geological Survey
D-3

D.1 INTRODUCTION

This appendix, which is organized into five sections, presents the regulatory analysis and 

backfitting discussion to determine the safety benefit of expedited transfer of spent fuel at a 

reference plant:

• Section D.1 describes the nature of the problem and a clear statement of the objective of 

the proposed action.

• Section D.2 describes and clearly explains the alternative approaches considered.

• Section D.3 describes the attributes affected, the methodology used to evaluate benefits 

and costs, the analysis model, key data and assumptions, and results for the alternatives 

evaluated.

• Section D.4 presents the analytical results and findings including discussion of 

supplemental considerations, uncertainties in estimates, and results of sensitivity 

analyses on the overall costs and benefits.

• Section D.5 presents the preferred alternative and the basis for selection, discusses any 

decision criteria used, identifies and discusses the regulatory instrument to be used (as 

applicable), and explains the statutory basis for the action.

D.1.1 Statement of the Problem

Various risk studies (most recently NUREG-1738, “Technical Study of Spent Fuel Pool Accident 

Risk at Decommissioning Nuclear Power Plants,” February 2001) have shown that storage of 

spent fuel in a high-density configuration in spent fuel pools is safe and that the risk is 

appropriately low. These studies used simplified and sometimes bounding assumptions and 

models to characterize the likelihood and consequences of beyond-design-basis spent fuel pool

accidents46. As part of NRC’s post-9/11 security assessments, spent fuel pool modeling using 

detailed thermal-hydraulic and severe accident progression models integrated into the MELCOR 

code were developed and applied to assess the realistic heatup of spent fuel under various pool 

draining conditions. Moreover, in conjunction with these post-9/11 security assessments, NRC 

issued a new regulation in 2009, 10 CFR 50.54(hh)(2), that requires reactor licensees to 

develop and implement guidance and strategies intended, in part, to maintain or restore spent 

fuel pool cooling capabilities following certain beyond design basis events.

Recently, the agency has restated its views on the safety of spent fuel stored in high-density 

configurations in a response to Petition for Rulemaking (PRM)-51-10 and PRM-51-12 

(73 FR 46204, August 8, 2008) as well as the revision to NUREG-1437 (the Generic 

Environmental Impact Statement for License Renewal, Draft Report for Comment). However, 

this position relies in part on the findings of the aforementioned security assessments, which are 

not publicly available. The events in Japan following the March 2011 earthquake has rekindled 

public and industry interest in understanding the consequences from postulated accidents 

associated with high-density spent fuel pool storage and the relative benefits of low-density 

spent fuel pool storage. This study provides an analysis of the health and safety benefits, if any, 

 46 An overview of previous studies is provided in section 10.2 to the Spent Fuel Pool Study.
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from moving from high-density to low-density spent fuel pool storage at the reference plant. This 

appendix assesses the costs and benefits of this activity, and then, for illustrative purposes,

assesses whether the benefits are cost justified and substantial enough to justify a backfit to 

impose these requirements on the reference plant.

In response to these recent events, the staff has determined that it should confirm that 

high-density spent fuel pool configurations continue to provide adequate protection, and assess 

whether any safety benefits (or detriments) would occur from expedited transfer of spent fuel to 

dry cask storage at the reference plant.

The purpose of this regulatory analysis is to help ensure that:

• Appropriate alternatives to regulatory objectives are identified and analyzed.

• No clearly preferable alternative is available to this action.

• The costs of implementation are justified by its effect on overall protection of the public 

health and safety.

D.1.2 Objective of Proposed Action

Following the March 2011 accident at the Fukushima Daiichi nuclear power plant in Japan that 

resulted after the Tohoku Earthquake and subsequent tsunami, several stakeholders submitted 

comments to the Commission and staff requesting that regulatory action be taken to require the 

expedited transfer of spent fuel stored in spent fuel pools to dry cask storage. The rationale 

was that transferring the spent fuel to dry storage would lessen the potential consequences 

associated with a loss of spent fuel pool coolant inventory by decreasing the amount of spent 

fuel stored in these pools and thereby decreasing the heat generation rate and radionuclide 

source term associated with the spent fuel in pool storage.

As directed by the Commission in SRM-SECY-12-0025, dated March 9, 2012, the staff has 

undertaken regulatory actions that originated from the NTTF recommendations to enhance 

reactor and spent fuel pool safety. On March 12, 2012, the staff issued Order EA-12-051, which 

requires that licensees install reliable means of remotely monitoring wide-range spent fuel pool

levels to support effective prioritization of event mitigation and recovery actions in the event of a 

beyond-design-basis external event. In addition, the staff issued Order EA-12-049 which 

requires that licensees develop, implement, and maintain guidance and strategies to maintain or 

restore core cooling, containment, and spent fuel pool cooling capabilities following a beyond 

design basis external event. Upon full implementation of these Orders, spent fuel pool safety 

will be significantly increased.

While the staff has concluded, based on previous studies without these enhancements, that 

both spent fuel pools and dry casks provide adequate protection of public health and safety, the 

staff has determined that it should confirm that spent fuel pools continue to provide adequate 

protection.

This analysis uses information contained within the “Consequence Study of a Beyond-DesignBasis Earthquake Affecting the Spent Fuel Pool for a U.S. Mark I Boiling Water Reactor” (Spent 

Fuel Pool Study or main document), to evaluate whether there is a benefit at the reference plant 

in the study to change from high- to low-density storage configurations in the spent fuel pool.
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This analysis calculates the potential benefit per reactor year resulting from expedited fuel 

transfer by comparing the safety of high-density fuel pool storage relative to low-density fuel 

pool storage. The comparison uses the initiating frequency and consequences from the Spent 

Fuel Pool Study as an indicator of any changes in the NRC’s understanding of safe storage of 

spent fuel. The staff also used calculated results from previous spent fuel pool studies (i.e., 

NUREG-1353 and NUREG-1738) to extend the applicability of this evaluation to include other 

initiators, which could challenge spent fuel pool cooling or integrity.

D.2 IDENTIFICATION AND PRELIMINARY ANALYSIS OF ALTERNATIVE 

APPROACHES

This section presents the analysis of the alternatives that the NRC considered to meet the 

regulatory goals identified in the previous section. The NRC considered the regulatory baseline 

and one alternative to change this baseline as discussed below.

D.2.1 Alternative 1 – Regulatory Baseline – Maintain the Existing Spent Fuel Storage 

Requirements

This alternative reflects a Commission decision not to expedite the storage of spent fuel to dry 

cask storage, but to continue with NRC’s existing licensing requirements for spent fuel storage. 

Under this alternative, spent fuel is moved into dry storage only as necessary to accommodate 

fuel assemblies being removed from the core during refueling operations. It also assumes that 

all applicable requirements and guidance to date have been implemented, but no 

implementation is assumed for related generic issues or other staff requirements or guidance 

that is unresolved or still under review.

The condition represented by this alternative is the storage of spent fuel in high-density racks47

in the spent fuel pool, a relatively full spent fuel pool, compliance with all current regulatory 

requirements including those requirements associated with the following:

• 10 CFR 50.54(hh)(2) with respect to spent fuel configuration, and spent fuel pool 

preventive and mitigative capabilities,

• Order EA-12-051 that requires licensees to install reliable means of remotely monitoring 

wide-range spent fuel pool levels to support effective prioritization of event mitigation 

and recovery actions in the event of a beyond-design-basis external event, and

• Order EA-12-049 that requires licensees to develop, implement, and maintain guidance 

and strategies to maintain or restore core cooling, containment, and spent fuel pool 

cooling capabilities following a beyond design basis external event.

Furthermore, because spent fuel pools have a limited amount of available storage, even after 

licensees expanded their storage capacity using high-density storage racks, the current practice 

 47 Most nuclear power plant spent fuel pools were originally designed for temporary storage of spent 

fuel. Starting in the 1980s, most pools were “re-racked” to utilize hardware that stores the assemblies 

in a more closely-spaced arrangement, thus allowing the storage of more assemblies in a 

high-density configuration.
D-6

of transferring spent fuel to dry storage in accordance with 10 CFR 72 is assumed to continue.48 

This alternative represents the baseline for estimating the incremental costs of alternative 2.

D.2.2 Alternative 2 – Low-density Spent Fuel Pool Storage

Under this alternative, older spent fuel assemblies49 are expeditiously moved from spent fuel 

pool storage to dry cask storage beginning in year 2014 to achieve and maintain a low-density 

loading of spent fuel in the existing high-density racks within five years. The situation where the 

spent fuel pool is re-racked to a low-density rack configuration was not evaluated because such 

a situation would be inefficient in terms of regulatory benefit given that much of the benefit could 

be achieved by storing less fuel in the existing high-density racks. Because of the low-density 

spent fuel pool loading, this alternative has less longer-lived radionuclide inventory in the spent 

fuel pool, a lower overall heat load in the pool, and a slight increase in the initial water inventory 

that displaces the removed spent fuel assemblies. In certain situations, this additional water 

could delay the clearing of the baseplate, which would temporarily inhibit natural air circulation 

cooling under and up through the racks should the spent fuel pool completely drain.

Due to the uncertainty associated with the schedule for the availability of a spent fuel repository, 

the reference plant has a plan to have sufficient on-site storage capacity (in-pool capacity and

dry storage) to store all of the spent fuel discharged over the operating life of the plant until 

sufficient repository capacity becomes available. As a result, the analyzed incremental increase 

in costs results primarily from the increase in net present value cost for the early transfer of 

spent fuel into dry storage resulting from the earlier capital costs for new casks and for a dry 

storage facility.

The staff recognizes that there are cost and risk impacts associated with the transfer of spent 

fuel from the spent fuel pool to cask storage after five years of cooling and during long-term 

cask storage50. These cost and risk impacts, if included, would reduce the overall net benefit in 

relation to the regulatory baseline. These effects (e.g., the added risks of handling and moving 

casks) were conservatively ignored to calculate the potential benefit per reactor year by only 

comparing the safety of high-density fuel pool storage relative to low-density fuel pool storage

and its implementation costs.

D.3 ESTIMATION AND EVALUATION OF VALUES AND IMPACTS

 48 Maintenance of the existing spent fuel pool storage requirements would not limit the Commission’s 

authority to add new requirements or update regulatory guidelines, as necessary. These actions and 

activities are a part of the regulatory baseline. However, these activities would be pursued as 

separate regulatory actions to resolve particular technical issues. Under this alternative, the NRC 

would take no action to require facilities to expedite the movement of spent fuel to achieve lowdensity loading in the spent fuel pool.

49 Older spent fuel assemblies are those that have been placed in the spent fuel pool to cool for at least 

five years after discharge from the reactor core.

50 EPRI report TR-1021049 assesses the cost and risk impacts (from a worker dose perspective) 

associated with transfer of spent nuclear fuel from spent fuel pools to dry storage after five years of 

cooling. The report concludes that expedited fuel movement would result in an increase cost to the 

U.S. nuclear industry of $3.6 billion, with the increase primarily related to the additional capital costs 

for new casks and construction costs for the dry storage facilities.
D-7

This section discusses the benefits and costs of each action alternative relative to the baseline. 

Ideally, all costs and benefits are converted into monetary values. The total of benefits and 

costs are then algebraically summed to determine for which alternative the difference between 

the values and impacts was greatest. However, in some cases the assignment of monetary 

values to benefits is not provided because meaningful quantification is not possible.

D.3.1 Identification of Affected Attributes

This section identifies the factors within the public and private sectors that the regulatory 

alternatives (discussed in section D.2) are expected to affect. These factors are classified as 

attributes using the list of potential attributes provided by the NRC in Chapter 5 of its Regulatory 

Analysis Technical Evaluation Handbook. The basis for selecting each attribute is presented 

below.

Affected attributes are the following:

• Public Health (Accident). This attribute measures expected changes in radiation 

exposure to the public due to changes in accident frequencies or accident 

consequences associated with the proposed action. The expected changes in 

radiation exposure are measured over a 50-mile radius from the plant site. The dose 

to the public is from reoccupation of the land and other activities following a severe 

accident. In addition, the dose to the public includes the occupational dose to 

workers for cleanup and decontamination of the contaminated land not onsite. The 

calculation for each alternative is made by subtracting the alternative from the 

regulatory baseline.

• Occupational Health (Accident). This attribute measures occupational health effects, 

for both immediate and long-term, associated with site workers because of changes 

in accident frequency or accident mitigation. Within the regulatory baseline, the 

short-term occupational exposure related to the accident occurs at the time of the 

accident and during the immediate management of the emergency and during 

decontamination and decommissioning of the onsite property. The radiological 

occupational exposure resulting from cleanup and refurbishment or decommissioning 

activities of the damaged facility to occupational workers are found within the 

long-term occupational exposure.

• Occupational Health (Routine). This attribute accounts for radiological exposures to 

workers during normal facility operations (i.e., non-accident situations). These 

occupational exposures occur during DSC loading and handling activities; ISFSI 

operations, maintenance, and surveillance activities; and preparing to ship the spent 

fuel offsite.

This attribute represents an estimate of health effects incurred during normal facility 

operations so accident probabilities are not relevant. As is true of other types of 

exposures, a net decrease in worker exposures is taken as positive; a net increase in 

worker exposures is taken as negative. This exposure is also subject to the dollar 

per person-rem conversion factor.

• Offsite Property. This attribute measures the expected total monetary effects on 

offsite property resulting from the proposed action. Changes to offsite property can 

take various forms, both direct, (e.g. land, food, and water) and indirect 
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(e.g. tourism). This attribute is typically the product of the change in accident 

frequency and the property consequences from the occurrence of an accident.

For the regulatory baseline, the offsite property costs are any property consequences 

resulting from any radiological release from the occurrence of an accident. Normal 

operational releases and those releases before severe accident are outside the 

scope of this regulatory analysis.

• Onsite Property. This attribute measures the expected monetary effects on onsite 

property, including replacement power costs, decontamination, and refurbishment 

costs, from the proposed action. There are two forms of onsite property costs that 

are evaluated. The first type is the cleanup and decontamination costs for the unit. 

The second type is the cost to replace the energy from the damaged or shutdown 

units.

• Industry Implementation. This attribute accounts for the projected net economic 

effect on the affected licensees to implement the mandated changes. Costs include 

procedural and administrative activities. Additional costs above the regulatory 

baseline are considered negative and cost savings are considered positive.

• Industry Operation. This attribute accounts for the projected net economic effect due 

to routine and recurring activities required by the proposed alternative on all affected 

licensees.

• NRC Implementation. This attribute accounts for the projected net economic effect 

on the NRC to place the proposed alternative into operation. NRC implementation 

costs and benefits incurred in addition to those expected under the regulatory 

baseline are included. Additional rulemaking, policy statements, new or expedited 

revision of guidance documents, and inspection procedures are examples of such 

costs.

• NRC Operation. This attribute accounts for the projected net economic effect on the 

NRC after the proposed action is implemented. Additional inspections, evaluations, 

or enforcement activities are examples of such costs.

Attributes that are not expected to be affected under any of the alternatives include the 

following: public health (routine), other government, general public, antitrust considerations, 

safeguards and security considerations, regulatory efficiency, improvements in knowledge, and 

environmental considerations.

D.3.2 Methodology Overview

This section describes the process used to evaluate benefits and costs associated with the 

proposed regulatory framework alternatives. The benefits (values) include desirable changes in 

affected attributes (e.g., monetary savings and improved security and safety). The costs 

(impacts or burdens) include undesirable changes in affected attributes (e.g., increased 

monetary costs, and decreased security and safety).

The regulatory analysis methodology is specified by various guidance documents. The two 

documents that govern the NRC’s voluntary regulatory analysis process are NUREG/BR-0058, 

Revision 4, “Regulatory Analysis (RA) Guidelines of the U.S. Nuclear Regulatory Commission,” 
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dated September 2004 (RA Guidelines), and NUREG/BR-0184, “Regulatory Analysis Technical 

Evaluation Handbook,” dated January 1997 (RA Handbook). The regulatory analysis identifies 

all attributes impacted by the proposed alternative and analyzes them either quantitatively or 

qualitatively as described in the previous section.

For the quantified regulatory analysis, the NRC staff develops expected values for each cost 

and benefit. The expected value is the product of the probability of the cost or benefit occurring 

and the consequences that would occur assuming the event actually happens. For each 

alternative, the staff first determines the probabilities and consequences for each cost and 

benefit, including the year the consequence is incurred. The NRC staff then discounts the 

consequences in future years to the current year of the regulatory action. Finally, the NRC staff 

sums the costs and the benefits for each alternative and compares them.

After performing a quantitative regulatory analysis, the NRC staff adds attributes that could only 

be qualified51. Based on the qualification of each attribute, uncertainties, sensitivities, and the 

quantified costs and benefits, the staff makes a recommendation for each alternative. If the 

benefits, both quantified and qualified, are greater than the quantified and qualified costs, then 

the staff recommends the alternative should be implemented. If the benefits, both quantified 

and qualified, are less than the quantified and qualified costs, then the staff recommends the 

alternative should not be implemented.52

D.3.2.1 Analysis Model

This regulatory analysis measures the incremental impacts of the proposed regulatory 

framework alternative to the “continue with the existing regulatory framework” baseline, which 

reflects anticipated behavior in the event that the proposed alternatives are not adopted. 

Section D.4 presents the estimated incremental costs and savings of each alternative relative to 

continuing with NRC’s existing regulatory framework (alternative 1).

Key inputs into the analysis model are discussed in the following subsections.

D.3.2.1.1 Baseline for the Analysis

The regulatory baseline used in the analysis is to continue with NRC’s existing approach to 

spent fuel pool storage. This baseline assumes full compliance with existing NRC 

requirements, including current regulations and relevant orders. This is consistent with 

NUREG/BR-0058, “Regulatory Analysis Guidelines of the U.S. Nuclear Regulatory 

Commission,” Rev. 4, which states that, “in evaluating a new requirement..., the staff should 

assume that all existing NRC and Agreement State requirements have been implemented.”

 51 See NRC’s Regulatory Analysis Technical Evaluation Handbook, Section 4.3, “Estimation and 

Evaluation of Values and Impacts.”

52 See NRC’s Regulatory Analysis Technical Evaluation Handbook, Section 4.5, “Decision Rationale.” 

Non-quantifiable attributes can only be factored into the decision in a judgmental way; the experience 

of the decisionmaker will strongly influence the weight that they are given. Qualitative attributes may 

be significant factors in regulatory decisions and should be considered, if appropriate.
D-10

D.3.2.1.2 Discount Rates

In accordance with guidance from the Office of Management and Budget (OMB) and 

NUREG/BR-0058, Rev. 4, present-worth calculations are used to determine how much society 

would need to invest today to ensure that the designated dollar amount is available in a given 

year in the future. By using present-worth, costs and benefits, regardless of when averted in 

time, are valued equally. Based on OMB guidance Circular No. A-4, September 17, 2003, 

present-worth calculations are presented using both 3 percent and 7 percent real discount 

rates. The 3 percent rate approximates the real rate of return on long-term government debt, 

which serves as a proxy for the real rate of return on savings. This rate is appropriate when the 

primary effect of the regulation is on private consumption. Alternatively, the 7 percent rate 

approximates the marginal pretax real rate of return on an average investment in the private 

sector, and is the appropriate discount rate whenever the main effect of a regulation is to 

displace or alter the use of capital in the private sector.

Although the NRC is not bound to follow OMB guidance, the NRC has voluntarily complied with 

the present-worth calculations developed in OMB Circular No. A-4 and has stated so in the RA 

Guidelines and RA Handbook.

D.3.2.2 Data

The data and assumptions used in analyzing the quantifiable impacts associated with the 

proposed alternative are discussed in this subsection. Information on attributes affected by the 

proposed regulatory framework alternatives were obtained from experienced NRC staff and 

other sources as referenced. The NRC staff considered the potential differences between the 

new requirements and the current requirements and has incorporated the proposed incremental 

changes into this regulatory analysis.

Available cost information is included in the backfitting analysis of the reference plant, which is 

provided for illustrative purposes. The NRC plans to use the insights from this analysis to 

inform a broader regulatory analysis to support decisionmakers in determining whether NRC’s 

regulations should be changed to impose new generic requirements on all operating nuclear 

reactors.

D.3.2.2.1 Spent Fuel Pool Initiator Release Frequency

Section 1.5 of the Spent Fuel Pool Study provides an overview of contributors to spent fuel pool

risk. The majority of spent fuel pool risk emanates from a loss of water from a sizeable leak in 

the spent fuel pool or a boil off in which operator action to inject water into the pool for an 

extended period is precluded. The release frequency from the spent fuel pool can then be 

characterized as the frequency of the initiator causing fuel uncovery multiplied by the probability 

of a release given fuel uncovery for the specific initiating event. The total release frequency is 

the sum of the frequency of releases from cask drops, seismic events, and other initiators. This 

value is given by:

Frelease = � Finitiatori x Preleasei i

Where Finitiator includes

Fdrop = frequency of spent fuel uncovery from cask drops

Fseismic-bin 3 = frequency of spent fuel uncovery from seismic bin 3 event
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Fseismic-bin 4 = frequency of spent fuel uncovery from seismic bin 4 event

Fother = frequency of spent fuel uncovery from sources other than cask drops 

and seismic

Prelease = probability of release given spent fuel uncovery for specific initiators

Source: Derived from Spent Fuel Pool Study, section B.4.

The Spent Fuel Pool Study provides a detailed analysis of the consequences, for a particular 

site and a calculation of Fseismic for seismic bin 3, a hazard exceedance frequency range 

provided in Table 4 of the Spent Fuel Pool Study and reproduced in Table 72.

Table 72 Seismic Bins and Initiating Event Frequencies

Bin No.

Bin Range 

(g)

Bin 

PGA (g)

Approximate Initiating Event 

Frequency (USGS 2008 model) (/yr)

1 0.05 - 0.3 0.12 5.2x10-4

2 0.3 - 0.5 0.4 2.7x10-5

3 0.5 - 1.0 0.7 1.7x10-5

4 > 1.0 1.2 4.9x10-6

The Spent Fuel Pool Study did not analyze initiators that contribute to spent fuel pool risk other 

than for seismic events defined by seismic bin no. 3. However past studies, such as 

NUREG-1353 and NUREG-1738, evaluated additional events that could contribute to risk and 

consequences from spent fuel pool fires. Table 74 summarizes these initiating-event-class fuel 

uncovery frequencies. Uncovery frequencies taken from past studies depend on the 

assumptions stated in those studies. Additionally, seismic bin no. 4 is included by extrapolating 

the results of this study. For seismic bin no. 3 and bin no. 4 events, the uncovery frequency is 

the product of the initiating event frequency, ac power fragility, and the liner fragility.

The main report uses an ac power fragility value of 0.84 taken from NUREG-1150 as a 

surrogate for the conditional probability of normal spent fuel pool cooling and makeup not being 

available following a 0.7g earthquake. This simplifying assumption was made in light of the fact 

that the main report is not a probabilistic risk assessment (but rather a consequence analysis 

with probabilistic considerations) and that this value already approximates the upper bound 

value of 1.00. For the seismic bin no. 4 event, ac power fragility upper bound value of 1.00 was 

used in this regulatory analysis. In reality, the availability of normal spent fuel pool cooling and 

makeup would be a combination of the ac power fragility, the fragility of the actual equipment 

and its support equipment, and operator actions to recover spent fuel pool cooling capabilities 

using additional mitigation equipment and strategies implemented in response to Order EA-12-

049, which were not considered in the main report. The modeling and consideration of these 

guidance and strategies to maintain or restore spent fuel pool cooling capabilities following a 

beyond design basis external event could result in a smaller value for spent fuel pool cooling 

and makeup failure conditional probability than the values used here and a resulting smaller 

initiating event fuel uncovery frequency.

Section 4.1.5 of the main report describes the results from the nonlinear finite element analysis 

to estimate the likelihood of leakage from concrete cracking and related spent fuel pool liner 

failure for the 0.7g earthquake. Figure 27 shows that the maximum membrane effective strain is 

about 3.7 percent. Based on this calculated liner strain for the 0.7g earthquake, a structural 

analysis of the pool estimates that the spent fuel pool in this study has a 90% probability of 

surviving the 0.7g earthquake with no liner leakage (or conversely, a 10% probability of 

damaging the liner such that leakage will occur). As a result, a liner fragility value of 0.1 is used 
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for the seismic bin no. 3 initiating event. For the seismic bin no. 4 initiating event (i.e., 1.2g 

earthquake), a comparable structural analysis was not performed to determine the liner fragility 

value. As detailed in section 4.1.1 of the main report, the specific conditions for liner failure vary 

according to site conditions and spent fuel pool design. NUREG-1353 predicted the likelihood 

of liner failure from all potential earthquakes to be between about two and six times in a million 

years. NUREG-1738 predicted the likelihood of liner failure from all potential earthquakes to be 

between about two times in a million years and two times in 10 million years. Because a 

documented liner fragility value for a 1.2g earthquake for the reference plant is not readily 

available, a conservative bounding approach was used. A liner fragility value of 1.00 is used in 

this regulatory analysis for the best estimate, even though a realistic analysis may be able to 

justify a value a factor of 2 or more lower.

Past studies have reached generally similar conclusions about the relative contribution to risk 

from the seismic initiating events considered. Table 73 Frequency of Spent Fuel Pool Fuel 

Uncovery for Seismic Events summarizes the impact of the above modeling assumptions when 

comparing the seismic initiating event fuel uncovery frequencies from previous spent fuel pool 

accident regulatory analyses.

Table 73 Frequency of Spent Fuel Pool Fuel Uncovery for Seismic Events

Reference Spent Fuel Pool Fuel Uncovery

(per reactor-year)

Percent Increase in Fuel 

Uncovery Frequency Value 

NUREG-1353 (1989)

(BWR, best estimate)1 7x10-6 (10%)

NUREG-17382 2x10-6 315%

This regulatory analysis3 6.3x10-6 100%

1. This number was not multiplied by the stated conditional probability of having a zirconium fire of 0.25.

2. NUREG-1738 presented results for the two different seismic hazard models in wide use at the time 

(the Electric Power Research Institute and Lawrence Livermore National Labs models). The larger of 

the two values is listed above.

3. The initiating event frequency values are from Table 72. The likelihood of fuel uncovery is a product 

of initiating event frequency (e.g., 1.6x10-5 for seismic bin no. 3), ac power fragility (0.84), and liner 

fragility (0.1). For seismic bin no. 4, the likelihood of fuel uncovery is a product of initiating event 

frequency (4.9x10-6

), ac power fragility of 1.0, and a liner fragility of 1.0 (i.e., 100-percent likelihood of 

ac power and pool liner failure).

As discussed in the SFPS report, the study was performed for a specific site and for a specific 

initiating event. Once fuel in the pool has become uncovered, it may still be coolable from 

natural circulation of air, depending on the amount of decay heat and the amount of cooling. In 

section 12.1 of the main report, the fuel is estimated to be air coolable for at least 72 hours for 

all but roughly 10 percent of the operating cycle. Factors affecting this value include the amount 

of fuel in the pool, its configuration, burnup, geometry of the fuel racks, etc. A partial draindown 

event with channeled fuel could impede airflow. In this case with no natural circulation of air 

through the racks, the cooling of the fuel by the spray flow would be the only effective cooling 

mechanism until the decay heat of the fuel is reduced.

For the seismic bin no. 4 event, the spent fuel is assumed to retain an air coolable geometry 

following this event that causes a moderate to large crack in the pool and results in full pool 

draindown. Information provided in NUREG/CR-5176 (Prassinos et al, 1989), which concludes

that there is high confidence that spent fuel pool racks are sufficiently robust to remain generally 

intact with their fuel channels open supports this assumption. Furthermore, prior studies 

conclude that severe earthquakes are not expected to result in catastrophic failure of spent fuel 
D-13

pool structural walls and floor or fuel racks. Section 4.2 of this study cites median fragility for 

the reactor building of about 1.6g. However, the main report did not perform a structural 

analysis to verify that the reference plant spent fuel and racks retain their structural integrity and 

air-coolable geometry following a 1.2g peak ground acceleration seismic event. Given the 

uncertainties involved, a bounding approach was used to evaluate the sensitivity of assuming 

the spent fuel is not air-coolable following a seismic bin no. 4 earthquake that causes a rapid 

draindown of the spent fuel pool. This was done by assuming a value of 1.0 for the high 

estimate of the conditional probability of release for the seismic bin no. 4 unsuccessful 

mitigation event.

For the cask drop event, spent fuel is assumed to retain an air coolable geometry because a 

postulated cask drop accident would most likely affect the fuel pool floor in the cask loading 

area. The overhead crane used to move the casks is designed to meet single failure proof 

criteria, and has interlocks and administrative controls that limit the motion of the crane over the 

spent fuel pool to the cask loading area, where no fuel is stored. Although improbable, crane 

failure is more likely to occur during hoisting operations when many components contribute to 

holding the cask than during translational motion when the hoist holding brakes are set. The 

hoisting activities occur over the cask loading area, and, in that location, the cask, if dropped,

could have sufficient potential energy to damage the spent fuel pool floor. . However, the main 

report did not perform a structural analysis to verify that the reference plant spent fuel and racks 

retain their structural integrity and air-coolable geometry following a cask drop event. Given the 

uncertainties involved, a bounding approach was used to evaluate the sensitivity of assuming 

the spent fuel is not air-coolable following a cask drop accident. This was done by assuming a 

value of 1.0 for the high estimate of the conditional probability of release for the cask drop 

unsuccessful mitigation event.

To calculate the total release frequency, the uncovery frequencies are multiplied by the 

conditional probability of release for each initiating event class. The conditional probability of 

release depends on the fraction of the operating cycle where the fuel is not air coolable. For the 

seismic bin no. 3 event analyzed in the SFPS, this was calculated to be the ratio of 60/730 days 

or 8.2% of the operating cycle. See Section 5.6.3 of the main document for further discussion. 

For the non-seismic and non-cask drop events taken from previous studies, the nature of the 

events may lead to a situation similar to a partial draindown where the rack baseplate is not 

cleared and airflow is impeded. For these events, the conditional release probability is assumed 

to be 100%.

When mitigation is credited, this study found that successful mitigation decreased the 

conditional probability by a factor of 19 for the seismic bin no. 3 event analyzed using mitigation 

measures required under 10 CFR 50.54 (hh)(2). The main report does not consider the 

post-Fukushima mitigation equipment and mitigation strategies for their use required under 

Order EA-12-049 and being implemented by the operating plants that are intended to increase 

the likelihood of restoring or maintaining power and mitigation capability during severe 

accidents. For the purposes of this regulatory analysis, it was assumed that successful 

mitigation decreased the conditional probability by a factor of 19 for all initiating events as 

determined in the main report. In reality, the effectiveness of post-Fukushima improvements to 

severe accident mitigation measures will depend on a variety of factors, which the SFPS did not 

consider, and which are expected to be more effective than what is assumed here. Although 

the likelihood of successful mitigation deployment is uncertain.

Table 74 summarizes the initiating event class fuel uncovery frequencies, the conditional 

probability of release, and the total release frequency with and without mitigation.
D-14

Table 74 Release Frequencies for Spent Fuel Pool Initiators

Spent fuel loading configuration 1x4 1x4

Initiating Event 

Class

Initiating 

Event Fuel 

Uncovery 

Frequency

(per r-yr)

Conditional 

Probability of 

Release 

(Unsuccessful 

mitigation)

Release 

Frequency 

(Unsuccessful 

mitigation)

(per r-yr)

Conditional 

Probability 

of Release 

(successful 

mitigation)

Release 

Frequency 

(successful 

mitigation)

(per r-yr)

Seismic bin no. 3 event 1.4x10-6 (3) 8.2% 1.18x10-7 0.43% (4) 6.18x10-9

Seismic bin no. 4 event 4.9x10-6 (3) 8.2% – 100% 4.03x10-7 – 4.9x10-60.43% (4) 2.12x10-8 – 2.58x10-7

Cask / heavy load drop 2x10-7 (2) 8.2% – 100% 1.64x10-8 – 2x10-70.43% (4) 8.65x10-10 – 1.05x10-8

LOOP – severe weather 1x10-7 (2) 100% 1.00x10-7 0.43% (4) 5.26x10-9

LOOP – other 3x10-8 (2) 100% 3.00x10-8 0.43% (4) 1.58x10-9

Internal fire 2x10-8 (2) 100% 2.00x10-8 0.43% (4) 1.05x10-9

Loss of pool cooling 1.5x10-8 (1) 100% 1.50x10-8 0.43% (4) 7.89x10-10

Loss of coolant inventory 3x10-9 (2) 100% 3.00x10-9 0.43% (4) 1.58x10-10

Inadvertent aircraft 

impacts

3x10-9 (2) 100% 3.00x10-9 0.43% (4) 1.58x10-10

Missiles – general 2.5x10-9 (1) 100% 2.50x10-9 0.43% (4) 1.32x10-10

Missiles - tornado 1x10-9 (2) 100% 1.00x10-9 0.43% (4) 5.26x10-11

Pneumatic seal failures n/a (5)

Total 7.11x10-7 – 5.39x10-6 3.74x10-8 – 2.84x10-7

1. Values from NUREG-1353. These numbers were multiplied by the stated conditional probability of 

having a zirconium fire of 0.25.

2. Values from NUREG-1738

3. Initiating event frequency values from Spent Fuel Pool Study, Table 4. The likelihood of fuel 

uncovery is a product of initiating event frequency (e.g., 1.6x10-5 for seismic bin no. 3), ac power 

fragility (0.84), and liner fragility (0.1). For seismic bin no. 4, the likelihood of fuel uncovery is a 

product of initiating event frequency (4.9x10-6 ), ac power fragility of 1.0, and a liner fragility of 1.0 

(e.g., 100-percent likelihood of ac power and pool liner failure).

4. The conditional probability of release with successful mitigation with deployed 50.54(hh)(2) equipment 

is the quotient of OCP probability (60/730 or 8.2%) divided by the mitigation benefit in reducing the 

release likelihood (factor of 19). See Section 5.6.3 of the main document for further discussion. 

Additional mitigation equipment and mitigation strategies under Order EA-12-049 would further 

enhance the likelihood of successful mitigation, thereby further reducing the value for the conditional 

probability of release with successful mitigation.

5. As discussed in Table 3 of the main report, the reference plant has gates with mechanical seals to 

prevent leakage. These seals are kept under pressure by passive mechanical means (i.e., do not 

depend on air pressure, ac power, or dc power). Therefore, pneumatic seal failures are not 

applicable for the reference plant.

Based on this information, the values used in this regulatory analysis for Frelease is are 

summarized in Table 75.

Table 75 Spent Fuel Pool Release Frequency Estimates

Parameter Unsuccessful mitigation Successful mitigation

Low Best High Low Best High

Frelease 7.11x10-7 5.39x10-6 3.74x10-8 2.84x10-7
D-15

These release frequency values are subject to the assumption of unsuccessful deployment of 

mitigation and the other assumptions contained in this analysis and those stated in Table 3 of 

the main report. A comparison of the release frequencies (total and delta) used in this 

regulatory analysis to the release frequencies used for only seismic bin no. 3 in the Spent Fuel 

Pool Study is provided in Table 76.

Table 76 Release Frequency Comparison Between Inclusion of All Initiator Event 

Classes to the Seismic Bin No. 3 Event

Mitigation Case

Release Frequency 

for All Initiator Events 

Classes (per r-yr)

Release Frequency 

for the Seismic Bin 

No. 3 Event (per r-yr)

Net Percent 

Increase in 

Release 

Frequency

Unsuccessful Mitigation 7.11x10-7 – 5.39x10-6 1.18x10-7 505% – 4489%

Successful Mitigation 3.74x10-8 – 2.84x10-7 6.18x10-9 505% – 4489%

Delta change 6.74x10-7 – 5.11x10-6 1.11x10-7 505% – 4489%

D.3.2.2.2 Duration of On-site Spent Fuel Storage Risk

The reference plant operating license expires in 2034. For this analysis, it is assumed that the 

plant operates through the term of its operating license and that the licensee continues to store 

spent fuel in the pool following commercial operation53 to allow the spent fuel to cool sufficiently 

before placing into dry storage. For all cases analyzed, it was assumed that spent fuel stored in 

the spent fuel pool is susceptible to the risk of spent fuel fires for up to one year after permanent 

cessation of operations.

D.3.2.2.3 Cost/Benefit Inflators

The consequences for some attributes are estimated based on the values published in the NRC 

Regulatory Analysis Handbook. Within the NRC Regulatory Analysis Handbook, the information 

in relation to severe reactor accident consequences is provided in previous year dollars. To 

evaluate the costs and benefits consistently, the consequences are inflated. The most common 

inflator is the Consumer-Price Index for all urban consumers (CPI-U), developed by the U.S. 

Department of Labor, Bureau of Labor Statistics. Using the CPI-U, the previous year dollars 

were converted to the year 2012. The formula to determine the amount in 2012 dollars is 

CPIU2012

CPIUBase Year

∗ ConsequenceBase Year = Consequence2012.

Values of CPI-U used in this regulatory analysis are summarized in Table 77.

Table 77 Consumer Price Index – All Urban Consumers Inflator

Base Year CPI-U Inflator for Year 2012

2005 1.1756

2006 1.1389

 53 Decommissioning of the unit must be completed within 60 years of permanent cessation of operations 

under 10 CFR 50.82, “Termination of License.” Completion of decommissioning beyond 60 years will 

be approved by the Commission only when necessary to protect public health and safety.
D-16

Base Year CPI-U Inflator for Year 2012

2007 1.1073

2008 1.0664

2009 1.0702

2010 1.0529

2011 1.0207

Source: www.bis.gov/data/inflation_calculator.htm

D.3.2.2.4 Dollar per Person-Rem Conversion Factor

Using the dollar value of the health detriment and a risk factor that establishes the nominal 

probability for stochastic health effects attributable to radiological exposure (fatal and non-fatal 

cancers and hereditary effects) provides a dollar per person-rem of $2,000, rounded to the 

nearest thousand, according to NUREG-1530, “Reassessment of NRC’s Dollar per Person-Rem 

Conversion Factor Policy,” dated December 1995.

The NRC currently use a value of statistical life (VSL)54 of $3 million based on NUREG-1530, 

and a cancer risk factor of 7.0 x 10-4

, which is a reduction to the closest significant digit of a 

recommendation by the International Commission on Radiation Protection (ICRP) in Publication 

No. 60. Therefore, the dollar per person-rem is equal to $3 million times 7.0 x 10-4 rounded to 

the nearest thousand (due to uncertainties) or $2,000.

D.3.2.2.5 Onsite Property Decontamination, Repair, and Refurbishment Costs

Spent fuel pool accident risks have significant contributions from onsite property monetary 

losses (e.g., repair and refurbishment) and plant decontamination. The risk dominant accident 

sequences involve the failure of the pool due to seismic or load drop events resulting in the loss 

of pool integrity. This scenario results in loss of spent fuel pool water inventory, zircaloy 

cladding fire initiation with propagation through the spent fuel assemblies stored in the pool, and 

an uncontrolled radiological release from the reactor building. The NRC assumes that, based 

on the current regulatory framework, with insights from the Fukushima Dai-ichi accident, that 

onsite property would be radiologically affected in the following way. The consequences of a 

spent fuel fire are expected to be similar to the Category II accident as defined in 

NUREG/CR-5281, section 3.2.4. Based on this reference, the cleanup and decontamination 

costs are estimated to be approximately $165 million (1983 dollars) and the cost for permanent 

disposal of the damaged fuel is $26 million (1983 dollars). Using Table C.95 from the RA 

Handbook, the pool repair to is expected to cost $72 million (1983 dollars). Adjusting these 

estimated costs using the CPI-U inflator formula and using a multiplier of three to model the high 

estimate and a divider of two to model the low estimate results in the values provided in Table 

78.

 54 The value of a statistical life (VSL) is the monetary value of a mortality risk reduction that would 

prevent one statistical (as opposed to an identified) death (Jones-Lee, 2004). The VSL is a key 

component in the calculation of the dollar per person-rem value, which is the product of the VSL 

multiplied by a risk coefficient.
D-17

Table 78 Onsite Property Decontamination, Repair, and Refurbishment Costs

D.3.2.2.6 Replacement Energy Costs

Replacement energy costs are the costs for replacing the energy from the nuclear power plant 

due to a plant shutdown to install required equipment or due to an accident.55 The NRC 

assumes that replacement energy costs would be required until onsite decontamination and 

repair efforts are completed or the unit is retired.

The NRC assumes that licensees engage in power purchase agreements (PPA)56 to 

economically purchase replacement power. A PPA is a legal contract between an electricity 

generator (licensee) and a power purchaser. The NRC assumes that a licensee will not be able 

to replace the power through other generation for seven years and would have to buy power 

from the market. Although not all licensees may have PPAs, the licensee will still replace the 

lost energy any time that the nuclear power plant is not operating to meet its electrical power 

supply obligations. The NRC assumes that after 7 years, the onsite decontamination and repair 

efforts are completed or the unit is retired and other power sources will be developed to replace 

the unit’s lost electrical generation capability.

For the replacement energy cost calculation in this regulatory analysis, the NRC assumes that 

the reference plant is located on a multi-unit site. For the high estimate case, the NRC assumes 

that replacement energy is purchased for both the accident unit and the co-located unit at the 

site.

D.3.2.2.7 Occupational Worker Exposure (Accident)

There are two types of occupational exposure related to accidents: short-term and long-term. 

The first occurs at the time of the accident and during the immediate management of the 

emergency. The second is a long-term exposure, presumably at significantly lower individual 

rates, associated with the cleanup and refurbishment or decommissioning of the damaged 

facility. The value gained in the avoidance of both types of exposure is conditioned on the 

change in frequency of the accident's occurrence.

 55 The replacement energy cost is only the cost to buy the energy for production on the market. 

Therefore, the cost would be the cost of buying the cheapest energy. These estimates do not include 

transmission or distribution costs.

56 A power purchase agreement is a contract between two parties, one who generates electricity for the 

purpose of sale (the seller) and one who is looking to purchase electricity (the buyer). The PPA 

defines all of the commercial terms for the sale of electricity between the two parties, including when 

the project will begin commercial operation, schedule for delivery of electricity, penalties for under 

delivery, payment terms, and termination.

Best Estimate High Estimate Low Estimate Best Estimate High Estimate Low Estimate

Cleanup and decontamination $165,000,000 $495,000,000 $82,500,000 $371,250,000 $1,113,750,000 $185,625,000

Repair Pool $72,000,000 $216,000,000 $36,000,000 $162,000,000 $486,000,000 $81,000,000

Disposal of damaged fuel $26,000,000 $78,000,000 $13,000,000 $58,500,000 $175,500,000 $29,250,000

Total $263,000,000 $789,000,000 $131,500,000 $591,750,000 $1,775,250,000 $295,875,000

Onsite Property Cost Element

1983 dollars 2013 dollars
D-18

The experiences at the Three Mile Island Unit 2 (TMI-2), the Chernobyl, and the Fukushima 

nuclear power plants illustrated that significant occupational exposures could result from 

performing activities outside the control room during a power reactor accident. At TMI-2, the 

average occupational exposure related to the incident was approximately 1.0 rem, with a 

collective dose of 1,000 person-rem occurring over a 4-month span, after which time 

occupational exposure approached pre-accident levels. For Chernobyl, the average dose for 

persons closest to the plant was 3.3 person-rem (RA Handbook p. 5.30), yielding an average 

value of 3,300 person-rem.

After the Fukushima unit 1 building explosion on March 12, 2011, the unit 3 building explosion 

on March 14, and the unit 4 building explosion and the exposure of the unit 2 reactor fuel rods 

on March 15, radioactive materials were release into the environment and surrounding areas of 

the Fukushima Dai-ichi nuclear power plant. Measurement and evaluation of radiation exposure 

levels for workers engaged in emergency work at the Fukushima Daiichi NPS have been

implemented continuously since the Tohoku Earthquake.

As shown in Figure 142, the dose rate in the vicinity of the main gate at the Fukushima Dai-ichi

site near the time of the Unit 4 explosion varied between 20 mrem and 1.0 rem per hour

(between 200 and 10,000 μSv per hour).

Source: Fukushima Nuclear Accident Analysis Report p. 371.

Figure 142 Dose Rate in Vicinity of Fukushima Daiichi Nuclear Plant Site Main Gate

between March 11 and March 16, 2011

On March 22 and 23, surveys of the airborne radioactivity and dose rates around the Fukushima 

Daiichi site were collected and documented. The dose rates are shown on Figure 143.
D-19

Source: INPO 11-005, p 41

Figure 143 Fukushima Daiichi Site Dose Rates between March 22 and March 23, 2011

The distribution of total monthly exposure for workers engaged in radiation work at the

Fukushima Daiichi NPS for the first three months following the March 2011 accident is provided

in Table 79.

Table 79 Average Accident Occupational Exposure at Fukushima Dai-ichi Nuclear Power 

Plant from March to May 2011

Total Radiation Exposure 

(mSv)

Number of Plant Workers Exposed

March 20111 April 20112 May 20113

≥ 250 6 0 0

200 - 249 2 0 0

150 - 199 14 0 0

100 - 149 77 0 0

50 - 99 309 3 0

20 - 49 859 81 19

10 - 19 1041 310 144

< 10 1434 3214 2854

Total number of workers 3742 3608 3017

1. Maximum March 2011 occupational exposure was 670.4 mSv.

2. Maximum April 2011 occupational exposure was 69.3 mSv.

3. Maximum May 2011 occupational exposure was 41.6 mSv.

4. One mSV is equal to 0.1 rem.

Source: Wada et al, Occupational and Environmental Medicine, 2012 August; 69(8): p. 600.

To estimate the monthly total occupational radiation exposure received by all workers, a high 

estimate, best estimate, and low estimate were calculated based on the maximum category 

value, the midpoint category value, and the first quartile category value. The results are 

tabulated in Table 80.
D-20

Table 80 Estimated Immediate Accident Occupational Monthly Exposure at Fukushima

The immediate accident occupational exposure for a spent fuel pool accident shown in Table 81

is estimated based on the Fukushima data and the following assumptions:

• The immediate accident period lasts for one year,

• The workforce during the immediate accident period is 3,700 workers, and

• The average worker radiation exposure remains constant at the May 2011 value from 

May 2011 through February 2012.

Table 81 Immediate Accident Occupational Exposure for a Spent Fuel Pool Fire

Case Immediate Accident Occupational 

Exposure (averted person-rem)

Low Estimate 18,070

Best Estimate 28,380

High Estimate 48,880

After the immediate response to a spent fuel pool fire, a long process of cleanup and 

refurbishment or decommissioning will follow. The Fukushima Nuclear Accident Analysis 

Report states, “The average value for 5,128 people in April of 2012 was 1.07 mSv per worker 

due to decreasing trends in environment dose rates (p 415). The NRC assumes that the 

process of cleanup and refurbishment or decommissioning will begin one year after the accident 

and will take seven years to complete. During those seven years, the NRC assumes that each 

occupational worker at the damaged reactor site will be exposed to 1.07 mSv per month 

(0.107 rem per month) for the duration of the cleanup and refurbishment or decommissioning. 

Assuming the average value for 5,128 workers would remain for the duration yields a 

cumulative long-term occupational dose of 46,000 person-rem.

In NUREG/CR-5281, Jo et al. (1989) conducted what essentially amounted to a regulatory 

analysis of a non-reactor nuclear fuel cycle facility using the 1983 Handbook (Heaberlin et al. 

1983) as guidance. The accidental occupational exposure was assumed to be similar to that 

from TMI-2, which is 4,580 person-rem.

As described in the RA Handbook (p 5.30), the DOE (1987) summarized results on the 

collective dose received by the populace surrounding the Chernobyl accident. Average dose 

equivalents of 3.3 rem per person, 45 rem per person, and 5.3 rem per person were estimated 

for residents within 3 km, between 3 km and 15 km, and between 15 km and 30 km of 

Chernobyl, respectively (Mubayi et al. 1995, p. A-5). Assuming 1,000 workers and a 

4.2 multiplier, an estimate radiation exposure of 14,000 person-rem results.

 March 2011 April 2011 May 2011 March 2011 April 2011 May 2011 March 2011 April 2011 May 2011

≥ 250 460.2 670.4 355.1

200 - 249 224.5 249 212.25

150 - 199 174.5 199 162.25

100 - 149 124.5 149 112.25

50 - 99 74.5 69.3 99 69.3 62.25 62.25

20 - 49 34.5 34.5 34.5 49 49 41.6 27.25 27.25 27.25

10 - 19 14.5 14.5 14.5 19 19 19 12.25 12.25 12.25

< 10 5 5 5 10 10 10 2.5 2.5 2.5

Total Monthly Dose 90,200 23,600 17,000 125,600 42,200 32,100 72,500 14,200 9,400

Avg Worker Dose 24.1 6.5 5.6 33.6 11.7 10.6 19.4 3.9 3.1

High Estimate

Category Radiation Exposure (mSv)

Low Estimate

Category Radiation Exposure (mSv)

Best Estimate

Radiation Exposure Category Radiation Exposure (mSv)

(mSv)
D-21

Site worker exposures following a spent fuel pool accident could be greater than that of a 

reactor core melt accident. This is because a spent fuel pool stores significantly more fuel 

assemblies than a reactor core. Additionally, radionuclides released during a spent fuel pool 

accident have longer half-lives (e.g., Cesium-137) than those that would be released during a 

reactor accident. Given the uncertainties in existing data and variability in severe accident 

parameters and worker response, Table 82 provides the long-term occupational dose used in 

this regulatory analysis to analyze spent fuel pool fires.

Table 82 Long-Term Accident Occupational Exposure for a Spent Fuel Pool Fire

Case Immediate Accident Occupational 

Exposure (averted person-rem)

Low Estimate 4,580

Best Estimate 14,000

High Estimate 46,000

D.3.2.2.8 Long-Term Habitability Criteria

The long-term phase is the period following the seven-day emergency phase and is modeled for 

50 years to calculate consequences from exposure of the average person. Radiation exposure 

during this phase is mainly from external radiation from trace contaminants that remain after the 

land is decontaminated, or in lightly contaminated areas where no decontamination was 

required. Internal radiation exposures may also occur during this period, including inhalation of 

resuspended radionuclides and ingestion of food and water with trace contaminants. 

Depending on the relevant protective action guides (PAGs) and the level of radiation, food, and 

water below a certain limit could be considered adequately safe for ingestion, and lightly 

contaminated areas could be considered habitable.

A long-term cleanup policy for recovery after a severe nuclear power plant accident does not 

currently exist. The actual decisions regarding how land would be recovered and populations 

relocated after an accident would be made by a number of local, state, and federal jurisdictions 

and would most likely be based on a long-term cleanup strategy, which is currently being 

developed by the NRC, EPA, and other Federal agencies. Furthermore, a cleanup standard 

may not have an explicit dose level for cleanup. Instead, the cleanup strategy may give local 

jurisdictions the ability to develop localized cleanup goals after an accident, to allow for a 

number of factors that include sociopolitical, technical, and economic considerations.

Site-specific values are used to determine long-term habitability. For habitability, most states 

adhere to EPA intermediate phase protective action guides that allow a dose of 2 rem in the first 

year and 500 mrem each year thereafter. This habitability criterion was used in previous spent 

fuel pool studies, which used 4 rem in 5 years to represent these PAG levels (e.g., 2 rem in year 

one, followed by 0.5 rem each successive year). However, consistent with the location of the 

reference plant, the Spent Fuel Pool Study analysis utilizes the State of Pennsylvania 

habitability criterion of 500 mrem beginning in the first year (and each following year). The use 

of this long-term habitability criterion reduces the predicted long-term population doses and 
D-22

health effects and increases the costs associated with interdiction, decontamination, and 

condemnation.57

Given the uncertainties in which long-term habitability criterion would be used, Table 83

provides the long-term phase habitability criterion used in this analysis to analyze the 

consequences of spent fuel pool fires on public health (accident).

Table 83 Long-Term Habitability Criterion

Case Long-Term Habitability Criterion Protective Action Basis

Low Estimate 500 mrem annually Pennsylvania dose limit to the public

Best Estimate 2 rem in the first year and 500 

mrem each year thereafter EPA intermediate phase PAGs

High Estimate 2 rem annually EPA intermediate phase PAG: first 

year

Based on the average population dose for a release estimated using a sensitivity analysis, the 

public dose for the two EPA protective action bases was estimated by scaling population dose 

calculated using the Pennsylvania dose limit. The habitability criterion scaling factors used are 

provided in Table 84.

Table 84 Habitability Criterion Scaling Factors

500 mrem

2 rem in the first year 

and 500 mrem each year 

thereafter 2 rem

Population Dose within 50 miles 100% 207% 278%

Total Population Dose 100% 165% 192%

The use of these habitability criteria also affects the values of offsite property damage used in 

this analysis. Certain metrics such as offsite property damage, the number of displaced 

individuals (either temporarily or permanently) and the extent to which such actions may be 

needed are inversely proportional to changes in collective dose resulting from changes in 

habitability criteria.

The impacts for alternate protective action levels were produced by examining the sensitivity 

analyses used to evaluate the effect of alternate protective action levels on land contamination, 

which were based on the results for a release from a high-density loading without credit for 

mitigation during OCP3. Scaling factors for different protective action levels were derived from 

this case. For a very large release that led to economic impacts beyond 50 miles, the sensitivity 

of the results within 50 miles to different protective action levels is less than the sensitivity of 

results beyond 50 miles. For significantly lower release magnitudes associated with the low 

density and successful mitigation cases, the scaling approach used can predict higher economic 

consequences within 50 miles than for the total. This implies that the economic impacts beyond 

50 miles would be small relative to the economic impacts within 50 miles, and the total scaled 

 57 Interdiction and condemnation refer to the relocation of people from contaminated areas according to 

the habitability criterion. Interdiction is the temporary relocation of the affected population while 

decontamination, natural weathering, and radioactive decay reduce the contamination levels. 

Condemnation is the permanent relocation of the affected population if decontamination, natural 

weathering, and radioactive decay cannot adequately reduce contamination levels to habitability limits 

within 30 years.
D-23

economic impact is therefore set equal to the scaled economic impact within 50 miles. The 

economic consequences scaling factors used are provided in Table 85.

Table 85 Economic Consequences Scaling Factors as a Function of Habitability Criteria

500 mrem

2 rem in the first year and 500 

mrem each year thereafter 2 rem

Economic Consequences within 50 miles 100% 67% 56%

Total Economic Consequences 100% 43% 31%

These criteria provide a benchmark for understanding the nature and the extent of the 

relationship between collective dose, economic consequences, and habitability criteria following 

a severe spent fuel pool accident. These measures are subject to large uncertainties, as it is 

difficult to model the impact of disruptions to many different aspects of local economies, the loss 

of infrastructure on the general U.S. economy, or the details of how long-term protective actions 

would be performed.

D.3.2.2.9 Other Key Data

All monetized costs are expressed in 2012 dollars. Ongoing costs of operation related to the 

alternatives are assumed to begin in 2014 unless otherwise stated, and are modeled on an 

annual cost basis.

Estimates were made for one-time implementation costs. The staff assumes that these costs 

will be incurred in the first year of the analysis unless otherwise noted.

Estimates were made for recurring annual operating expenses. The values for annual operating 

expenses are modeled as a constant expense for each year of the analysis horizon. An annuity 

calculation was performed to discount these annual expenses to 2012 dollar values.

Reference plant site population data was projected to year 2011 using the latest version of the 

computer code SECPOP2000. SECPOP2000 uses 2000 census data and applies a multiplier 

value of 1.1051 from the U.S. Census Bureau to account for the average population growth in 

the United States from 2000 to 2011 as discussed in section 7.1.3 of the main report. No further 

population growth was evaluated in this appendix.

D.3.2.3 Assumptions

The Spent Fuel Pool Study is used to inform this analysis is a consequence study based on the 

occurrence of a postulated beyond-design-basis earthquake (with an estimated frequency of 

occurrence of one event in 60,000 years) to a selected U.S. Mark I BWR spent fuel pool with a 

unit-specific spent fuel pool. The Spent Fuel Pool Study major assumptions are listed in 

section 2 of the main document. Additional assumptions used for this analysis are discussed 

below. The costs presented in this analysis are based on estimates by the authors or cited 

documents. It should be noted that this is a generic cost estimate and should be used 

accordingly. Site-specific features may result in higher or lower costs than those estimated.

D.3.2.3.1 Projected Number of Outages and Spent Fuel Assemblies

The reference plant is on a 24-month refueling cycle and is estimated to require eleven refueling 

outages between 2012 and the end of its operating license in 2034. It is assumed that 
D-24

284 assemblies are offloaded to the spent fuel pool during each outage based on information in 

section 5.1 of the main document. The full core of 764 assemblies is offloaded to the spent fuel 

pool upon operating license expiration.

The analysis for the reference plant is based on a high-density spent fuel pool inventory of 

3,055 assemblies in a high-density 1x4 loading configuration, a number based on the pool 

capacity of 3,819 assemblies, reduced by 764 assemblies to accommodate a full core offload 

capability using the existing high-density racking. In a low density 1x4 with empties 

configuration, the spent fuel pool stores 852 assemblies. The number of spent fuel assemblies 

required up to operating license expiration is calculated based on the existing high-density spent 

fuel pool inventory, the number added from refueling outages, and the full reactor core inventory 

and is provided in Table 86.

Table 86 Number of Spent Fuel Assemblies Remaining through Operating License 

Expiration

Category Inventory Number

No. of spent fuel 

assemblies

Current spent fuel pool inventory 3,055 1 3,055

refueling 284 11 3,124

reactor core 764 1 764

Total 6,943

D.3.2.3.2 Dry Storage Capacity

Three companies supply most of the dry storage technologies to U.S. commercial nuclear 

power plants. These companies are Holtec International, Inc. (Holtec), NAC International, Inc. 

(NAC), and Transnuclear, Inc. (Transnuclear). The dry storage cask systems58 (DSCs) for all 

three companies are certified by the NRC for storage of high burnup spent fuel (i.e., burnups 

greater than 45 GWd/MTU), using both regional and uniform loading of spent fuel in the 

packages. A summary of a representative sampling of dry storage canisters commercially 

available to the reference plant for BWR fuel storage is provided in Table 87.

Table 87 Representative Sampling of Commercially Available BWR Spent Fuel Dry 

Storage Technology

Vendor Package Fuel Type Canister 

Type

Capacity 

(Assemblies)

Maximum Decay Heat 

Per Package1 (kW)

Holtec HI-STORM BWR MPC-68 68 34

Holtec HI-STORM FW BWR MPC-89 89 46.36

NAC MAGNASTOR BWR 87B 87 33

Transnuclear NUHOMS BWR 61BTH 61 31.2

Transnuclear TN-68 BWR Bolted 68 30

The maximum decay heat per assembly for uniform loading is estimated by dividing the package decay heat by the 

number of assemblies. The maximum decay heat per assembly under regional loading schemes will generally be 

higher than the maximum decay heat per assembly assuming uniform loading for a smaller number of assemblies. 

Cask certificates of compliance provide the specific maximum assembly decay heat limits for each storage location in 

the basket.

Source: EPRI TR-1025206, p. 2-11.

 58 The term dry storage cask system (DSC) includes dual-purpose canister based systems, dualpurpose casks, and storage-only dry storage casks and canister systems.
D-25

D.3.2.3.3 Fuel Assembly Decay Heat as a Function of Burnup and Cooling Time

As fuel assembly burnups increase, the decay heat of the fuel assembly (watts per assembly) 

increases. Decay heat also can vary significantly with initial enrichment and assembly 

irradiation parameters. Spent fuel burnups have gradually increased since the 1990s with 

average BWR burnups about 43 GWd/MTU and range between 40 and 50 GWd/MTU. Spent 

fuel assembly average decay heat for a 40 GWd/MTU BWR assembly that has cooled for five 

years is approximately 360 watts/assembly. The average decay heat for a 50 GWd/MTU 

assembly that has cooled for five years is approximately 520 watts per assembly (EPRI 

TR-1021049, p. 2-3, Regulatory Guide 3.54). The average BWR spent fuel assembly that has 

cooled for five years is approximately 410 watts/assembly.

Table 88 Canister Storage Capacity Based on Heat Rate Limitations

Based on the average BWR spent fuel assembly that emits 410 watts after it has cooled for five 

years, Table 88 shows that all of the dry storage canisters can be filled to capacity with the 

exception of the NAC MAGNASTOR, without exceeding the maximum decay heat per package 

rating, subject to restrictions on loading pattern. For 50 GWd/MTU assemblies that emit 520 

watts after they have cooled for five years, fewer assemblies can be stored in a cask to ensure 

that it does not exceed the maximum decay heat rating. The number of additional dry storage 

casks required depends on the vendor package selected and range between no additional 

canisters to almost 40% additional canisters. Additional DSCs, which are required because of 

high heat load, are estimated in this appendix. For this regulatory analysis, the Transnuclear 

TN-68 dry casks are evaluated because the reference plant’s ISFSI for dry cask storage utilizes 

the TN-68 cask design as discussed in section 1.3 of the main document. The currently 

approved minimum cooling time for fuel stored in the TN-68 dry casks is seven years (10 years 

for some fuel types), and Transnuclear would need to demonstrate, in an amendment request, 

that spent fuel that was cooled for a shorter period can be stored safely. The costs for 

Transnuclear to prepare such an amendment request and for the NRC review are not included

in this regulatory analysis. The methodology used to estimate the capacity of the DSCs for 

spent fuel that has cooled for five years is subject to uncertainties resulting from decay heat and 

loading pattern restrictions. As a result, the actual DSC capacity may be higher or lower than 

those estimated.

D.3.2.3.4 Dry Storage Upfront Costs

Upfront costs include engineering, design, and licensing costs; equipment costs; construction 

costs; and start up and testing costs. Each of these cost components are further described in 

EPRI TR-1021048, “Industry Spent Fuel Storage Handbook.” As noted in EPRI TR-1025206, 

“Impacts Associated with Transfer of Spent Nuclear Fuel from Spent Fuel Storage Pools to Dry 

Storage after Five Years of Cooling, Rev. 1,” the independent spent fuel storage installation

410w per 

assembly

520w per 

assembly

% Additional 

Canisters

Holtec HI-STORM 68 34 68.00 65.38 4.0%

Holtec HI-STORM FW 89 46.36 89.00 89.00 0.0%

NAC MAGNASTOR 87 33 80.49 63.46 37.1%

Transnuclear NUHOMS 61 31.2 61.00 60.00 1.7%

Transnuclear TN-68 68 30 68.00 57.69 17.9%

Max. Capacity based on decay heat

Vendor Package

Capacity 

(Assemblies)

Maximum Decay Heat 

Per Package1 (kW)
D-26

(ISFSI) upfront costs vary widely from site to site and the upfront costs for those in operation 

vary from several million to tens of millions of dollars. (EPRI TR-1025206, p. 2-23) Values for 

upfront costs were estimated based on two publically available cost estimates that identified the 

specified number of DSC to be stored. The estimate amortized upfront costs for each site is 

provided in Table 89.

Table 89 Amortized DSC Upfront Costs

D.3.2.3.5 Incremental Costs Associated with Earlier DSC Purchase and Loading

Incremental costs are the costs associated with the purchase and loading of DSCs on a periodic 

basis. These costs include the capital costs for the DSC and the loading costs for the storage 

systems. The unit cost estimates used in this analysis are provided in Table 90. These cost 

estimates are based on the DSC unit costs that EPRI used for a Generic Interim Storage Facility 

(EPRI TR-1018722) and documented in EPRI TR-1025206. Operating nuclear power plants 

sites may experience incremental DSC purchase and loading costs that are higher or lower than 

the amount assumed in this analysis.

Table 90 Incremental Unit Cost Estimates

Item Unit Cost

(Constant $2012)

Canister $780,000

Concrete overpack $208,000

Loading of canister-based storage $312,000

Total $1,300,000

D.3.2.3.6 Incremental Annual ISFSI Operating Costs

Annual operating costs for an ISFSI during reactor operation include the costs associated with 

NRC inspections; security; radiation monitoring; ISFSI operational monitoring; technical 

specification and regulatory compliance, including implementation of new certificate of 

compliance (CoC) amendments; personnel cost and code maintenance associated with fuel 

selection for dry storage; personnel costs for spent fuel management and fabrication 

surveillance activities; electric power usage for lighting and security systems; road maintenance 

to the ISFSI site; and miscellaneous expenses associated with ISFSI maintenance. NRC license 

fees for dry storage are included as part of the 10 CFR 50 operating license fees.

Because the reference plant has already implemented dry storage, there are no incremental 

annual ISFSI operating costs expected to implement dry storage at an earlier date if a policy 

decision is made to accelerate the transfer of spent fuel stored in spent fuel pools to dry 

storage. Annual operating costs are a function of when a company begins dry storage. 

ISFSI Facility

Upfront Cost 

Estimate 

(base year)

Upfront Cost Est. 

(2012 $)

DSC Storage 

Capacity

Attributed 

Upfront Cost per 

DSC (2012 $)

Monticello $21.5 million 

(2005 $) $25,275,400 30 $842,500

Pilgrim $22 million 

(2006$) $25,055,800 53 $472,800

Average (Best Estimate) $25,165,600 $657,700
D-27

Therefore, incremental costs associated with annual ISFSI operating costs are insignificant for 

this analysis.

D.3.2.3.7 Dry Storage Occupational Exposure (Routine)

Routine occupational exposure associated with dry storage of spent fuel includes worker dose 

associated with additional DSC loading, unloading and handling activities; additional ISFSI 

operations, maintenance, and surveillance activities; additional DSC storage at an ISFSI; and 

additional transportation cask loading, unloading, and handling activities.

Worker dose associated with DSC loading operations vary depending upon the cask technology 

being loaded, the characteristics of the fuel being loaded (e.g., fuel age and burnup), and fuel 

loading patterns in the DSC (e.g., the location of short-cooled, high burnup spent fuel or colder 

spent fuel within DSC baskets using regional loading). For the regulatory baseline, a worker 

dose of 400 person-mrem per DSC loaded was assumed. This radiation dose is consistent with 

that used in EPRI TR-1021049 and in EPRI TR-1018058, which analyzed worker impacts 

associated with loading spent fuel for transport to the proposed Yucca Mountain repository. 

Some sites achieve per package dose ranges in the range of 200 to 300 person-mrem per 

package loaded, while other sites experience higher per package dose rates. For the 

low-density storage case, each cask loaded in addition to the number required by the regulatory 

baseline is estimated to result in an incremental 400 person-mrem dose.

There is routine occupational dose associated with ISFSI annual operation and maintenance 

activities (i.e., inspection, surveillance, and security operations). The regulatory baseline 

assumes an annual dose of 120 person-mrem per site per year for inspection, surveillance, and 

security activities and 1,500 person-mrem per site per year for ISFSI operations and 

maintenance. These estimated radiation doses are consistent with assumptions used by EPRI 

in EPRI TR-1021049 and TR-1018058. Because additional shielding is assumed to be provided 

by concrete overpacks, the worker dose associated with ISFSI operations and maintenance is 

not expected to increase. Therefore, there is no incremental occupational dose predicted for 

performing annual ISFSI operation and maintenance.

There is routine occupational dose associated with the storage of each DSC at an operational 

ISFSI. The regulatory baseline assumes a worker dose of 170 person-mrem for each additional 

DSC loaded at an ISFSI site. This estimated radiation dose is consistent with assumptions 

used by EPRI in EPRI TR-1021049 and TR-1018058. Because additional shielding is assumed 

to be provided by concrete overpacks, the worker dose associated with each DSC stored at an 

operational ISFSI is not expected to increase. For the low-density spent fuel pool storage case, 

each cask stored in addition to the number required by the regulatory baseline is estimated to 

result in an incremental 170 person-mrem dose.

Table 91 summarizes the occupational dose estimates for each activity.

Table 91 Incremental Occupational Dose (Routine) Estimates

Activity Incremental Occupational Dose (Routine)

(person-mrem per activity)

Load a DSC 400

ISFSI Operation and maintenance 0

Loading a DSC at an ISFSI 170

Total 570
D-28

D.3.2.3.8 Number of Dry Storage Casks

In 2012, the reference plant has 3,819 fuel assemblies stored in the spent fuel pool in a highdensity 1x4 loading configuration. During each refueling outage, 284 assemblies are offloaded 

from the reactor vessel to the spent fuel pool. For the regulatory baseline, the plant is expected 

to load the required number of Transnuclear TN-68 DSCs with a 68-assembly capacity each 

refueling outage to retain sufficient space in the spent fuel pool to discharge one full core of fuel 

(full core reserve). The estimated inventory for use by this regulatory analysis is shown in Table 

92.

Table 92 Regulatory Baseline Loading of Dry Storage Casks

Year

Initial 

SFP 

inventory Refueling

Placed 

into dry 

storage

Final SFP 

Inventory

No. of 

casks 

loaded

Cask 

Capacity

2012 3055 284 -340 2999 5 68

2014 2999 284 -272 3011 4 68

2016 3011 284 -272 3023 4 68

2018 3023 284 -272 3035 4 68

2020 3035 284 -272 3047 4 68

2022 3047 284 -340 2991 5 68

2024 2991 284 -272 3003 4 68

2026 3003 284 -272 3015 4 68

2028 3015 284 -272 3027 4 68

2030 3027 284 -272 3039 4 68

2032 3039 284 -272 3051 4 68

2034 3051 764 0 3815 0 68

2040 3815 0 -816 2999 12 68

2041 2999 0 -816 2183 12 68

2042 2183 0 -816 1367 12 68

2043 1367 0 -680 687 10 68

2044 687 0 -687 0 11 68

Total number of casks 103

At the expiration of the operating license in 2034, the full core is offloaded into the spent fuel 

pool. The analysis further assumes that the entire spent fuel pool inventory will gradually be 

placed into dry storage beginning in 2040 and completed by 2044, 10 years after termination of 

unit commercial operation.

For the low-density spent fuel pool storage case, it is assumed that there is an NRC policy 

decision that requires licensees to offload the spent fuel inventory to dry storage to obtain a lowdensity 1x4 with empties configuration within five years (e.g., by end of 2019). In this 

configuration, the reference plant spent fuel pool stores 852 assemblies (Spent Fuel Pool Study, 

Table 15). Using the same initial conditions as above, and using the DSC with a 57-assembly 

derated capacity beginning in year 2019, the inventory model is provided in Table 93.
D-29

Table 93 Low-density Spent Fuel Pool Case Loading of Dry Storage Casks

Year

Initial SFP 

inventory Refueling

Placed into 

dry storage

Final SFP 

Inventory

No. of casks 

loaded

Cask 

Capacity

2012 3055 284 -340 2999 5 68

2013 2999 0 0 2999

2014 2999 284 -544 2739 8 68

2015 2739 0 -544 2195 8 68

2016 2195 284 -544 1935 8 68

2017 1935 0 -544 1391 8 68

2018 1391 284 -544 1131 8 68

2019 1131 0 -285 846 5 57

2020 846 284 -285 845 5 57

2022 845 284 -285 844 5 57

2024 844 284 -285 843 5 57

2026 843 284 -285 842 5 57

2028 842 284 -285 841 5 57

2030 841 285 -285 841 5 57

2032 841 286 -285 842 5 57

2034 842 764 -798 808 14 57

2043 808 0 -408 400 6 68

2044 400 0 -400 0 6 68

Total number of casks 111

At the expiration of the operating license in 2034, the full core is offloaded into the spent fuel 

pool. The analysis further assumes that the entire spent fuel pool inventory will gradually be 

placed into dry storage beginning in 2043 and completed by 2044, taking only two years 

because of the smaller remaining inventory. Additionally, in years 2038 and 2039, the spent 

fuel has cooled for a sufficient length of time that the DSC is no longer derated.

D.3.3 Sensitivity Analysis

D.3.3.1 Present Value Calculations

Current trends in the marketplace have provided returns on investments well below the 

3 percent and 7 percent discount rates, which OMB Circular No. A-4 is based. The NRC is 

providing a zero discount rate (e.g., undiscounted values) as a sensitivity analyses. Historically, 

regulatory analyses have provided the undiscounted values for the costs and benefits for 

information purposes, but have not provided them as a sensitivity analysis. However, the NRC 

is reporting the undiscounted costs and benefits as part of the sensitivity analysis based on 

current market trends and future predictions.

D.3.3.2 Dollar per Person-Rem Conversion Factor

The NRC is currently revising the dollar per person-rem averted conversion factor based on 

recent information regarding the value of a statistical life (VSL). However, until the NRC 

completes the update and publishes the appropriate guidance documents, the NRC will perform 
D-30

sensitivity analysis to estimate the impact on the calculated results when more current VSL and 

cancer risk factor are used. The NRC used the U.S. Environmental Protection Agency’s (EPA) 

VSL as an interim value in the sensitivity analysis. The EPA’s VSL was developed through a 

rigorous process, reviewing many published academic papers, and includes review from the 

Scientific Advisory Board, an independent review board.

The EPA’s VSL in 2009 dollars is approximately $7.2 million.59 The VSL is derived from “using 

a mixed effects model (random intercept with fixed effects for study characteristics), the authors 

regressed the VSL estimates on average income, probability of death, and several study design 

variables” (EPA, page 41). Therefore, using the CPI-U based inflator to adjust from 2009 

dollars to 2012 dollars yields a VSL of approximately $7.7 million. The International 

Commission on Radiation Protection (ICRP) updated the mortality risk factor in ICRP 

Publication No. 103, the updated risk coefficient is 5 x 10-4

. Using the updated ICRP risk 

coefficient and escalated EPA-based VSL, the dollar per person-rem conversion, rounded to the 

nearest thousand, is $4,000 per person-rem.

Therefore, the NRC will provide the $2,000 per person-rem conversion value for the 

recommendation and the $4,000 per person-rem conversion value as a sensitivity analysis for 

this regulatory analysis.

D.3.3.3 Replacement Energy Costs

The NRC is currently updating its estimates for replacement energy costs based on a 

U.S. competitive electricity market area model. The updated model provides the replacement 

energy costs by day, week, and year, based on market area, in 2010 dollars. For each 

U.S. power market area, a lowest cost and highest cost replacement energy cost estimate was 

calculated, normalizing for reactor megawatt rating differences. The estimated replacement 

energy cost per reactor per year ranges from a high estimate of $54.4 million to a low estimate 

of $692,000 across all U.S. power markets. The average estimated cost per reactor per year 

across all U.S. power markets is $9.6 million and the median estimated cost is $6.4 million in 

2010 dollars. Using the CPI-U inflator formula and the 2010 CPI-U inflator value from Table 77, 

the estimated replacement energy costs range from $57.3 million to $729,000 in 2012 dollars. 

The average estimated cost per reactor per year across all US power markets is $10.1 million 

and the median estimated cost is $6.7 million in 2012 dollars.

D.3.3.4 Consequences Extending Beyond 50 Miles

NUREG/BR-0184 states that in the case of nuclear power plants, changes in public health and 

safety from radiation exposure and offsite property impacts should be examined over a 50-mile 

distance from the plant site. However, in this circumstance it is beneficial for the analysis to 

include supplemental information (e.g., analyses and results) that go beyond the guidance

provided in this document. The Spent Fuel Pool Study uses a plume release model that 

predicts slow deposition of aerosols. This results in public health consequences that extend 

beyond 50 miles from the postulated accident site. While the accuracy of the model decreases 

with distance, the amount of public exposure beyond 50 miles in the event of a release is 

expected to be significant. To capture effects beyond 50 miles, this regulatory analysis 

 59 Environmental Protection Agency, National Center for Environmental Economics, “Valuing Mortality 

Risk Reductions for Environmental Policy: A White Paper”, dated December 2010.
D-31

evaluates the public health and safety and economic consequences estimated by the plume 

model beyond the 50-mile distance from the plant site as a sensitivity analysis.

D.3.4 Alternative – Low-Density Spent Fuel Pool Storage

D.3.4.1 Public Health (Accident)

This attribute measures expected changes in radiation exposure to the public due to change in 

accident frequencies or accident consequences associated with the proposed action. The 

expected changes in radiation exposure are predicted over a 50-mile radius from the plant site. 

The calculated radiation dose to the public is primarily from reoccupation of the land and other 

activities following the spent fuel pool accident. In addition, the calculated radiation dose to the 

public includes the occupational dose to workers for cleanup and decontamination of 

contaminated land not onsite. The incremental radiation doses are calculated by subtracting the 

values for the alternative from those of the regulatory baseline. The difference (delta) is the 

averted dose benefit of this alternative in units of person-rem. The quantitative results for public 

health (accident) considering the contribution of all initiators that could affect spent fuel pool risk 

is provided in Table 94. These values are based on the MACCS2 analyses and probabilistic 

considerations described in further detail in the Spent Fuel Pool Study and other referenced 

documents. The assumptions with regard to the release frequencies are discussed in section 

D.3.2.2.1 and with regard to the habitability criteria are found in section D.3.2.2.8 of this 

regulatory analysis.

Table 94 Summary of Public Health (Accident) for Low-density Spent Fuel Pool Storage 

[All Initiators]

As Table 94 shows, the best estimate of the delta benefit for averted public health (accident) 

radiation exposure from a spent fuel pool accident, which results in spent fuel damage, is 

124 person-rem. This dose represents the reduction of public health risk that results from a 

policy decision to transfer spent fuel from the spent fuel pool to dry storage in order to achieve 

low-density spent fuel loading in the pool at the reference plant. This value is based on a spent 

fuel pool accident that results in an averted delta dose exposure of approximately 

5.6 person-rem per reactor-year over a remaining licensed lifetime of 22 years (until year 2034). 

The best estimate values are based on the reference site’s population density of 722 people per 

square mile within a 50-mile radius from the site and result from the uncontrolled release of 

radionuclides from a full spent fuel pool. The low estimate case reflects the health benefit of a 

spent fuel pool with low-density storage compared to a pool with high-density storage if the 

more stringent Pennsylvania protective action guides are used following an event challenging 

spent fuel pool cooling. The high estimate case reflects the calculated health benefits that result 

if a less stringent 2 rem annual dose protective action guide is used.

A case to evaluate the sensitivity of the results to a change in the dollar per person-rem 

conversion value from $2,000 to $4,000 per person-rem averted was performed. The results of 

this case are provided in Table 95.

Undiscounted

Best Est. Low Est. Best Est. High Est. Low Est. Best Est. High Est.

Low-density storage 60 124 1,260 $247,700 $86,700 $179,500 $1,825,500 $60,200 $124,600 $1,267,000

Case

Dose (averted person-rem) Benefits (2012 dollars)

Low Est. Best Est. High Est. 3% Net Present Value 7% Net Present Value
D-32

Table 95 Sensitivity Analyses of Public Health (Accident) Benefits for Low-density Spent 

Fuel Pool Storage for All Initiating Events (within 50 miles)

Because a spent fuel pool fire under certain scenarios and environmental conditions could result 

in impacts to public health that extend beyond 50 miles, the next two cases evaluate the 

sensitivity of averted public health exposures extending beyond 50 miles from the site. The first 

sensitivity case extends the analysis beyond 50 miles from the plant site and uses the same 

low, best, and high estimate case assumptions for habitability described above and uses the 

standard $2000 per person-rem conversion factor. The second sensitivity case evaluates the 

sensitivity of extending the analysis beyond 50 miles and uses a $4,000 per person-rem 

conversion factor. Table 96 shows the sensitivity on public health (accident) benefits for these 

two cases.

Table 96 Sensitivity Analyses of Public Health (Accident) Benefits for Low-density Spent 

Fuel Pool Storage for All Initiating Events (extending beyond 50 miles)

D.3.4.2 Occupational Health (Accident)

Occupational health measures both short-term and long-term health effects associated with site 

workers as a result of changes in accident frequency or accident mitigation. Within the 

regulatory baseline, the short-term occupational exposure related to the accident occurs at the 

time of the accident and during the immediate management of the emergency and during 

decontamination and decommissioning of the onsite property. The radiological occupational 

exposure resulting from cleanup and refurbishment or decommissioning activities of the 

damaged facility to occupational workers are estimated within the long-term occupational 

exposure. The quantitative results for occupational health (accident) considering the 

contribution of all initiators that could affect spent fuel pool risk is provided in Table 97 and is 

based on the release frequencies discussed in section D.3.2.2.1 and the occupational health 

(accident) assumptions found in section D.3.2.2.7.

Table 97 Occupational Health (Accident) Benefits for Low-density Spent Fuel Pool 

Storage Considering All Initiating Events

Undiscounted

Best Est. Low Est. Best Est. High Est. Low Est. Best Est. High Est.

Dollar per person-rem 

value 60 124 1,260 $495,500 $173,400 $358,900 $3,650,900 $120,400 $249,100 $2,534,000

Low Est. Best Est. High Est.

3% Net Present Value

Case

Dose (averted person-rem) Benefits (2012 dollars)

7% Net Present Value

Low Est. Best Est. High Est. Low Est. Best Est. High Est.

Base case extended 

beyond 50 miles 541 892 7,868 $1,783,450 $783,250 $1,291,900 $11,399,100 $543,650 $896,700 $7,911,700

Dollar per person-rem 

value 541 892 7,868 $3,566,900 $1,566,500 $2,583,800 $22,798,200 $1,087,300 $1,793,400 $15,823,400

Undiscounted

Benefits (2012 dollars)

3% Net Present Value 7% Net Present Value Case

Dose (averted person-rem)

Low Est. Best Est. High Est.

Undiscounted

Best Est. Low Est. Best Est. High Est. Low Est. Best Est. High Est.

short-term 0.3 0.4 5.5 $841 $388 $628 $7,959 $269 $453 $5,524

long-term 0.1 0.2 5.2 $415 $98 $310 $7,490 $68 $223 $5,198

Total 0.3 0.6 10.7 1,260 490 940 15,450 340 680 10,720

Case

Dose (averted person-rem) Benefits (2012 dollars)

Low Est. Best Est. High Est. 3% Net Present Value 7% Net Present Value
D-33

As Table 97 shows, the total delta benefit for short- and long-term occupational health 

(accident) is 0.6 person-rem averted per reactor. The estimated total benefit of the occupational 

health (accident) attribute for low-density spent fuel pool storage relative to the regulatory 

baseline, using the $2,000 per person-rem averted conversion factor, net present value ranges 

are insignificant for this analysis and do not warrant further sensitivity analysis.

D.3.4.3 Occupational Health (Routine)

Occupational health (routine) accounts for radiological exposures to workers during normal 

facility operations (i.e., non-accident situations). These occupational exposures occur during 

DSC loading and handling activities, ISFSI operations, and maintenance and surveillance 

activities. The assumptions in relation to the exposures for occupational health (routine) are 

found in section D.3.2.3.7 of this regulatory analysis.

Table 98 Occupational Health (Routine) Costs for Low-density Spent Fuel Pool Storage

As Table 98 shows, the delta benefit for occupational health (routine) is an increase of 

4.56 person-rem in worker exposure resulting from DSC loading and handling activities; ISFSI 

operations; and maintenance and surveillance activities. The estimated cost to the occupational

health (routine) for low-density spent fuel storage relative to the regulatory baseline and 

calculated in accordance with the current regulatory framework, ranges from $24,000 (3 percent 

net present value) to $27,000 (7 percent net present value) using the $2,000 per person-rem 

averted conversion factor. These ranges are insignificant for this analysis and do not warrant 

further sensitivity analysis.

Low-Density 

SFP Loading

Regulatory 

Baseline Difference

Exposure per 

DSC

Additional 

Dose No Discount 3% NPV 7% NPV

2012 5 5 0 0.57 0 $0 $0 $0

2013 0 0 0 0.57 0 $0 $0 $0

2014 8 4 -4 0.57 -2.28 -$4,560 -$4,298 -$3,983

2015 8 0 -8 0.57 -4.56 -$9,120 -$8,346 -$7,445

2016 8 4 -4 0.57 -2.28 -$4,560 -$4,052 -$3,479

2017 8 0 -8 0.57 -4.56 -$9,120 -$7,867 -$6,502

2018 8 4 -4 0.57 -2.28 -$4,560 -$3,819 -$3,039

2019 5 0 -5 0.57 -2.85 -$5,700 -$4,635 -$3,550

2020 5 4 -1 0.57 -0.57 -$1,140 -$900 -$663

2022 5 5 0 0.57 0 $0 $0 $0

2024 5 4 -1 0.57 -0.57 -$1,140 -$800 -$506

2026 5 4 -1 0.57 -0.57 -$1,140 -$754 -$442

2028 5 4 -1 0.57 -0.57 -$1,140 -$710 -$386

2030 5 4 -1 0.57 -0.57 -$1,140 -$670 -$337

2032 5 4 -1 0.57 -0.57 -$1,140 -$631 -$295

2034 14 0 -14 0.57 -7.98 -$15,960 -$8,329 -$3,602

2040 12 12 0.57 6.84 $13,680 $5,979 $2,058

2041 12 12 0.57 6.84 $13,680 $5,805 $1,923

2042 12 12 0.57 6.84 $13,680 $5,636 $1,797

2043 6 10 4 0.57 2.28 $4,560 $1,824 $560

2044 6 11 5 0.57 2.85 $5,700 $2,214 $654

Total: -4.56 -$9,000 -$24,000 -$27,000

Year

No. of DSCs Dose (person-rem) Costs (2012 dollars)
D-34

D.3.4.4 Offsite Property

The offsite property attribute measures the expected total monetary effects on offsite property 

resulting from the proposed action. Changes to offsite property can take various forms, both 

direct, (e.g. land, food, and water) and indirect (e.g. tourism). This attribute is the product of the 

change in accident frequency and the property consequences from the occurrence of a spent 

fuel pool accident at the reference plant.

For the regulatory baseline, the offsite property costs are any property consequences resulting 

from any radiological release from the occurrence of an accident. Normal operational releases 

and any plant releases not related to the severe accident analyzed are outside the scope of this 

regulatory analysis.

The cost offsets for the analyzed spent fuel pool accident are quantified relative to the 

regulatory baseline based on the MACCS2 calculation results and probabilistic considerations 

provided in the main document. The results for the consequences from a low-density spent 

pool accident are compared to those from the regulatory baseline spent fuel pool accident. The 

calculation is the difference between the calculated consequences resulting from a low-density 

and a high-density spent fuel pool accident and are provided in Table 99.

Table 99 Offsite Property Cost Offsets for Low-density Spent Fuel Pool Storage

As Table 99 shows the estimate of offsite property damage can vary significantly with the 

criterion used to measure or estimate the level of contamination. This regulatory analysis uses 

three protective action levels – the Pennsylvania PAG of 500 mrem annually for the low 

estimate, the EPA intermediate phase PAG level of 2 rem in the first year, and 500 mrem 

annually thereafter for the best estimate, and 2 rem annually for the high estimate – to evaluate 

post-accident collective dose and offsite property costs. As discussed in section D.3.2.2.8, 

offsite property costs are inversely proportional to changes in collective dose resulting from 

changes in habitability criteria (i.e., lower PAG guidelines result in lower collective dose value 

and higher offsite property costs). Furthermore, the high estimate is also affected by the 

bounding assumption used in establishing the high estimate spent fuel pool release frequency 

shown in Table 75. As shown in Table 99 the estimated total cost offsets for the low-density 

storage option relative to the regulatory baseline range from $0.5 to $3.3 million (3 percent net 

present value) and from $0.4 to $2.3 million (7 percent net present value) considering

consequences within 50 miles from the site. As a sensitivity study, the analysis of potential 

consequences was extended beyond 50 miles from the site and were quantified based on the 

MACCS2 model. These estimate results are also shown in Table 99 and result in cost offsets 

approximately 2.5 to 4.6 times greater than those in the base case result.

This analysis does not address potential changes to current methodologies and tools to 

regulatory analysis guidance that may result from applying SOARCA insights and improving 

guidance and analysis tools (such as the MACCS2 computer code) based on up-to-date data in 

Undiscounted

Best Est. Low Est. Best Est. High Est. Low Est. Best Est. High Est.

Base case, consequences 

within 50 miles $723,300 $777,500 $524,000 $3,323,400 $539,700 $363,700 $2,306,700

Sensitivity study, 

consequences extend beyond 

50 miles

$2,139,300 $3,599,100 $1,549,700 $8,393,400 $2,498,000 $1,075,600 $5,825,500

Case

Offsite Property Cost Offsets (2012 dollars)

3% Net Present Value 7% Net Present Value
D-35

addition to advancements in accident consequence assessment knowledge as it relates to this 

attribute.

D.3.4.5 Onsite Property

This attribute measures the expected monetary effects on onsite property, including 

replacement power costs, decontamination, and refurbishment costs, from the proposed action. 

There are two forms of onsite property costs that each alternative must disposition. The first 

type of onsite property costs are the cleanup and decontamination costs for the unit. The 

second type of onsite property costs is the cost to replace the energy from the damaged or 

shutdown unit(s). The cost offsets for low-density spent fuel pool storage are quantified relative 

to the regulatory baseline based on the probabilistic considerations provided in the main 

document and the onsite property estimates described in section D.3.2.2.5.

As stated in section D.3.2.2.6, another unit is co-located on the reference plant’s site. 

Therefore, both units may not operate (e.g., due to significant site damage or contamination 

resulting in high occupational exposure to the undamaged unit) due to the spent fuel pool 

accident. In modeling the replacement energy costs based on this scenario, it is assumed for 

the high estimate that replacement energy would be purchased for both units.

Based on these modeling assumptions, the onsite property results are provided in Table 100.

Table 100 Summary of Onsite Property Cost Offsets for Low-density Spent Fuel Pool 

Storage

As Table 100 shows, based on these calculations, the delta cost offset for probability weighted 

onsite property best estimate ranges from $6,890 (3 percent net present value) to $4,280

(7 percent net present value). Low and high estimates are also provided in Table 100.

D.3.4.6 Industry Implementation

Industry implementation accounts for the projected net economic effect on the affected 

licensees to implement the mandated changes. Costs evaluated for dry storage include upfront 

and incremental DSC capital and loading costs. Additional costs above the regulatory baseline 

are considered negative and cost savings are considered positive. The quantitative results for 

industry implementation are given in terms of expected costs if a policy decision is made to 

accelerate the transfer of spent fuel stored in spent fuel pools to dry storage. These expected 

costs are not frequency weighted. Assumptions used for developing the industry 

implementation cost model are discussed in sections D.3.2.3.2, D.3.2.3.5, and D.3.2.3.6, with 

the results provided in Table 101.

Undiscounted

Best Est. Low Est. Best Est. High Est. Low Est. Best Est. High Est.

Onsite Property - 

Replacement Energy $1,639 $50 $1,091 $117,100 $30 $682 $73,200

Onsite Property - Cleanup, 

Decontamination, Repair, & 

Refurbishment

$8,800 $2,900 $5,800 $132,500 $1,800 $3,600 $82,600

Total $10,440 $2,950 $6,890 $249,600 $1,830 $4,280 $155,800

Case 3% Net Present Value 7% Net Present Value

Onsite Property Cost Offsets (2012 dollars)
D-36

Table 101 Industry Implementation Cost Model for Low-density Spent Fuel Pool Storage

For this analysis, the Transnuclear TN-68 dry casks are evaluated for the best estimate 

because the reference plant’s ISFSI for dry cask storage utilizes the TN-68 cask design as 

discussed in Section 1.3 of the main report. The results provided in Table 102 show that eight 

additional DSCs are needed to store the hotter spent fuel.

Table 102 Industry Implementation Costs for Low-density Spent Fuel Pool Storage

Case

Costs (2012 dollars)

No Discount

3% Net 

Present 

Value

7% Net 

Present Value

Industry 

Implementation -$15,660,000 -$41,820,000 -$46,770,000

Table 102 shows, the incremental costs associated with DSC upfront costs and the earlier 

purchasing and loading of DSCs on a periodic basis. The estimated industry implementation 

costs for low-density spent fuel storage relative to the regulatory baseline and calculated in 

accordance with the current regulatory framework, ranges from $41.8 million (3 percent net 

present value) to $46.8 million (7 percent net present value).

D.3.4.7 Industry Operation

Industry operation accounts for the projected net economic effect due to routine and recurring 

activities required by the proposed alternative. Annual operating costs for an ISFSI during 

reactor operation include the costs associated with NRC inspections; security; radiation 

LowDensity Regulatory 

Baseline Difference

One Time 

ISFSI Mod

Upfront 

costs per 

DSC Purchase 

and Loading No Discount 3% NPV 7% NPV

2012 5 5 0 $657,632 $1,300,000 $0 $0 $0

2013 0 0 0 $657,632 $1,300,000 $0 $0 $0

2014 8 4 -4 $657,632 $1,300,000 -$7,830,528 -$7,381,024 -$6,839,486

2015 8 0 -8 $657,632 $1,300,000 -$15,661,056 -$14,332,085 -$12,784,087

2016 8 4 -4 $657,632 $1,300,000 -$7,830,528 -$6,957,323 -$5,973,872

2017 8 0 -8 $657,632 $1,300,000 -$15,661,056 -$13,509,364 -$11,166,116

2018 8 4 -4 $657,632 $1,300,000 -$7,830,528 -$6,557,944 -$5,217,811

2019 5 0 -5 $657,632 $1,300,000 -$9,788,160 -$7,958,670 -$6,095,574

2020 5 4 -1 $657,632 $1,300,000 -$1,957,632 -$1,545,373 -$1,139,360

2022 5 5 0 $657,632 $1,300,000 $0 $0 $0

2024 5 4 -1 $657,632 $1,300,000 -$1,957,632 -$1,373,044 -$869,212

2026 5 4 -1 $657,632 $1,300,000 -$1,957,632 -$1,294,225 -$759,203

2028 5 4 -1 $657,632 $1,300,000 -$1,957,632 -$1,219,932 -$663,118

2030 5 4 -1 $657,632 $1,300,000 -$1,957,632 -$1,149,902 -$579,193

2032 5 4 -1 $657,632 $1,300,000 -$1,957,632 -$1,083,893 -$505,889

2034 14 0 -14 $657,632 $1,300,000 -$27,406,848 -$14,303,428 -$6,186,086

2040 0 12 12 $657,632 $1,300,000 $23,491,584 $10,267,625 $3,533,186

2041 0 12 12 $657,632 $1,300,000 $23,491,584 $9,968,568 $3,302,043

2042 0 12 12 $657,632 $1,300,000 $23,491,584 $9,678,222 $3,086,022

2043 6 10 4 $657,632 $1,300,000 $7,830,528 $3,132,111 $961,377

2044 6 11 5 $657,632 $1,300,000 $9,788,160 $3,801,105 $1,123,105

Total: -8 Total: -$15,660,000 -$41,820,000 -$46,770,000

Year

No. of DSCs Unit Costs Costs (2012 dollars)
D-37

monitoring; ISFSI operational monitoring; technical specification and regulatory compliance, 

including implementation of new certificate of compliance (CoC) amendments; personnel cost 

and code maintenance associated with fuel selection for dry storage; personnel costs for spent 

fuel management and fabrication surveillance activities; electric power usage for lighting and 

security systems; road maintenance to the ISFSI site; and miscellaneous expenses associated 

with ISFSI maintenance. NRC license fees for dry storage are included as part of the 10 CFR 50 

operating license fees. As discussed in section D.3.2.3.6, incremental costs associated with 

annual ISFSI operating costs are insignificant for this analysis.

Industry operation also includes annual operating costs following reactor shutdown for 

decommissioning, which includes the costs associated with transporting spent fuel offsite. 

These costs were beyond the scope of the evaluation of expedited transfer of spent fuel to dry 

cask storage and are not included in this analysis.

D.3.4.8 NRC Implementation

These costs, if calculated, would further reduce the calculated net benefit for this reference plant 

regulatory and backfit analysis.

D.3.4.9 NRC Operation

These costs, if calculated, would further reduce the calculated net benefit for this reference plant 

regulatory and backfit analysis.

D.3.4.10Other Considerations

D.3.4.10.1 Modeling Uncertainties

There remain significant uncertainties in estimating the frequency of events for natural 

phenomena, which are postulated to challenge spent fuel pool cooling or integrity. There are 

also significant uncertainties in the calculation of event consequences in terms of the dispersion 

and disposition of radioactive material into the site environs. This is due in part to significant 

uncertainties regarding the degree to which topographical features and other phenomena are 

modeled at distances away from the reference plant. Estimating economic consequences also 

includes large uncertainties, as it is difficult to model the impact of disruptions to many different 

aspects of local economies and the loss of infrastructure on the general U.S. economy. An 

example of this is the supply chain disruptions that followed the 2011 Tohoku earthquake and 

subsequent tsunami on Japan or the 2004 Indian Ocean earthquake and tsunami on Thailand.

D.3.4.10.2 Cask Handling Risk

The NRC recognizes that there are costs and risks associated with the handling and movement

of spent fuel casks in the reactor building. These cost and risk impacts, if included in this 

analysis, would further reduce the overall net benefit in relation to the regulatory baseline. 

These effects (e.g., the added risks of handling and moving casks) were conservatively ignored 

in order to calculate the maximum potential benefit by only comparing the safety of high-density 

fuel pool storage relative to low-density fuel pool storage and its implementation costs without 

consideration of cask movement risk.
D-38

D.3.4.10.3 Mitigating Strategies

The release of fission products to the environment from events that may cause the loss of spent 

fuel pool cooling or integrity, such as seismic events, missiles, heavy load drops, loss of cooling 

or make-up, inadvertent drainage or siphoning and pneumatic seal failures, are estimated to be 

approximately 5.5x10-7 per reactor-year without successful mitigation. Operator diagnosis and 

recovery are important factors considered in the development of the event frequencies for these 

events and portions of this evaluation are premised on licensees having taken appropriate 

actions to understand the potential consequences of spent fuel pool accident events and 

develop appropriate procedures and mitigating strategies to respond and mitigate the 

consequences.

The main report evaluated the potential benefits of mitigation measures required under Title 10, 

Code of Federal Regulations (10 CFR), Part 50.54 (hh)(2), which were implemented following 

the September 11, 2001 attacks. These mitigation measures are intended to maintain spent 

fuel pool cooling in the event of a loss of large areas of the plant due to explosions or fire. The 

main report does not consider the post-Fukushima improvements required by NRC and being 

implemented by the plants. These improvements are intended to increase the likelihood of 

restoring or maintaining power and mitigation capability during severe accidents.

The new spent fuel pool level instrumentation required under Order EA-12-051 and the 

mitigation strategies now required under Order EA-12-049, significantly enhance the likelihood 

of successful mitigation because of the following features:

• Portable equipment with redundant sets (e.g., N+1) that is sufficient to supply all 

functions, simultaneously for the entire site, including equipment for the spent fuel pool. 

This portable equipment provides additional protection from seismic events, which are a 

dominant contributor to spent fuel pool risk.

• The mission time for this equipment is indefinite, versus the 12-hour mission time for the 

50.54(hh)(2) equipment.60

• The new EA-12-049 mitigating strategies are capable of being deployed in all modes,

which means that the new strategies can address spent fuel pool cooling issues that 

could occur in any operating cycle phase.

• The new spent fuel pool level instrumentation required under Order EA-12-051, ensures 

a reliable indication of the water level in the spent fuel pool for identification of the 

following pool water level conditions:

 A level that is adequate to support operation of the normal fuel pool cooling 

system

 A level that is adequate to provide substantial radiation shielding for a person 

standing on the spent fuel pool operating deck, and 

 A level where fuel remains covered and actions to implement make-up water 

addition should no longer be deferred.

 60 This section of the regulations deals with the development and implementation of guidance and strategies 

intended to maintain or restore core cooling, containment, and SFP cooling capabilities under the circumstances 

associated with loss of large areas of the plant resulting from explosions or fire.
D-39

• The minimum spent fuel pool makeup flow rate under Order EA-12-049 is set to match 

the design basis heat load for the spent fuel pool, which is typically a full core offload in 

addition to the recently removed fuel from the last refueling outage. This results in a 

lower flow rate than that in NEI guidance for Part 50.54 (hh)(2) equipment and an earlier 

transition to spray, if necessary, due to leaks.

• The method of filling the spent fuel pool is via a connection to the normal spent fuel pool 

makeup system located away from the spent fuel pool floor, reducing the impacts on 

human performance due to potentially adverse environmental conditions (e.g., high 

temperature, humidity, and radiation) following an event.

This additional equipment, strategies, and features provided by Orders EA-12-049 and 

EA-12-051, provide additional accident mitigation capability and would further enhance the 

likelihood of successful mitigation, thereby further reducing the value for the conditional 

probability of release.

D.3.4.10.4 Other Favorable Spent Fuel Loading Configurations

In section 9.2 of the Spent Fuel Pool Study, a sensitivity analysis is provided in which a more 

favorable fuel pattern is considered. In this more favorable pattern, eight cold assemblies 

surround each hot assembly (i.e., 1x8 fuel assembly pattern). Although only a few sensitivity 

analysis were performed using this configuration, the results are promising. The sensitivity 

calculations for the high-density 1x8 fuel pattern showed a shorter time to air coolability (i.e. no 

releases in OCP3). Even for the cases that led to the release of radioactive materials in OCP2, 

the release magnitude was much smaller than for the 1x4 fuel pattern, and comparable to the 

low-density cases. Furthermore, the high-density loading configuration, which allows for 764 

empty cells for a full core offload may result in similar reductions in risk to the low-density 

storage option evaluated without the significant capital costs for implementation. Further 

evaluation of this alternative and possibly other loading configurations for all operating cycle 

phases is recommended.

D.4 PRESENTATION OF RESULTS

This section presents the analytical results, including discussion of supplemental 

considerations, uncertainties in estimates, and results of sensitivity analyses on the overall 

benefits. The results are presented in two different ways, in order to address the differing 

decision criteria between regulatory analyses and backfit analyses (10 CFR 50.109).

D.4.1 Regulatory Analysis

D.4.1.1 Summary Table

Table 103 summarizes the quantified net benefits used to perform a safety goal screening.
D-40

Table 103 Summary of Net Benefits for Low-density Spent Fuel Pool Storage 

Considering All Initiator Events (within 50 miles)

1. nc = not calculated

2. Results are expressed in current dollars (year 2012 dollars) except for the undiscounted 

cases, which are expressed in constant dollars.

As shown in Table 103, the calculated net benefits for requiring low-density spent fuel pool 

storage at the reference plant does not achieve a positive net benefit using the current 

regulatory framework. This means that the calculated licensee costs to implement a low-density 

spent fuel pool storage alternative at the referenced site outweighs the estimated benefits.

Furthermore, for the seismic event analyzed for the Spent Fuel Pool Study, no offsite early 

fatalities are calculated to occur. This result is expected for two main reasons:

1. In comparison to reactors, spent fuel pools have a larger proportion of longer-lived 

radionuclides, which are less likely to cause the significant doses required for acute 

health effects.

2. Despite the large releases for certain predicted spent fuel pool accident progressions, 

the release from the most recently discharged fuel (which contains the shorter-lived 

radionuclides) is predicted to be insufficiently fast and insufficiently large to reach the 

acute thresholds associated with offsite early fatalities. When doses do exceed 

minimum levels for early fatalities, emergency response, as treated in the main report,

effectively prevents any early fatality risk, at least in part because the modeled accident 

progression results in releases that are long compared with the time needed for 

relocation.

In addition, the predicted long-term exposure of the population, which could result in latent 

cancer fatality risk, is also low for the following reasons:

1. The individual latent individual latent cancer fatality risk within 0-10 miles for the studied 

scenarios is predicted to be on the order of 10-10 to 10-11 per year, based on the linear no

threshold (LNT) dose response model.

2. The risk within 10 miles of the analyzed accident is dominated by low dose received at a 

low dose rate. According to alternate dose response models, excluding the uncertain 

effects of low radiation dose could reduce the quantified individual latent cancer fatality 

risk within 10 miles to be approximately 10-14 per year, a reduction of approximately 

3,000 times.

Undiscounted 3% NPV 7% NPV Undiscounted 3% NPV 7% NPV Undiscounted 3% NPV 7% NPV

Public Health (Accident) $247,700 $179,500 $124,600 $119,700 $86,700 $60,200 $2,520,000 $1,825,500 $1,267,000

Occupational Health (Accident) $1,300 $900 $700 $700 $500 $300 $21,300 $15,400 $10,700

Offsite Property $723,300 $524,000 $363,700 $1,073,300 $777,500 $539,700 $4,587,800 $3,323,400 $2,306,700

Onsite Property $10,400 $6,900 $4,300 $4,480 $2,950 $1,830 $378,600 $249,600 $155,800

Total Benefits $982,700 $711,300 $493,300 $1,198,200 $867,700 $602,000 $7,507,700 $5,413,900 $3,740,200

Occupational Health (Routine) -$9,000 -$24,000 -$27,000 -$9,000 -$24,000 -$27,000 -$9,000 -$24,000 -$27,000

Industry Implementation -$15,660,000 -$41,820,000 -$46,770,000 -$15,660,000 -$41,820,000 -$46,770,000 -$15,660,000 -$41,820,000 -$46,770,000

Industry Operation -$730,000 -$252,000 -$64,000 -$730,000 -$252,000 -$64,000 -$730,000 -$252,000 -$64,000

NRC Implementation nc nc nc nc nc nc nc nc nc

NRC Operation nc nc nc nc nc nc nc nc nc

Total Costs -$16,399,000 -$42,096,000 -$46,861,000 -$16,399,000 -$42,096,000 -$46,861,000 -$16,399,000 -$42,096,000 -$46,861,000

Net Benefit -$15,416,000 -$41,385,000 -$46,368,000 -$15,200,800 -$41,228,300 -$46,259,000 -$8,891,300 -$36,682,100 -$43,120,800

Attribute Best Estimate Low Estimate High Estimate
D-41

3. Average individual latent cancer fatality risk is low and decreases slowly as a function of 

distance from the plant. Additionally, the predicted individual risks latent cancer fatalities 

are dominated by long-term exposures to very lightly contaminated areas for which 

doses are small enough to be considered habitable. Therefore, the use of alternate 

dose response models would significantly reduce the quantified latent cancer fatalities 

by at least an order of magnitude.

D.4.1.2 Implementation and Operation Costs

Table 104 Summary of Total Implementation and Operation Costs for Low-density Spent 

Fuel Pool Storage for All Initiator Events

Attribute Costs (2012 dollars in millions)

3% Net Present Value 7% Net Present Value

Occupational Health (Routine) $0.024 $0.027

Industry Implementation $41.800 $46.770

Industry Operation $0.252 $0.064

NRC Implementation nc nc

NRC Operation nc nc

Total $42.096 $46.861

As shown in Table 104, the total estimated costs for the referenced plant unit to achieve and 

maintain a low-density spent fuel pool loading range from $42 million (3 percent net present 

value) to $47 million (7 percent net present value). These costs are dominated by the capital 

costs for the DSCs and the loading costs for the storage systems to achieve low-density storage 

in the spent fuel pool than that required for the regulatory baseline.

D.4.1.3 Total Benefits and Cost Offsets

Table 105 Summary of Total Benefits and Cost Offsets for Low-Density Spent Fuel Pool 

Storage for All Initiator Events

Attribute Benefits and Cost Offsets (2012 dollars in millions)

Undiscounted 3% Net Present Value 7% Net Present Value

Public Health (Accident) $0.12 to $2.52 $0.09 to $1.83 $0.06 to $1.27

Occupational Health (Accident) $0.001 to $0.021 $0.0005 to $0.015 $0.0003 to $0.011

Offsite Property $0.72 to $4.59 $0.52 to $3.32 $0.36 to $2.31

Onsite Property $0.004 to $0.38 $0.003 to $0.25 $0.002 to $0.16

Total $0.85 to $7.51 $0.61 to $5.42 $0.42 to $3.75

The total benefits, which include the public health (accident) and occupational health (accident) 

is summed with the cost offsets, which include offsite property and onsite property relative to the 

regulatory baseline, are shown in the Table 105. The offsite property cost offset is the largest 

contributor to the benefits, of which the majority of those costs occur during the long-term 

phase.

D.4.1.4 Sensitivity Analysis

This section summarizes the results of the sensitivity analyses that were performed as an 

additional consideration in performing safety goal screening for requiring low-density spent fuel 

pool storage at the reference plant.
D-42

D.4.1.4.1 Dollar per Person-Rem Conversion Factor

The NRC is currently revising the dollar per person-rem averted conversion factor based on 

recent information regarding the value of a statistical life. However, until the NRC completes the 

update and publishes the appropriate guidance documents, the NRC performs sensitivity 

analysis to estimate the impact on the calculated results when more current VSL and cancer 

risk factor are used. The NRC used the U.S. Environmental Protection Agency’s (EPA) VSL as 

an interim value in the sensitivity analysis as described in section D.3.3.2. The affect of this 

variable on the calculated results are provided in Table 106.

Table 106 Dollar Per Person-Rem Sensitivity Analysis of Net Benefits for Low-density 

Spent Fuel Pool Storage Considering All Initiating Events (within 50 miles)

1. nc = not calculated

2. Results are expressed in current dollars (year 2012 dollars) except for the undiscounted 

cases, which are expressed in constant dollars.

As shown in Table 106, the dollar per person-rem sensitivity analysis does not achieve a 

positive net benefit when using a person-rem conversion factor twice as large as the conversion 

factor in NUREG-1530.

D.4.1.4.2 Consequences Extending Beyond 50 Miles

The RA Handbook states that in the case of nuclear power plants, changes in public health and 

safety from radiation exposure and offsite property impacts should be examined over a 50-mile 

distance from the plant site, although alternative distances from the plant may be used for 

sensitivity analyses. For this regulatory analysis, supplemental information (e.g., analyses and 

results) based on MACCS2 calculated results, which extends the analysis beyond 50 miles from 

the postulated accident site is provided in Table 107.

Undiscounted 3% NPV 7% NPV Undiscounted 3% NPV 7% NPV Undiscounted 3% NPV 7% NPV

Public Health (Accident) $495,400 $359,000 $249,200 $239,400 $173,400 $120,400 $5,040,000 $3,651,000 $2,534,000

Occupational Health (Accident) $2,600 $1,800 $1,400 $1,400 $1,000 $600 $42,600 $30,800 $21,400

Offsite Property $723,300 $524,000 $363,700 $1,073,300 $777,500 $539,700 $4,587,800 $3,323,400 $2,306,700

Onsite Property $10,400 $6,900 $4,300 $4,480 $2,950 $1,830 $378,600 $249,600 $155,800

Total Benefits $1,231,700 $891,700 $618,600 $1,318,600 $954,900 $662,500 $10,049,000 $7,254,800 $5,017,900

Occupational Health (Routine) -$18,000 -$48,000 -$54,000 -$18,000 -$48,000 -$54,000 -$18,000 -$48,000 -$54,000

Industry Implementation -$15,660,000 -$41,820,000 -$46,770,000 -$15,660,000 -$41,820,000 -$46,770,000 -$15,660,000 -$41,820,000 -$46,770,000

Industry Operation -$730,000 -$252,000 -$64,000 -$730,000 -$252,000 -$64,000 -$730,000 -$252,000 -$64,000

NRC Implementation nc nc nc nc nc nc nc nc nc

NRC Operation nc nc nc nc nc nc nc nc nc

Total Costs -$16,408,000 -$42,120,000 -$46,888,000 -$16,408,000 -$42,120,000 -$46,888,000 -$16,408,000 -$42,120,000 -$46,888,000

Net Benefit -$15,176,000 -$41,228,000 -$46,269,000 -$15,089,400 -$41,165,100 -$46,225,500 -$6,359,000 -$34,865,200 -$41,870,100

Attribute Best Estimate Low Estimate High Estimate
D-43

Table 107 Consequences Extending Beyond 50 Miles Sensitivity Analysis of Net Benefits 

for Low-density Spent Fuel Pool Storage Considering All Initiating Events

1. nc = not calculated

2. Results are expressed in current dollars (year 2012 dollars) except for the undiscounted 

cases, which are expressed in constant dollars.

As shown in Table 107, calculated net benefits for requiring low-density spent fuel pool storage 

at the reference plant do not achieve a positive net benefit for eight of the nine cases presented. 

One case, the undiscounted high estimate, shows a positive net benefit of about $11.3 million, 

which reflects the value and impacts at the time in which they are incurred with no present worth 

conversion. It is informative to compare this value to the other high estimate values of ($22.0 

million) and ($33.0 million), which differ from this case by adjusting these future costs into year 

2012 dollars using 3-percent and 7-percent discount rates as described in section D.3.2.1.2.

D.4.1.4.3 Combined Effect of Consequences Extending Beyond 50 Miles and Dollar per

Person-Rem Conversion Factor

This sensitivity analysis considers all initiating events that can challenge the reference plant’s 

spent fuel pool cooling or integrity while taking into account the combined effects of extending

the analysis of consequences beyond 50 miles from the site and increasing the dollar per 

person-rem conversion value from $2,000 to $4,000 per person-rem averted. The combined 

effects of these two variables on the calculated net benefits are provided in Table 108.

Table 108 Combined Sensitivity Analysis that Analyzes Consequences Beyond 50 Miles 

using a Revised Dollar per Person-Rem Conversion Factor on the Net Benefits for Lowdensity Spent Fuel Pool Storage for All Initiator Events

1. nc = not calculated

2. Results are expressed in current dollars (year 2012 dollars) except for the undiscounted 

cases, which are expressed in constant dollars.

Undiscounted 3% NPV 7% NPV Undiscounted 3% NPV 7% NPV Undiscounted 3% NPV 7% NPV

Public Health (Accident) $1,783,400 $1,291,900 $896,700 $1,081,200 $783,300 $543,600 $15,735,800 $11,399,100 $7,911,700

Occupational Health (Accident) $1,300 $900 $700 $700 $500 $300 $21,300 $15,400 $10,700

Offsite Property $2,139,300 $1,549,700 $1,075,600 $4,968,300 $3,599,100 $2,498,000 $11,586,600 $8,393,400 $5,825,500

Onsite Property $10,400 $6,900 $4,300 $4,680 $3,150 $2,030 $378,600 $249,600 $155,800

Total Benefits $3,934,400 $2,849,400 $1,977,300 $6,054,900 $4,386,100 $3,043,900 $27,722,300 $20,057,500 $13,903,700

Occupational Health (Routine) -$9,000 -$24,000 -$27,000 -$9,000 -$24,000 -$27,000 -$9,000 -$24,000 -$27,000

Industry Implementation -$15,660,000 -$41,820,000 -$46,770,000 -$15,660,000 -$41,820,000 -$46,770,000 -$15,660,000 -$41,820,000 -$46,770,000

Industry Operation -$730,000 -$252,000 -$64,000 -$730,000 -$252,000 -$64,000 -$730,000 -$252,000 -$64,000

NRC Implementation nc nc nc nc nc nc nc nc nc

NRC Operation nc nc nc nc nc nc nc nc nc

Total Costs -$16,399,000 -$42,096,000 -$46,861,000 -$16,399,000 -$42,096,000 -$46,861,000 -$16,399,000 -$42,096,000 -$46,861,000

Net Benefit -$12,465,000 -$39,247,000 -$44,884,000 -$10,344,100 -$37,709,900 -$43,817,100 $11,323,300 -$22,038,500 -$32,957,300

Attribute Best Estimate Low Estimate High Estimate

Undiscounted 3% NPV 7% NPV Undiscounted 3% NPV 7% NPV Undiscounted 3% NPV 7% NPV

Public Health (Accident) $3,566,900 $2,583,800 $1,793,400 $2,162,500 $1,566,500 $1,087,300 $31,471,600 $22,798,200 $15,823,400

Occupational Health (Accident) $2,500 $1,900 $1,400 $1,300 $1,000 $700 $42,700 $30,900 $21,400

Offsite Property $2,139,300 $1,549,700 $1,075,600 $4,968,300 $3,599,100 $2,498,000 $11,586,600 $8,393,400 $5,825,500

Onsite Property $10,400 $6,900 $4,300 $4,680 $3,150 $2,030 $378,600 $249,600 $155,800

Total Benefits $5,719,100 $4,142,300 $2,874,700 $7,136,800 $5,169,800 $3,588,000 $43,479,500 $31,472,100 $21,826,100

Occupational Health (Routine) -$18,000 -$49,000 -$54,000 -$18,000 -$49,000 -$54,000 -$18,000 -$49,000 -$54,000

Industry Implementation -$15,660,000 -$41,820,000 -$46,770,000 -$15,660,000 -$41,820,000 -$46,770,000 -$15,660,000 -$41,820,000 -$46,770,000

Industry Operation -$730,000 -$252,000 -$64,000 -$730,000 -$252,000 -$64,000 -$730,000 -$252,000 -$64,000

NRC Implementation nc nc nc nc nc nc nc nc nc

NRC Operation nc nc nc nc nc nc nc nc nc

Total Costs -$16,408,000 -$42,121,000 -$46,888,000 -$16,408,000 -$42,121,000 -$46,888,000 -$16,408,000 -$42,121,000 -$46,888,000

Net Benefit -$10,689,000 -$37,979,000 -$44,013,000 -$9,271,200 -$36,951,200 -$43,300,000 $27,071,500 -$10,648,900 -$25,061,900

Attribute Best Estimate Low Estimate High Estimate
D-44

As shown in Table 108, calculated net benefits for requiring low-density spent fuel pool storage 

at the reference plant do not achieve a positive net benefit for eight of the nine cases presented. 

One case, the undiscounted high estimate, shows a positive net benefit of about $27.1 million, 

which reflects the value and impacts at the time in which they are incurred with no present worth 

conversion. It is informative to compare this value to the other high estimate values of ($10.6 

million) and ($25.1 million), which differ from this case by adjusting these future costs into year 

2012 dollars using 3-percent and 7-percent discount rates as described in section D.3.2.1.2.

D.4.2 Backfit Analysis

As discussed above, the NRC has determined that the reference plant would not achieve a 

substantial increase in the protection of public health and safety from a change to low-density 

spent-fuel-pool storage. The NRC has therefore determined that imposing a requirement to use 

only low-density spent fuel pool storage at the reference plant would not meet the requirements 

of the backfit rule. However, to ensure that there is a complete discussion of these issues, the 

NRC has drafted an analysis of the costs associated with imposing these requirements as a 

backfit for illustrative purposes. This analysis of the direct and indirect costs of implementing 

the new requirements provides an assessment of the costs associated with imposing these 

requirements and the relative safety benefits in terms of the NRC’s backfit rule. This backfit 

analysis examines the impacts of requiring low-density spent fuel pool storage at the reference 

plant relative to the baseline used in the regulatory analysis, which consists of existing 

requirements including the recently issued orders.

This plant-specific backfit analysis differs from most NRC’s backfit analyses in that the NRC is 

not imposing or proposing to impose any requirements on its licensees. Instead, the NRC is 

assessing the safety benefits and costs of hypothetical requirements that, if implemented, would 

result in the use of low-density spent fuel pool storage and a corresponding increase in on-site 

dry cask storage for the reference plant. An NRC rulemaking to impose requirements like the 

ones analyzed in this appendix would need to include a backfit analysis. This section of the 

appendix provides a discussion of some of the elements that would be analyzed as part of a 

backfit analysis of these requirements. Prior to imposing these requirements through a 

rulemaking the NRC would, at the very least, issue a separate regulatory bases for public 

comment. If it is determined that rulemaking is required, the NRC would issue a proposed rule 

for public comment.

Low-density Spent Fuel Pool Storage Alternative Requirements that Constitutes a PlantSpecific Backfit for the Reference Plant

• All spent fuel assemblies that have cooled for at least five years (older spent fuel 

assemblies) after discharge from the reactor core are expeditiously moved from spent 

fuel pool storage from spent fuel pool storage to dry cask storage.

• The completion of the initial movement of older spent fuel assemblies to dry cask 

storage is achieved within five years of the effective date of the requirement.

• Following each refueling outage, the older spent fuel assemblies stored in the pool shall 

be moved to dry cask storage in a timely manner.

In performing this analysis, the NRC considered the nine factors in 10 CFR 50.109, as 

described in the following subsections.
D-45

D.4.2.1 General Description of the Activity Required at the Reference Plant to Complete 

the Backfit

The alternative would require that the licensee of the reference plant incur upfront costs,

including engineering, design, and licensing costs; equipment costs; construction costs; and 

start up and testing costs, as necessary for their independent spent fuel storage installation to 

accept the dry storage cask systems. The licensee would also need to purchase and load dry 

storage casks on a periodic basis in compliance with the regulatory requirement.

D.4.2.2 Potential Change in the Risk to the Public from the Accidental Offsite Release of 

Radioactive Material

Table 109 Public Health (Accident) Person-Rem Averted

Results are expressed in current dollars (year 2012 dollars) except for the undiscounted cases, 

which are expressed in constant dollars.

If the NRC were to implement the low-density storage proposal, the storage of spent fuel in dry 

storage casks would decrease the accidental offsite release of radioactive material from a 

postulated spent fuel pool accident. As Table 109 shows, dry cask storage at the reference 

plant would decrease the radiation exposure to the public by between 60 and 1,260 person-rem. 

The dose to the public mostly comes from the reoccupation of land after decontamination and 

the exposure to the workers who are decontaminating the public land. This analysis also 

assumes that 0.5% of the public will not evacuate during the accident. This resultant radiation 

dose is included within the public health exposure. As shown in the regulatory analysis, the 

best estimate benefits range from $0.18 million (3 percent net present value) to $0.12 million 

(7 percent net present value). A more in-depth review of the person-rem exposure to the public 

is found in section D.3.4.2.

D.4.2.3 Potential Impact on Radiological Exposure of Facility Employees

Table 110 Facility Employee Exposure

Results are expressed in current dollars (year 2012 dollars) except for the undiscounted cases, 

which are expressed in constant dollars.

If imposed on licensees, these requirements would provide added assurance that nuclear 

industry workers are not subjected to unnecessary radiological or hazardous chemical 

exposures as the result of mitigative and clean-up activities associated with a spent fuel pool 

accident that results in a radioactive release. Storage of spent fuel in dry storage casks would 

decrease the post-accidental offsite radiation exposure to facility employees from a postulated 

spent fuel pool accident. The exposure of facility employees comes from a short-term dose, 

based on the exposure during the accident, and a long-term dose, based on the exposure from 

Undiscounted

Best Est. Low Est. Best Est. High Est. Low Est. Best Est. High Est.

Low-density storage 60 124 1,260 $247,700 $86,700 $179,500 $1,825,500 $60,200 $124,600 $1,267,000

Case

Dose (averted person-rem) Benefits (2012 dollars)

Low Est. Best Est. High Est. 3% Net Present Value 7% Net Present Value

Undiscounted

Best Est. Low Est. Best Est. High Est. Low Est. Best Est. High Est.

accident short-term 0.268 0.421 5.493 $841 $388 $628 $7,959 $269 $453 $5,524

accident long-term 0.068 0.208 5.170 $415 $98 $310 $7,490 $68 $223 $5,198

routine -4.560 -4.560 -4.560 -$9,000 -$24,000 -$24,000 -$24,000 -$27,000 -$27,000 -$27,000

Total -4.224 -3.932 6.103 -$7,744 -$23,514 -$23,063 -$8,552 -$26,662 -$26,324 -$16,278

Case

Dose (averted person-rem) Benefits (2012 dollars)

Low Est. Best Est. High Est. 3% Net Present Value 7% Net Present Value
D-46

the onsite cleanup costs. Facility employees, however, receive additional radiation exposure 

during DSC loading and handling activities, ISFSI operations, and maintenance and surveillance 

activities, resulting in a net increase in radiation exposure as shown in Table 110 for the low and 

best estimates. A more in-depth discussion of the person-rem exposure to facility employees 

can be found in sections D.3.4.2 and D.3.4.3.

D.4.2.4 Installation and Continuing Costs Associated with the Backfit, including the Cost 

of Facility Downtime or the Cost of Construction Delay

Table 111 Installation and Continuing Costs Associated with the Backfit

Case Costs (2012 dollars)

Undiscounted 3% NPV 7% NPV

Implementation 

costs -$15,660,000 -$41,820,000 -$46,770,000

Operation costs -$730,000 -$252,000 -$64,000

Total -$16,390,000 -$42,072,000 -$46,834,000

Implementation and continuing costs include the upfront costs, which include engineering, 

design, and licensing costs; equipment costs; construction costs; and start up and testing costs, 

as necessary, for the reference plant’s independent spent fuel storage installation to accept the 

dry storage cask systems. In addition, the licensee would need to purchase and load dry 

storage casks on a periodic basis in compliance with regulatory requirements. As these actions 

are assumed not to affect normal power operations, there are no assumed replacement energy 

costs or construction delays. A more detailed analysis of the industry implementation and 

operation costs is provided in sections D.3.4.6 and D.3.4.7.

D.4.2.5 Potential Safety Impact of Changes in Plant or Operational Complexity, including 

the Relationship to Proposed and Existing Regulatory Requirements

If imposed on licensees, these requirements are not expected to have a significant effect on 

facility complexity. The scheduling and performance of loading spent fuel assemblies from the 

spent fuel pool into casks and transporting them to the ISFSI would add additional complexity to 

plant operations, especially during the initial 5-year loading phase. The added plant operations 

complexity is not significant and will not substantially affect the reference plant operational 

practices or result in substantial indirect costs. However, should a cask drop accident occur

during plant operation, even though its likelihood is remote, the event could challenge plant 

safety systems in mitigating the consequences.

D.4.2.6 Estimated Resource Burden on the NRC Associated with the Proposed Backfit 

and the Availability of Such Resources.

The establishment of the requirements needed to require the reference plant to move 

expeditiously all spent fuel assemblies that have cooled for at least five years (older spent fuel 

assemblies) after discharge from the reactor core from spent fuel pool storage to dry cask 

storage would require rulemaking. The rulemaking would not result in a substantial increase in 

annual expenditures of agency resources.
D-47

D.4.2.7 Potential Impact of Differences in Facility Type, Design, or Age on the Relevancy 

and Practicality of the Proposed Action

There is no expected significant differentiation in how individual plants would implement the 

requirement to expeditiously move all spent fuel assemblies that have cooled for at least five 

years (older spent fuel assemblies) after discharge from the reactor core from spent fuel pool 

storage to dry cask storage. If imposed on licensees, these requirements do not directly relate 

to the facility type, design, or age.

D.4.2.8 Whether the Proposed Backfit is Interim or Final and, if Interim, the Justification 

for Imposing the Proposed Backfit on an Interim Basis

This consideration is not relevant to the analysis at this time because no requirements are being 

proposed.

D.4.2.9 Other Information Relevant and Material to the Proposed Backfit

Table 112 Summary of Backfitting Net Benefits for Low-density Spent Fuel Pool Storage 

for All Initiator Events (within 50 miles)

1. nc = not calculated

2. Results are expressed in current dollars (year 2012 dollars) except for the undiscounted 

cases, which are expressed in constant dollars.

Table 112 summarizes the described benefits and costs associated with the proposed backfit to 

require the reference plant to expeditiously move all older spent fuel assemblies after discharge 

from the reactor core from spent fuel pool storage to dry cask storage. The analyzed alternative 

would also incur onsite and offsite property cost offsets from an accident. These cost offsets 

are summarized in Table 113

Table 113 Summary of Cost Offsets for Onsite and Offsite Property

1. nc = not calculated

2. Results are expressed in current dollars (year 2012 dollars) except for the undiscounted 

cases, which are expressed in constant dollars.

Undiscounted 3% NPV 7% NPV Undiscounted 3% NPV 7% NPV Undiscounted 3% NPV 7% NPV

Public Health (Accident) $247,700 $179,500 $124,600 $119,700 $86,700 $60,200 $2,520,000 $1,825,500 $1,267,000

Occupational Health (Accident) $1,300 $900 $700 $700 $500 $300 $21,300 $15,400 $10,700

Occupational Health (Routine) -$9,000 -$24,000 -$27,000 -$9,000 -$24,000 -$27,000 -$9,000 -$24,000 -$27,000

Total Benefits $240,000 $156,400 $98,300 $111,400 $63,200 $33,500 $2,532,300 $1,816,900 $1,250,700

Industry Implementation -$15,660,000 -$41,820,000 -$46,770,000 -$15,660,000 -$41,820,000 -$46,770,000 -$15,660,000 -$41,820,000 -$46,770,000

Industry Operation -$730,000 -$252,000 -$64,000 -$730,000 -$252,000 -$64,000 -$730,000 -$252,000 -$64,000

NRC Implementation nc nc nc nc nc nc nc nc nc

NRC Operation nc nc nc nc nc nc nc nc nc

Total Costs -$16,390,000 -$42,072,000 -$46,834,000 -$16,390,000 -$42,072,000 -$46,834,000 -$16,390,000 -$42,072,000 -$46,834,000

Net Benefit -$16,150,000 -$41,916,000 -$46,736,000 -$16,279,000 -$42,009,000 -$46,801,000 -$13,858,000 -$40,255,000 -$45,583,000

Attribute Best Estimate Low Estimate High Estimate

Undiscounted 3% NPV 7% NPV Undiscounted 3% NPV 7% NPV Undiscounted 3% NPV 7% NPV

Offsite Property $723,300 $524,000 $363,700 $1,073,300 $777,500 $539,700 $4,587,800 $3,323,400 $2,306,700

Onsite Property $10,400 $6,900 $4,300 $4,480 $2,950 $1,830 $378,600 $249,600 $155,800

Total Benefits $733,700 $530,900 $368,000 $1,077,800 $780,500 $541,500 $4,966,400 $3,573,000 $2,462,500

Attribute Best Estimate Low Estimate High Estimate

Total Cost Offsets
D-48

Table 114 Combined Sensitivity Analysis of the Backfitting Net Benefits for Low-density 

Spent Fuel Pool Storage for All Initiator Events (extending analysis beyond 50 miles and 

using a Revised Dollar per Person-Rem Conversion Factor)

1. nc = not calculated

2. Results are expressed in current dollars (year 2012 dollars) except for the undiscounted 

cases, which are expressed in constant dollars.

Table 114 summarizes the results of the combined sensitivity analyses that extended the 

backfitting net benefit analysis beyond 50 miles from the plant site and used a higher per 

person-rem conversion factor to monetize averted dose. The analyzed alternative would also 

incur onsite and offsite property cost offsets from an accident. These cost offsets for the 

combined sensitivity analysis are summarized in Table 115.

Table 115 Summary of Combined Sensitivity Analysis Cost Offsets for Onsite and Offsite 

Property

1. nc = not calculated

2. Results are expressed in current dollars (year 2012 dollars) except for the undiscounted 

cases, which are expressed in constant dollars.

D.4.3 Disaggregation

In order to comply with the guidance provided in Section 4.3.2 (“Criteria for the Treatment of 

Individual Requirements”) of the Regulatory Analysis Guidelines, the NRC conducted a 

screening review to ensure that the aggregate analysis does not mask the inclusion of individual 

provisions that are not cost-beneficial when considered individually and not necessary to meet 

the goals of imposing these requirements on the reference plant. Consistent with the 

Regulatory Guidelines, the NRC evaluated, on a disaggregated basis, each new regulatory 

provision expected to result in incremental costs. Based on this screening review, the NRC did 

not identify any requirements needing further consideration. The NRC believes that each of 

these provisions described in section D.4.2 is necessary in the aggregate for the expedited 

transfer of spent fuel to DSCs. However, as noted above, the Commission has not found that 

accelerated transfer to DSCs to provide a substantial safety benefit, nor to be cost justified.

Undiscounted 3% NPV 7% NPV Undiscounted 3% NPV 7% NPV Undiscounted 3% NPV 7% NPV

Public Health (Accident) $3,566,900 $2,583,800 $1,793,400 $2,162,500 $1,566,500 $1,087,300 $31,471,600 $22,798,200 $15,823,400

Occupational Health (Accident) $2,500 $1,900 $1,400 $1,300 $1,000 $700 $42,700 $30,900 $21,400

Occupational Health (Routine) -$18,000 -$49,000 -$54,000 -$18,000 -$49,000 -$54,000 -$18,000 -$49,000 -$54,000

Total Benefits $3,551,400 $2,536,700 $1,740,800 $2,145,800 $1,518,500 $1,034,000 $31,496,300 $22,780,100 $15,790,800

Industry Implementation -$15,660,000 -$41,820,000 -$46,770,000 -$15,660,000 -$41,820,000 -$46,770,000 -$15,660,000 -$41,820,000 -$46,770,000

Industry Operation -$730,000 -$252,000 -$64,000 -$730,000 -$252,000 -$64,000 -$730,000 -$252,000 -$64,000

NRC Implementation nc nc nc nc nc nc nc nc nc

NRC Operation nc nc nc nc nc nc nc nc nc

Total Costs -$16,390,000 -$42,072,000 -$46,834,000 -$16,390,000 -$42,072,000 -$46,834,000 -$16,390,000 -$42,072,000 -$46,834,000

Net Benefit -$12,838,600 -$39,535,300 -$45,093,200 -$14,244,200 -$40,553,500 -$45,800,000 $15,106,300 -$19,291,900 -$31,043,200

Attribute Best Estimate Low Estimate High Estimate

Undiscounted 3% NPV 7% NPV Undiscounted 3% NPV 7% NPV Undiscounted 3% NPV 7% NPV

Offsite Property $2,139,300 $1,549,700 $1,075,600 $4,968,300 $3,599,100 $2,498,000 $11,586,600 $8,393,400 $5,825,500

Onsite Property $10,400 $6,900 $4,300 $4,680 $3,150 $2,030 $378,600 $249,600 $155,800

Total Benefits $2,149,700 $1,556,600 $1,079,900 $4,973,000 $3,602,300 $2,500,000 $11,965,200 $8,643,000 $5,981,300

Attribute

Total Cost Offsets

Best Estimate Low Estimate High Estimate
D-49

D.4.4 Safety Goal Evaluation

Safety goal evaluations are applicable only to regulatory initiatives considered to be generic

safety enhancement backfits subject to the substantial additional protection standard in 10 CFR

50.109(a)(3).

The frequency of damage to the spent fuel is estimated to be range from 7.11x10-7 to 5.39x10-6

per reactor-year when considering all initiators that could challenge spent fuel pool cooling or 

integrity. These values, when compared to a target value of 1x10-4

, which is the quantitative 

health objective for latent cancer fatalities derived using reactor accident characterizations, 

represents a 0.71% to 5.39% of the overall frequency of core damage.

The frequency of a release of radioactive material to the environment is assumed to be the 

same as the frequency of spent fuel damage. The reactor building, which houses the spent fuel 

pool, does not provide a containment barrier similar to the containment structure surrounding 

the reactor core, especially under the conditions postulated to dominate the release of 

radioactive materials from spent fuel.

It is difficult to compare the estimated 7.11x10-7 to 5.39x10-6 per reactor-year release 

frequencies for the postulated spent fuel pool accident when considering all initiators to a target 

value of 1x10-5 per reactor year for a large early release frequency (LERF). The spent fuel pool 

source term is not similar to the core damage (or melt) source term for which the consequences 

of a spent fuel pool accident are predicted to have no early fatalities and public health risk is

dominated by latent cancer risks resulting from long-term exposures. Because the analyzed 

spent fuel accident is a slow progression with at least eight hours before an environmental 

release occurs, and the resultant release is not expected to result in any offsite early fatalities, 

the analysis suggests that the spent fuel pool release does not fall within the definition of a large 

early release. Although this analyzed accident is different from a reactor accident, the spent 

fuel pool estimated release frequencies of 7.11x10-7 to 5.39x10-6 per reactor-year meets the 

1x10-5 LERF guidelines.

Societal risk is based on the statistically expected number of early and latent cancer fatalities. 

The Safety Goals for the Operation of Nuclear Power Plants: Policy Statement (51 FR 28044)

defines the early fatality area calculation as that within one mile from the site boundary. As 

discussed above, the resultant release is not expected to result in any offsite early fatalities. A 

ten-mile radius is defined for calculating latent cancer fatalities. The second quantitative 

objective of the Policy Statement is for the risk to the population in the vicinity of a nuclear 

power plant from an accident at a nuclear power plant should not exceed 0.1 percent of the sum 

of cancer fatality risks resulting from all other causes. Based on recent data 

(http://www.cancer.org/research/cancerfactsfigures/index) the total fatality rate from cancer in 

the U.S. is 580,350 per 315,747,500 persons (http://www.census.gov/popclock/) or a risk of 

1.84x10-3 per year, which results in a safety goal of 1.84x10-6 per year. Using the bounding 

frequency of damage to the spent fuel of 5.39x10-6 per reactor-year, which considers all 

initiators that could challenge spent fuel pool cooling or integrity, and the conditional individual 

latent cancer fatality risk within a ten-mile radius is 4.4x10-4 yields a bounding latent cancer 

fatality risk of 2.37x10-9 of cancer fatality per year. This calculated value of 2.37x10-9 latent 

cancer fatalities per reactor-year associated with a spent fuel pool accident is less than 

represents a 0.13% fraction of the 1.84x10-6 per year societal risk goal value based on the 

calculation area specified in the Safety Goal Policy Statement.
D-50

Therefore, the risk of a spent fuel pool accident at the reference plant appears to meet the 

Safety Goal Policy Statement public health objectives. They also meet the 1x10-5 per reactoryear LERF guideline. Therefore, the Regulatory Baseline is justified for the alternative 

described in section D.2.2 as evaluated for the reference plant.

D.4.5 CRGR Results

This section addresses regulatory analysis information requirements for rulemaking actions or 

staff positions subject to review by the Committee to Review Generic Requirements (CRGR). 

All information called for by the CRGR is presented in this regulatory analysis.

D.5 DECISION RATIONALE

This section presents the decision rationale, including the basis for selection, any decision 

criteria used, the regulatory instrument to be used (if applicable), and the statutory basis for the 

selected regulatory action. The decision rationale is presented in two different ways, in order to 

address the differing decision criteria between regulatory analyses and backfit analyses 

(10 CFR 50.109).

D.5.1 Regulatory Analysis

Table 103 shows that a requirement for low-density spent fuel storage alternative does not 

achieve a cost-beneficial increase in public health and safety for the reference plant using the 

current regulatory framework when all event initiators, which may challenge spent fuel cooling or 

pool integrity, are considered. Furthermore, the three sensitivity studies provided in section 

D.4.1.4 also showed that the low-density spent fuel storage alternative was not cost-justified for 

any of the discounted sensitivity cases.

The NRC believes that there are other considerations discussed in section D.3.4.10 that would 

further reduce the quantified benefits and make the proposed alternative less justifiable. Based 

on the NRC's assessment of the costs and benefits, the agency has concluded that the risk due 

to beyond design basis accidents in spent fuel pools, while not negligible, is sufficiently low that 

the added costs involved with expediting the movement of spent fuel from the pool to achieve 

the low-density fuel pool storage alternative evaluated for the reference plant is not warranted.

D.5.2 Backfit Analysis

The NRC conducted a backfit analysis for the reference plant relative to the backfit 

requirements in 10 CFR 50.109 for illustrative purposes. The NRC does not believe that this 

alternative results in a cost-justified substantial safety enhancement for the reference 

plant. First, the risk of a spent fuel pool accident at the reference plant appears to meet the 

Safety Goal Policy Statement public health objectives. The estimated spent fuel pool accident 

release frequency is also less than the 1x10-5 per reactor-year LERF guideline. Second, the 

cost-justified criteria are not met when evaluating the averted accident consequences within 50 

miles of the site consistent with the regulatory framework. Sensitivity analyses that extend the 

analyses beyond 50 miles also show that the low-density spent fuel storage alternative was not 

cost-justified for any of the discounted sensitivity cases. Therefore, the Regulatory Baseline is 

justified for the alternative described in section D.2.2 as evaluated for the reference plant.

In light of the findings above, the NRC concludes that the quantified safety benefits of the 

proposed requirements that qualify as backfits, considered in the aggregate, would not 
D-51

constitute a substantial increase in protection to public health or safety or the common defense 

and security, and the costs of these requirements would not be justified in view of the increase 

in protection to safety and security provided by the backfits embodied in these requirements on 

the reference plant.

D.5.3 Conclusion

The regulatory screening analysis and the backfitting discussion in this appendix indicate that 

for the reference plant a requirement for low-density spent fuel pool storage, and an associated 

requirement for expedited transfer of spent fuel from the spent fuel pool to meet a low-density 

spent fuel pool storage requirement, are not justified.

The risk due to beyond design basis accidents in the spent fuel pool analyzed in this study, is 

sufficiently low that the added costs involved with expediting the movement of spent fuel from 

the pool to achieve the low-density fuel pool storage alternative are not warranted. While the 

expedited fuel movement alternative evaluated is not cost-beneficial, the report has discovered 

that an alternative 1x8 high-density fuel configuration may have significantly lower costs in 

implementation and potentially similar benefits to the low-density configuration. This alternative 

should be evaluated further, in addition to other possible spent fuel pool loading configurations, 

as part of the regulatory analysis for expedited fuel movement described in SECY-12-0095 to 

evaluate the transfer of spent fuel to dry cask storage for existing and new (future) nuclear 

power plants.
D-52

D.6 REFERENCES

Consequence Study of a Beyond-Design-Basis Earthquake Affecting the Spent Fuel Pool at a 

Selected US Mark I Boiling Water Reactor (Spent Fuel Pool Study), Interim Report 

(Draft), dated June 2012 (ADAMS Accession No. ML12159A180).

Devcan Version 6.6.1, National Cancer Institute (http://surveillance.cancer.gov/devcan/). 

Source: NCHS public use data file for the total U.S., April 2012.

EPRI TR-1018058, “Occupational Risk Consequences of the Department of Energy’s Approach 

to Repository Design, Performance Assessment and Operation in the Yucca Mountain License 

Application,” dated August 2008.

EPRI TR-1018722, “Cost Estimate for an Away-From-Reactor Generic Interim Storage Facility 

(GISF) for Spent Nuclear Fuel,” dated May 2009.

EPRI TR-1021048, “Industry Spent Fuel Storage Handbook,” dated July 2010.

EPRI TR-1021049, “Impacts Associated with Transfer of Spent Nuclear Fuel from Spent Fuel 

Storage Pools to Dry Storage after Five Years of Cooling,” dated 2010.

EPRI TR-1025206, “Impacts Associated with Transfer of Spent Nuclear Fuel from Spent Fuel 

Storage Pools to Dry Storage after Five Years of Cooling, Revision 1, dated 

August 2012.

INPO 11-005, “Special Report on the Nuclear Accident at the Fukushima Daiichi Nuclear Power 

Station, Rev. 0, November 2011.

International Commission on Radiological Protection (ICRP), “1990 Recommendations of the 

International Commission on Radiological Protection,” Publication 60, Ann. ICRP 21 

(1-3), 1991.

_____. 2008. “The 2007 Recommendations of the International Commission on Radiological 

Protection,” Publication 103. Ann. ICRP 37 (2-4), 2008.

Jones-Lee, M.W., “Valuing International Safety Externalities: Does the ‘Golden Rule’ Apply?” 

Journal of Risk and Uncertainty, 29.3:277-287, 2004.

Kiyoshi, Kurokawa, et al. Japan. The National Diet of Japan. “Fukushima Nuclear Accident 

Independent Investigation Commission,” The National Diet of Japan, 2012.

Office of Management of the Budget Circular A-4, “Regulatory Analysis,” issued September 

2003.

Order EA-12-049, “Order Modifying Licenses with Regard to Requirements for Mitigation 

Strategies for Beyond-Design-Basis External Events,” dated March 12, 2012.

Order EA-12-051, “Order Modifying Licenses with Regard to Reliable Spent Fuel Pool 

Instrumentation,” dated March 12, 2012.
D-53

Prassinos, P.G. e al. “Seismic Failure and Cask Drop Analyses of the Spent Fuel Pools at Two 

Representative Nuclear Power Plants,” NUREG/CR-5176, January 1989.

PRM-51-10, “Proposed Amendment to 10 CFR Part 51,” dated August 25, 2006.

PRM-51-12, “Proposed Amendment to 10 CFR Part 51 (Rescinding finding that environmental 

impacts of pool storage of spent nuclear fuel are insignificant), dated March 16, 2007.

SECY-12-0095, “Tier 3 Program Plans and 6-Month Status Update in Response to Lessons 

Learned from Japan’s March 11, 2011, Great Tohoku Earthquake and Subsequent 

Tsunami,” dated July 13, 2012 (ADAMS Accession No. ML12165A089).

SRM-SECY-93-086, “Backfit Considerations,” dated June 30, 1993 (ADAMS Accession 

No. ML003760758).

SRM-SECY-12-0025, “Proposed Orders and Requests for Information in Response to Lessons 

Learned from Japan’s March 11, 2011, Great Tohoku Earthquake and Tsunami,” dated 

March 9, 2012 (ADAMS Accession No. ML120690347).

U.S. Nuclear Regulatory Commission (NRC), 2001. “Recommendations for Enhancing Reactor 

Safety in the 21st Century.”

_____. 1999. “Spent Fuel Heat Generation in an Independent Spent Fuel Storage Installation,” 

Regulatory Guide 3.54, Rev. 1.

_____. 2004. “Regulatory Analysis Guidelines of the U.S. Nuclear Regulatory Commission,” 

NUREG/BR-0058, Revision 4.

_____. 1997. “Regulatory Analysis Technical Evaluation Handbook,” NUREG/BR-0184.

_____. 1989. "Value/Impact Analyses of Accident Preventive and Mitigative Options for Spent 

Fuel Pools,” NUREG/CR-5281 (ADAMS Accession No. ML071690022).

_____. 2005. "Identification and Analysis of Factors Affecting Emergency Evacuations,” 

NUREG/CR-6864.

_____. 1989. “Regulatory Analysis for the Resolution of Generic Issue 82, Beyond Design 

Basis Accidents in Spent Fuel Pools,” NUREG-1353.

_____. 1990. “Backfitting Guidelines,” NUREG-1409.

_____. “Generic Environmental Impact Statement for License Renewal of Nuclear Plants,” Draft 

Report for Comment, NUREG-1437.

_____. 1995. “Reassessment of NRC’s Dollar Per Person-Rem Conversion Factor Policy,” 

NUREG-1530.

_____. 2001. “Technical Study of Spent Fuel Pool Accident Risk at Decommissioning Nuclear 

Power Plants,” NUREG-1738.

_____. 1992. “NRC Labor Rates in the Generic Cost Catalog,” NUREG/CR-4627, Rev. 2.
D-54

Wada, Koji, Toru Yoshikawa, Takeshi Hayashi, and Yoshiharu Aizawa, “Emergency Response 

Technical Work at Fukushima Dai-ichi Nuclear Power Plant: Occupational Health 

Challenges Posed by the Nuclear Disaster,” Occupational and Environmental Medicine 

2012; 69:599-602, April 12, 2012.
E-1

APPENDIX E: SFPS PUBLIC COMMENTS SUMMARY

The Consequence Study of a Beyond-Design-Basis Earthquake Affecting the Spent Fuel Pool 

for a U.S. Mark I Boiling Water Reactor (SFPS) draft report was publicized in a press release on 

June 24, 2013, with a public comment period from July 2, 2013 through August 1, 2013. 

Comments related to the SFPS project covered a wide range of topics. This Appendix provides 

excerpts of the comments received along with NRC responses. All comments are located in the 

NRC’s Agencywide Documents Access and Management System (ADAMS) with the below 

referenced accession numbers.

The comments below are organized by public commenter.

From: Marv Lewis (ADAMS Accession No. ML13196A302, ML13207A141, and ML13211A177):

1) Comment:

“The entire report seems to skirt the issue whether putting spent fuel pools on the roof of a 

reactor is logical or a design error” 

NRC Response:

The elevated pool design used with the boiling water reactor (BWR) Mark I and II 

containment designs provides for the safe transfer of fuel from the reactor and the safe 

storage of fuel. The pool location allows for underwater transfer of fuel from the reactor 

vessel with substantial shielding provided by more than 10 feet of water above the fuel 

during transfer. The storage configuration maintains in excess of 20 feet of water above the 

stored fuel to provide for cooling of the fuel, protection from credible high-energy missiles 

that may fall into the pool, and mitigation of any radiological release that may result from a 

design-basis fuel handling accident. Systems attached to the pool provide for adequate 

cooling and have design features to prevent a substantial loss of coolant inventory under 

accident conditions, such as design-basis earthquakes and pipe breaks. The pool itself 

makes up a small part of the reactor building and, in addition to being constructed from very 

thick reinforced concrete walls with a leak-tight stainless steel liner, is surrounded by the 

reactor building structure, which consists of thick reinforced concrete walls at the level of the 

spent fuel pool and below. The heavy structural members that support the enclosure 

above the refueling floor have been designed to withstand extreme natural phenomena, 

but, for the several facilities in this class that have relatively lightweight sheathing, the 

sheathing has been designed to separate from the enclosure to prevent over-loading the 

structure. Thus, the spent fuel pools have been designed to provide for safe storage and 

transfer of the fuel used in these BWRs.
E-2

2) Comment: 

“Water supply is evaluated only in the proximity of the fuel pool” 

NRC Response:

The study assesses situations where deployment of mitigation is successful, and where 

mitigation is not successful, considering only portable pumping equipment because the 

study assumes that the normal ac-powered systems will be unavailable. The cases in 

Chapter 6 with unsuccessful deployment of mitigation subsume instances where the water 

source (which as the comment indicates is not proximate to the refuel floor) is unavailable 

because of either an event-driven failure or random failure. 

3) Comment:

I applaud the use of color and simple way that seismology is explained. The tables and 

figures are easily read and understood. The problem emerges in that the graphs and 

explanation may be an oversimplification. I do not see many 2 sigma and 6 sigma 

confidence levels on the graphs and figures. If there is any confidence in the figures, data 

and graphs, the confidence levels should be clearly marked. Is the reader to assume the 

data is unassailable or to assume there is no confidence in the data?”

NRC Response:

The USGS 2008 model does not provide the various confidence levels, thus the report 

doesn’t include this information. However, the point estimates provided by this model and 

used in the study account for confidence levels by providing for each peak ground 

acceleration average hazard estimates which are associated with confidence levels above 

the 50th percentile confidence level of the median point estimates. As compared to the 

median estimates, the confidence level percentiles for the average hazard curves increase 

when the uncertainties increase. As shown in Figure 2 (Chapter 3) of the report, the hazard 

curve used in the study predicts average seismic hazards greater than the average hazard 

curves used in previous NRC studies for peak ground accelerations greater than about 0.4g.

4) Comment:

“I respectfully suggest an extension of the comment period”

NRC Response:

Because this is a research study we did not extend the comment period. This research 

study does not authorize any licensee action or set regulatory requirements. This study also 

does not establish any Commission policy. The comment period was appropriate for a 

research study and provided sufficient opportunity to receive comments.
E-3

5) Comment:

“Are there any configurations and accidents that can ignite zirconium if the surface, alloying 

or other parameters are outside of the design due to error?”

NRC Response:

Nuclear fuel is fabricated with a high degree of quality assurance, with very tight design 

tolerances. The only credible scenario under which a zirconium fire is expected to occur in 

the spent fuel pool (SFP) is during accidents that lead to loss of water. This is investigated 

in this report. The air oxidation kinetics models in MELCOR for Zr-based alloys (including 

Zirlo and M5) are based on the research sponsored by NRC and documented in 

NUREG/CR-6846. 

From: Robert Vandenbosh (ADAMS Accession No. ML13190A353):

6) Comment:

“Study terminated after 3 days. One learns from Fukushima that adequate offsite emergency 

response cannot be guaranteed within 3 days” 

NRC Response:

The assumptions regarding offsite support in this study are similar to those used in the 

State-of-the-Art Reactor Consequence Analyses (NUREG-1935). In NUREG-1935, the staff 

reviewed available resources and emergency plans for Peach Bottom (same reference plant 

as for this spent fuel study) and determined that adequate mitigation measures could be 

brought onsite within 24 hours, and be connected and functioning within 48 hours.

The 3 day truncation time was chosen to perform a detailed assessment of the accident 

progression and to capture the release characteristics. In all unmitigated cases studied in 

the report for which a zirconium fire is predicted to occur, fuel uncovery occurs before 48 

hours (see page 81) and radiological releases occur before 72 hours.

In the event of a prolonged severe accident, radiation and other hazards could make 

termination of an ongoing SFP release challenging. The staff included a sensitivity case 

reported in Section 9.8 to investigate releases that continue beyond 72 hours.

7) Comment:

“Too much credit is given to mitigation. It is difficult to predict all scenarios that could 

compromise mitigation.” 

NRC Response:

The study analyzed scenarios with and without successful deployment of mitigation which 

reasonably characterizes the range of possibilities.
E-4

8) Comment:

“Inclusion in tables of cancer results assuming the existence of threshold doses goes 

against recommendations of radiation health professional organizations that LNT should be 

used.”

NRC Response:

Radiation health professional organizations differ in their assessment of how the 

consequences from low dose radiation should be reported. The SFPS uses the Linear No 

Threshold (LNT) model as the base for the reporting of results. The dose truncation 

methodology, introduced in the SOARCA analyses documented in NUREG-1935, is a 

sensitivity analysis. Because the health effects of low dose radiation are uncertain, this 

approach yields insights into how results might change if low dose radiation were less 

effective at cancer induction than predicted by the LNT model.

9) Comment:

“You give an example of a translation of a specific ground movement (0.12) into an 

earthquake magnitude. A similar example for your 0.7 g ground motion would be helpful.” 

NRC Response:

The ground motion at a site is a function of both the magnitude of the earthquake and the 

distance from the source to the specific site as well as other factors such as source 

characteristics, how the regional geology transmits the seismic energy and how local 

geologic effects affect the ground motions. Calculation of the probability of exceeding a 

given peak ground acceleration at the site, accounts for the ground motions that may result 

from all potential seismic sources in the region around the site. Therefore, the calculated 

probability is an aggregate of the contribution of earthquakes with different magnitudes and 

distances to the site. Although a peak ground acceleration at the site is not generally 

associated with a single magnitude earthquake, probabilistic seismic hazard models provide 

an indication of the magnitude and range of the earthquakes that would contribute the most 

to the hazard at a given site. Section 3.3, first paragraph of page 37, of the report provides 

the magnitude and distance of the earthquakes expected to contribute the most to the 0.7g 

event by describing them as earthquakes with magnitude less than about 6.0 at distances of 

less than about 20 km.

From: Union of Concerned Scientists (ADAMS Accession No. ML13210A139):

10) Comment:

“In several tables and figures (e.g., Table 37, Figs. 96 and 97) you use the term "populationweighted" for Individual Latent Cancer Fatality Risk. I don't understand what this means for 

an individual. A search on the term "population-weighted" reveals it is never defined or 

explained in the text of the Report.” 
E-5

NRC Response:

“Population-weighted, individual LCF risk" is the total amount of latent cancer fatalities 

predicted in a specified area, divided by the population that resides within that area. This 

definition has been added to Table 33 of the report. 

11) Comment:

“Most of your High Density results are for a (1 x 4) configuration. But not all reactor pools are 

loaded in this configuration, and the length of time that regulations allow for achieving this 

density is said to be a secret (p. 208). Thus you are emphasizing a risk less than that 

actually associated with some pools.” 

NRC Response:

The calculations in the report were for a specific reference SFP that implements a prearranged fuel loading pattern (see Table 15 and Section 9.3). Even though the specific time 

requirement for achieving the 1x4 arrangement is not publicly available information, the 

sensitivity calculations presented in Section 9.3 show that somewhat higher releases are 

expected in OCP1 and OCP2 for a contiguous fuel pattern, and the large releases for 1x4 in 

OCP3 are comparable to OCP2 (with contiguous fuel loading). The objective of this study 

was not to provide releases for all pools, but to specify a detailed accident progression 

analysis for a specific reference pool with additional sensitivity calculations to provide 

insights on releases from different loading patterns. In fact, Peach Bottom uses a 1x8 

loading pattern, where most plants employ a 1x4 loading pattern.

12) Comment:

“The relative risk of these two storage options was the real need being served by the study, 

but the study utterly fails that goal” 

“That was the central question that needed to be answered by the NRC’s study – what is the 

relative risk from the two onsite spent fuel storage options. That question remains 

shamefully unanswered in the study.”

“...relative risks and identified strategies and tactics that enhanced the risk reductions 

achievable... But this draft study fails to provide that very useful service.”

“This study could have, and should have, provided useful insights to the relative risk of spent 

fuel pool versus dry storage. The final report must remedy this fundamental flaw”

NRC Response:

Both storage options are safe and pose low risk to the public. The objective of this study 

was to compare the consequences of a beyond design basis earthquake on a 

representative spent fuel pool in two configurations, low and high density. The study 

concluded that both high and low density spent fuel configurations are safe, and that 

expedited transfer of spent fuel from the spent fuel pool to dry cask storage is not warranted 

for the SFP studied. Thus, the NRC staff does not agree with the statement that this study is 

fundamentally flawed.
E-6

13) Comment:

“In March 2012, the NRC ordered plant owners (of all plants, not just Peach Bottom Unit 3) 

to install instrumentation to monitor conditions inside the spent fuel pools at their facilities 

(see http://pbadupws.nrc.gov/docs/ML1205/ML12056A044.pdf). At the same time, the NRC 

also ordered plants owners to develop mitigation strategies to provide assurance of 

adequate cooling of reactor cores and spent fuel pools when permanent electrical supplies 

are unavailable for indefinite periods (see 

http://pbadupws.nrc.gov/docs/ML1205/ML12054A735.pdf). The NRC has applied a double 

standard” 

“If the extreme earthquake scenario examined for a single reactor in the NRC’s study is 

sufficiently thorough and bounding, what was the basis for the NRC’s March 2012 orders 

that owners install spent fuel pool instrumentation and provide spent fuel pool cooling 

capabilities during extended power outages of infinite duration?” 

“This NRC study wants us to believe that damage inflicted by an extreme earthquake on a 

highdensity loaded spent fuel pool poses little threat to public health and safety, even if it is 

not successfully mitigated within 3 days”

“The NRC’s March 2012 orders and its June 2013 draft study cannot both be right” 

NRC Response:

Based on this study and previous studies, NRC believes that spent fuel pools protect public 

health and safety. Orders EA-12-051 and EA-12-049 currently being implemented by all

operating U.S. nuclear power plants should serve to further reduce spent fuel pool accident 

risk by increasing the capability of nuclear power plant operators to monitor spent fuel pool 

water inventory and mitigate beyond-design-basis external events. 

14) Comment:

“The study’s assumption that the battle is won or lost within 72 hours contradicts the mission 

time for other scenarios and the experience at Fukushima” 

NRC Response:

Please see NRC response to Comment #6
E-7

15) Comment:

“The 30-day mission time is not applied across the board to every structure, system, and 

component in every scenario. When shorter mission times are applied, they are 

accompanied by solid justifications. The NRC study’s 3-day mission time is an unverified 

assumption.”

“The study’s imposition of the 3-day mission time serves to dismiss other plausible scenarios 

that could cause damage to irradiated fuel in spent fuel pools after 3 days. The study must 

consider longer mission times, and other scenarios that longer mission times permit. Proper 

consideration of longer mission times and other scenarios might show the risk is low – but 

that and that alone would be the proper method for dismissing these scenarios. Dismissing 

them via an arbitrary, unjustified assumption is poor science.”

“Other credible scenarios were summarily dismissed from consideration because they took 

longer than 3 days to play out” 

NRC Response:

The NRC does not agree that the 3-day assumption is arbitrary or unjustified. In this regard, 

please see the response to Comment #6.

16) Comment:

“The extreme earthquake considered in the draft study may represent the fastest way to 

place the public in harm’s way from a spent fuel pool hazard. But to fixate on it and exclude 

other scenarios seems to replicate the tunnel-vision that factored into the March 1979 

meltdown at Three Mile Island Unit 2.” 

NRC Response:

The study is focused on a large seismic event (beyond the design-basis for the plant) 

because past studies (i.e., NUREG-1353 and NUREG-1738) have suggested that this is the 

largest contributor to SFP risk. In addition, in Appendix D (regulatory analysis), the staff also 

used calculated results from previous spent fuel pool studies to extend the applicability of 

this evaluation to include other initiators, which could challenge spent fuel pool cooling or 

integrity.

17) Comment:

“At a time when Americans cannot board a commercial airliner with more than 3 ounces of 

shampoo in a single container and without first removing footwear and all outwear to thwart 

terrorism, this study seems woefully deficient in summarily dismissing any and all acts of 

malice.” 

NRC Response:

This study is a safety study of a severe earthquake impact on a reference SFP. Separate 

security studies were performed after 9/11 and are non-public. The sensitive unclassified
E-8

and classified analyses were made available to the Advisory Committee on Reactor 

Safeguards, the Government Accountability Office, and the National Academy of Sciences.

18) Comment:

“But what if a spent fuel pool only partially drains? In that case, the temperature rise is not 

checked (as shown in Figure 53) or turned around (as shown in Figure 55). Thus, a partially 

drained spent fuel pool represents a greater hazard than a fully drained one. Again, the 

NRC draft study opted for the better of these two choices to consider.” 

“Yet there are other failure modes that may be caused by these lower accelerations, and 

these scenarios should be considered”

“Because a refueling cavity seal has failed in the past due to reasons other than forces 

resulting from an earthquake’s ground motion, it is unjustified to merely assume such failure 

cannot possibly be caused by a seismic event.”

“The study needs to examine the risk from draindown events, such as the one that 

happened at Haddam Neck, initiated by seismic events of less severity than that likely 1 in 

60,000 year before dismissing such scenarios in favor of but one “bounding” case”

NRC Response:

The study did not analyze a partial draindown, which would correspond to a leak on the 

walls of the SFP, because the staff’s structural analysis for the reference pool studied 

showed that the walls are not potential leak locations. As discussed in Section 4 of the 

report the structural analysis predicted the earthquake would cause relative motion between 

the pool walls and the pool floor that would have the potential to result in cracking in the 

reinforced concrete structure and liner tears at the bottom of the walls. The staff found the 

pool structure to be strong enough to prevent liner tearing that could lead to leakage at any 

other location in the pool. Other components, including the refueling cavity gates and the 

piping attached to the SFP, whose failure could have increased the rate of coolant loss,

were evaluated and found to be sufficiently strong and flexible enough to resist the ground 

motion without leakage. 

19) Comment:

“The study must consider scenarios other than that involving a 1 in 60,000 year earthquake 

leading to complete draindown of a spent fuel pool. Just as Three Mile Island demonstrated 

the fallacy of the large-break loss of coolant shield, the extreme earthquake obsession does 

not answer all the relevant questions that must be answered” 

NRC Response:

The study is a plant-specific analysis focused on the large seismic event shown by past 

studies to dominate risk , and is not a comprehensive search for vulnerabilities. The 

objective of this study was to provide a detailed accident progression analysis for a specific 

pool with additional sensitivity calculations to gain insights on potential radioactive releases 

from different loading patterns. 
E-9

20) Comment:

“Page 11 of the draft study states 

“The conditional probability of a Zircaloy cladding fire given a complete loss of water was 

found to be 1.0 for PWRs [pressurized water reactors] and 0.25 for BWRs [boiling water 

reactors] in high-density configurations based on differences in assumed rack 

geometry.” A 1.0 conditional probability means that there’s a 100 percent chance of a 

PWR spent fuel pool fire if it lost water. (Conditional probability literally cannot get any 

higher than that, yet it’s curiously been omitted from the draft study.)” 

“In this, and too many other instances, when faced with multiple choices, the NRC picked a 

nonconservative, non-bounding option.”

“In this specific instance, the NRC must either examine a PWR spent fuel pool scenario in its 

final report or justify excluding a 100 percent conditional probability of fire along with its 

disastrous consequences” 

NRC Response:

The study is a best-estimate analysis that did not intentionally bias the results in a 

conservative or non-conservative direction. The spent fuel pool study used a BWR as a 

reference plant for a detailed accident progression analysis resulting in a predicted range of 

releases. Therefore, any conclusions in the report only apply to the SFP considered in the 

study for a BWR with a Mark I containment. Any reference to PWR in the report is provided 

in the context of past studies in Section 1.7. 

21) Comment:

“The final study must consider full core offloads (at least parametrically) – or the NRC should 

ban them from happening.” 

NRC Response:

The rationale for choosing a "core shuffle" rather than a full core offload is because the 

former is the typical case for BWRs. Emergent core offloads later in the operating cycle are 

not typical (please see Table 3 for additional information). Moreover, the full core's decay 

heat is actually considered during the outage when the reactor and the SFP are 

hydraulically connected. The additional decay heat from the reactor is provided in Table 26 

for OCP 1 and 2.

22) Comment:

“Table 3 of the study states that "Failure of nearby dams is not explicitly addressed." The 

disaster at Fuksuhima has been attributed to flooding from the tsunami caused by the 

earthquake rather than by earthquake damage directly. This study ignores that reality by 

assuming that an earthquake of severe magnitude likely to occur only once every 60,000 

years (and yielding geometric mean acceleration of 0.7 g) has zero chance of causing 

nearby dam(s) to fail.” 
E-10

NRC Response:

As mentioned in Table 3 of the report, failure of nearby dams is not expected to cause 

flooding above grade level at the reference plant. If flooding were to occur, efforts to 

mitigate the accident could be challenged. 

23) Comment:

“The final study must not summarily dismiss the criticality concern associated with a highdensity spent fuel pool” 

NRC Response:

The NRC does not agree that this concern has been summarily dismissed. As discussed in 

Section 2.3, the offsite consequences of a criticality event are believed to be less severe 

than the consequences of a prolonged uncovering of the SFP, and are expected to be 

bounded by the results of this study.

24) Comment:

“It is not apparent from the information in Tables 8 and 9 that removing 55 percent of the 

mass from spent fuel pools has only positive safety margin implications.”

“The NRC must, as a minimum, explain qualitatively why it only considered one 

configuration in its seismic evaluations.” 

NRC Response:

The difference in the weight of the spent fuel assemblies and mass of water between the 

two configurations while not negligible is not expected to lead to significant differences in the 

calculated probability of damage to the liner. The differences in the estimation of liner failure 

probabilities from differences in those quantities, which are small in the context of the total 

loads on the pool, are expected to be within the range of variability of the response 

predictions for these beyond design basis loads. The study assumed the same mass of 

water for the two configurations without reducing this mass to account for the volume of the 

assemblies in either case. This provides an upper bound on the mass of the water in the 

pool for both configurations. 

From: Mack Sim (ADAMS Accession No. ML13211A175):

25) Comment:

“The exclusion of a refueling gate seal failure is unwarranted. There is no argumentation for 

this exclusion in the study (beyond baldly stating such an event is "unlikely") and no study or 

experiment to support this exclusion is cited.” 
E-11

NRC Response:

The description of the gates system provided by the licensee of the reference plant to the 

study team during a site visit as well as engineering drawings shown by that licensee to the 

study team during the same site visit, showed that the gate facing the spent fuel pool has a 

backup gate with an air gap in between the two gates. Those description and drawings also 

showed that both gates have mechanical seals to prevent leakage. Passive mechanical 

means (i.e., do not depend on air pressure, ac power, or dc power) that are unlikely to fail 

under the earthquake conditions provide leaks tighteness by keeping the seals pressured 

between the gates and walls. 

26) Comment:

“Hydrogen production from steam radiolysis (specifically, steam radiolysis taking place in

bubbles on the surface of fuel elements in a pool that has lost circulation cooling) is not 

considered, which leads to overly optimistic predictions [with respect to] probability and 

timing of hydrogen combustion events.” 

NRC Response:

During a severe accident, cladding oxidation modeled in MELCOR is by far the most 

important source of hydrogen generation, and could be about two orders of magnitude 

higher than radiolysis (based on the decay heat levels in the SFP used in this study). For 

the scenarios leading to hydrogen combustion in this study, the time required to get to 

ignition is only a few hours following the onset of significant heatup. In case of boiloff 

scenarios, the time required to generate sufficient hydrogen to cause combustion is much 

longer than the 72 hour truncation time considered, and the significant steam generated 

during boiloff events would act to temporarily inert the building.

From: EDF - France (ADAMS Accession No. ML13225A582):

 

27) Comment:

“the study is performed for a BWR nuclear reactor, can you give us the characteristics of the 

fuel assembly (type, pitch, enrichment, weight) and the name of the nuclear power plant?”

NRC Response:

Table 15 provides the information on the assembly type. The study used Peach Bottom as a 

reference and used all of the available plant specific design and operational data. 

28) Comment:

“we would like to know the irradiation history of the fuel assemblies (burnup, number of 

irradiation cycle and intercycle, lenght of irradiation and intercycle in days)”
E-12

NRC Response:

As indicated in the Response to Comment # 27, this information is proprietary. But such 

information was used to calculate the radionuclides inventories as explained in Section 6.1.5 

of the report. 

29) Comment: 

“the study considers configuration 1x4 and 1x8 (one hot fuel surrounded by 4 or 8 cold fuel), 

according to you is it possible to store spent fuel like these configurations (are all the 

permutations of fuels practically manageable in exploitation)?”

NRC Response:

In fact, Peach Bottom uses a 1x8 loading pattern, where most plants employ a 1x4 loading

pattern. In the study, the base case used Peach Bottom plant data with a 1x4 loading 

pattern. The 1x8 loading pattern used at Peach Bottom was analyzed as a sensitivity and 

had more favorable results, in terms of coolability and radiological releases than a 1x4 

loading pattern, although both are safe arrangements in terms of public health and safety.

30) Comment: 

“for the configuration 1x4 high density, could you give us the spent fuel heat (KW) for hot 

and cold fuel?”

NRC Response:

Table 25 of the report provides a comparison of various decay heat levels (averaged over a 

number of assemblies).

31) Comment: 

“the study concludes that for BWR fuel assemblies, the latent cancer fatal risk is very low. 

According to you, would the conclusion be similar if the study had been performed for PWR 

fuel assemblies? Is BWR more conservative than [PWR] for this study?”

NRC Response:

The results of this study only apply to the reference plant (BWR Mark I) SFP. However, 

Section 10 provides a comparison of the consequences of this study and past spent fuel 

pool analyses (see Table 62 for latent cancer fatalities information). 
E-13

32) Comment: 

“Figure ES-1 shows that for small leak without mitigation measures, maximum release are 

42% Cs for high density and 3.1% for low density. How can you explain that the latent 

cancer fatal risk is similar and very low for the two cases although there is a great release of 

Cs for high density? What are your assumptions for the radiological consequences study on 

the population?”

NRC Response:

The consequence analysis includes evacuating and relocating people. Therefore, the 

predominant outcome of the SFP release is land contamination, rather than public health 

effects, because emergency planning procedures are assumed to be effective in reducing 

dose to the public under both scenarios.

Section 7 of the report provides a more detailed explanation of the study for offsite 

consequences. Section 7.2.2 discusses the individual latent cancer fatality risk.

33) Comment:

“did the study take into account a propagation of zirconium fire (Sandia National Laboratory 

experiment)?.” 

NRC Response:

Yes. The model development was based on validation of MELCOR against the BWR 

zircornium fire experiments as documented in NUREG/CR-7143.

34) Comment:

“could you briefly explain what the 10 CFR 50.54(hh)2 mitigation measures are?”

NRC Response:

Each NRC reactor licensee must develop its own set of strategies to mitigate the effects of 

potential explosions or fires, and such strategies must include those relating to fire fighting, 

the mitigation of fuel damage, and minimizing radiological releases. As discussed further in 

Section 5.3 of the report, to meet the 10 CFR 50.54(hh)(2) requirements, licensees typically 

rely on equipment such as additional electrical power sources and pumps to provide water 

to the SFP for cooling the spent fuel (as was required by NRC orders after the terrorist 

attacks of September 11, 2001). 

35) Comment:

“Finally, in your conclusion you say that 'Analysis also shows that for the scenarios and 

spent fuel pool studied, spent fuel is only susceptible to a radiological release within a few 

months after the fuel is moved from the reactor into the spent fuel pool. After that time, the 

spent fuel is coolable by air'. Can you explain this phenomenon and particularly the 

assumptions that you made to conclude that the spent fuel is coolable by air? What are the 

spent fuel heats after a few month of storage in the pool?”
E-14

NRC Response:

In an unlikely event of a liner failure in the reference plant SFP, the structural analysis in 

Section 4 predicts the leakage to occur at the bottom of the pool. Without successful 

deployment of mitigation, the leakage will lead to an eventual draining of all water in the 

pool. This leads to the natural circulation of air through the fuel assemblies. The timing of a 

potential zirconium fire is based on detailed accident progression analysis documented in 

Section 6 of the report. While releases are predicted for OCP3 (see page 142), a zirconium 

fire is not predicted in OCP4 (see Figures 52-57). Therefore, the spent fuel is coolable by air 

for at least 72 hours in OCP4. Table 25 of the report provides a comparison of various 

decay heat levels (averaged over a number of assemblies).

36) Comment: 

“Is there any containment protection in the pool?”

NRC Response:

The spent fuel pool is open to the refueling room as shown in Figure 42. The refueling room 

is part of the reactor building adjacent to the reactor containment. This building is not 

designed to withstand elevated pressures. Page 107 of the report provides a description of 

the reactor building and the failure pressure criteria for the blowout panels and the roof.

Section 4 of the report provides more details on the reactor drywell and the spent fuel pool.

From AGREE New York (ADAMS Accession No. ML13217A130):

37) Comment:

“The Draft study is not responsive to the Court's specific direction in June 2012 regarding 

what must be thoroughly evaluated.”

“However, the Court decision in June 2012, US Court of Appeals for the District of Columbia

Circuit, concerning NRC's waste confidence decision, required NRC to thoroughly evaluate 

the following:

1. the environmental effects of failing to secure permanent disposal

2. the risks of spent fuel pool leaks

3. the consequences of spent fuel pool fires.”

NRC Response:

The purpose of the SFPS is to determine whether accelerated transfer of older, colder spent 

fuel from the spent fuel pool at a reference plant to dry cask storage significantly reduces 

risks to public health and safety. 

A separate NRC activity, the ongoing Waste Confidence rulemaking, will respond to the 

June 2012 decision by the U.S. Court of Appeals for the District of Columbia Circuit that 

vacated and remanded the NRC’s Waste Confidence Rule. The Waste Confidence 

rulemaking will address the three deficiencies the Court—and the commenter—identified. 
E-15

For more information about the Waste Confidence rulemaking, visit 

http://www.nrc.gov/waste/spent-fuel-storage/wcd.html.

38) Comment: 

“A number of scientific studies have been completed by the Nuclear Regulatory Commission 

and the National Academy of Sciences which document the severity of the problems of 

overcrowded spent fuel pools and the potential catastrophic risks. This body of scientific 

work, which includes that of Allison Macfarlane, the current Commission Chairperson, 

cannot now simply be ignored by the NRC.” 

NRC Response:

The SFPS builds upon numerous past studies of spent fuel pool risk as discussed in Section 

1.7 of the report.

From: Charles Pennington (ADAMS Accession No.ML13217A132):

39) Comment:

“I question the use of such highly conservative “safety analysis” codes (thermal, structural, 

source and release calculations, especially in the dispersion, inhalation, ingestion, direct 

dose rates, population dose, etc. determinations)” 

“It appears there has been no “sanity check” of the individual case results with an actual 

event such as Chernobyl”

“Such inconsistencies and significant modeling conservatisms should be carefully reviewed 

as to whether the displayed results are consistent with the modeling (a good QA review is 

probably in order) and whether they should appear in this report as has been said by others, 

these results are not hyper-conservative; they are wrong.”

“It would seem logical, if modeling is the issue here (such safety analysis models classically 

being unable to project reasonable outcomes, which is especially troubling when used to 

discuss BDB events with the public), that for at least the dose consequence portion of the 

analyses, simple use of Chernobyl correlations would result in still conservative dose 

projections without the need to overstate impacts on the public; I have developed correlation 

models myself for that purpose, and that is the course I would recommend.” 

NRC Response:

The accident progression analysis contains no intentional conservatisms. MELCOR and 

MACCS2 represent state of the art codes for evaluation of accident progression and 

accident consequences, respectively. 

MELCOR is the NRC's best estimate tool for severe accidents analysis, and has been 

validated against experimental data. MELCOR has the capability to mechanistically model 

deposition of aerosols on the structures within the reactor building. This is factored into the 

building decontamination factor as shown in Figure 86. Section 6.1.1 of the report details the 

modeling approach used for this analysis including the experimental basis for the breakaway 
E-16

oxidation kinetics model as well as other required models when modeling SFPs. The 

zirconium fire experiments were used to help validate MELCOR (see NUREG/CR-7143). 

The code’s predictions showed good agreement with the experimental data for the initiation 

and propagation of zirconium fire.

The MACCS2 code has been compared to a number of alternate atmospheric transport and 

deposition codes. These include: Gaussian puff models and a state-of-the-art Lagrangian 

particle tracking code, for estimating concentrations and deposition out to distances as great 

as 100 miles from the point of release. The study was documented in NUREG-6853. 

Generally, MACCS2 performed as well as either of the Gaussian puff models when 

compared with the state-of-the-art Lagrangian code for calculating mean consequence 

results. 

Extrapolation from the results of historical releases of radioactivity to the environment, such 

as the Chernobyl accident, to yield estimates of the exposures and risks of fuel pool 

accidents would require great care in order to yield valid insights. This is due to major 

differences in source term characteristics (e.g., heat content, aerosol characteristics), 

transport characteristics, and exposure factors (such as, shielding factors, dietary habits, 

and protective action strategies). However, the NRC is participating in a number of 

international and domestic efforts underway to benchmark codes against available 

Fukushima data, although releases from spent fuel pools were not experienced and 

therefore not available for comparison. 

40) Comment:

“The fact that the most highly regarded nuclear regulatory body in the world still uses LCFs 

is simply beyond credulity.” 

NRC Response:

Individual latent cancer fatality (LCF) risk is computed to provide a perspective on how the 

accident evaluated in the report could affect societal risks. Such use of LCF risk data is

consistent with the NRC safety goal policy statement.

41) Comment:

“Finally, I would suggest that some part of the report provide more discussion of the doses 

generated by the events and the risk issue, as well as offering the public a comparison of 

the collective doses from this report with the collective doses the public receives every year 

from a selection of non-nuclear industries (out of the 15 to 20 that are most impactful on 

U.S. population doses).” 

NRC Response:

The NRC Safety Goal regarding LCF risk from nuclear power plant operation (i.e., 2x10-6 or 

2 in 1 million per year) is set 1,000 times lower than the sum of cancer fatality risks resulting 

from all other causes (i.e., ~2x10-3 or 2 in 1 thousand per year). As discussed further in the 

report’s Executive Summary, this study estimated that the likelihood of a radiological release 

from the spent fuel pool resulting from the selected severe seismic event the study analyzed 

is on the order of one time in 10 million years or lower. This in part is a reason why the LCF 

risks within 10 miles in this study are low, in the range of one in a trillion (10-12) to 1 in 10 
E-17

billion (10-10) per year for the analyzed scenario. Comparisons of the individual LCF risk 

within 10 miles calculated in this study to the NRC Safety Goal are provided in Figure ES-3. 

Such comparisons provide context that may help the reader to understand the contribution 

to cancer risks from the accident scenarios that were studied. The results of this study are 

scenario-specific and related to a single spent fuel pool. However, staff concludes that 

since these risks are several orders of magnitude smaller than the 2x10-6 (2 in 1 million) 

individual LCF risk – which corresponds to the safety goal for latent cancer fatalities – it is 

unlikely that the results here would contribute significantly to a risk that would challenge the 

Commission’s safety goal policy.

From: Electric Power Research Institute (ADAMS Accession No. ML13217A133):

42) Comment:

“The SFPS is a consequences analysis as opposed to a full risk assessment. This is 

highlighted and acknowledged in the title as well as several times throughout the report. A 

more comprehensive probabilistic evaluation, while more resource intensive, would provide 

greater refinement and insight and increase the usefulness of the study, such as in riskinformed regulatory activities.” 

NRC Response:

This study is intended to provide a detailed assessment of the consequences associated 

with spent fuel pool accidents, and to compare the consequences of high and low density 

pool configurations. Although the study did not examine all scenarios typically considered in 

a probabilistic risk assessment (PRA), it focused on a large seismic event (beyond the 

design-basis for the plant) because past studies have suggested that this is the largest 

contributor to SFP risk. In fact, any analytical technique, including PRA, will have inherent 

limitations of scope and method. Nevertheless, the staff is currently performing a 

comprehensive site Level 3 PRA for a US PWR as discussed in SECY-11-0089.

43) Comment:

“The SFPS acknowledges the potential for interactions between the reactor and the spent 

fuel pool, and identifies such interactions in Section 2.2. These interactions are not, 

however, evaluated within the scope of the study. This decoupling of the SFP from the 

reactor affects several parameters including the potential for hydrogen from a reactor 

accident to collect in the SFP area, with the possibility of combustion; the timing of assumed 

operator actions; and the timing of the general emergency order (potential to affect 

evacuation). Additional discussion or basis would be helpful.” 

NRC Response:

Although such interactions between the reactor and SFP may be significant, they are 

outside the scope of the current study. The possibility that a concurrent reactor event may 

preclude operator actions is part of the motivation for the "unmitigated" cases analyzed in 

the study. Further, Section 9.4 presents sensitivity calculations to show the importance of 

the reactor building in the progression of accidents in the SFP and the source term with a 

concurrent reactor accident. Moreover, the full core's decay heat is actually considered 

during the outage when the reactor and the SFP are hydraulically connected.
E-18

Additionally, as noted above, the staff is currently performing a comprehensive site Level 3 

PRA for a US PWR as discussed in SECY-11-0089.

44) Comment: 

“The report assumes a uniform spray flux of water over the top of the SFP. Depending on 

timing and locations at which operator actions would need to be taken (e.g., proximity to 

pool and radiation fields), it may be difficult to align the spray. It could also be difficult to 

achieve 100% coverage if debris has blocked the top of some fuel assemblies.” 

NRC Response:

The study analyzed scenarios with and without successful deployment of mitigation and thus 

reasonably characterizes the range of possibilities. Some aspects of the mitigation success 

are provided in Section 8 in the HRA analysis. As stated in Table 3 of the report, no 

significant debris generated by the seismic event is expected to enter the SFP based on the 

structural response of the building and overhead crane. In addition, as stated in Table 3, 

some debris could be generated and could fall into the pool as a result of hydrogen 

combustion. However, the occurrence of a hydrogen combustion event in this study denotes 

that the fuel in the SFP has already become uncovered and is undergoing a fission product 

release.

45) Comment: 

“The option to spray into the SFP from outside the Reactor Building is possible if the 

refueling floor is open to the environment. Depending on the size of the assumed opening, 

this option could limit the ability to achieve full spray coverage of the pool, but might be more 

likely to result in atmospheric scrubbing of the radionuclides” 

“Spraying into the refueling floor and spent fuel pool from an external position creates the 

potential for direct release to the environment that is not currently included. Any operator 

actions to ventilate the building are considered should be considered in this context as well.” 

“The report indicates that “for PBAPS, the capacities of the available equipment are 

somewhat higher. The use of ... 200 gpm here attempts to account for uncertainties in the 

... spray that goes outside the boundary of the pool”. However, it is not clear whether this 

particular conservatism could balance the potential lack of full spray effectiveness noted 

above.” 

NRC Response:

The study analyzed scenarios with and without successful deployment of mitigation which 

reasonably characterizes the range of possibilities. Section 8 in the HRA analysis provides 

some aspects of the mitigation success.

Regarding the comment on considering operator action to ventilate the building, and the 

need to capture the associated direct radiological release to the environment that might 

accompany this, the study did not consider negative or positive impacts of operator actions 

associated with ad hoc measures to ventilate the building. This is in keeping with the 

approach of focusing on mitigation measures covered by 10 CFR 50.54(hh)(2) (and more 
E-19

specifically those in the NRC-endorsed guidance document NEI-06-12, Revision 2). When 

modeled actions or phenomenologically-driven events did affect radiological release 

characterization (e.g., spray scrubbing, hydrogen combustion-induced building damage), 

these effects were mechanistically modeled.

46) Comment:

“Air cooling of the assemblies is dependent upon having clear channels for air flow. Debris 

from the Reactor Building superstructure or miscellaneous equipment on the refueling floor 

could conceivably fall into the pool, potentially damaging the liner or impeding local air flow 

in the SFP channels. The analysis does not consider debris interaction with the SFP 

resulting either from a large seismic event or from a hydrogen combustion event (see 

Section 1, Table 3 for technical basis). Additional technical justification or a sensitivity 

analysis may be desirable. We do note that experience from recent large seismic events, 

including Kashiwazaki-Kariwa 2007, Fukushima Daiichi, Fukushima Daini, Onagawa, and 

North Anna, indicates little or no debris generation and deposition into SFPs due to 

earthquake itself.”

NRC Response:

The occurrence of a hydrogen combustion event associated with the spent fuel itself in this 

study denotes that the fuel in the SFP has already become uncovered and is undergoing a 

fission product release. The reduction in flow area and losses associated with debris 

generated from a hydrogen combustion resulting from a reactor accident is explicitly 

considered in Section 9.4 of the report. 

No significant debris generated by the seismic event is expected to enter the SFP based on 

the structural response of the building. In addition, as stated on page 25, some debris could 

be generated and could fall into the pool as a result of hydrogen combustion.

47) Comment:

“The reports states that “The cool gases leaving the lower regions of the building are not 

brought into thermal equilibrium with gases above the SFP.” (Page 107). This assumption 

allows for cooler air at the inlet of the fuel. In order to achieve the lower inlet temperatures 

other actions may be necessary such as opening doors or other access ways to enhance 

airflow. It may be useful to provide additional justification or sensitivity analysis.” 

“Only the refueling floor is modeled in MELCOR (pp. 105, 107, 108). The equipment shaft 

from lower Reactor Building elevations is modeled as supplying the air inlet to the SFP. This 

leads to air temperatures lower than what might be expected. In practice, air ingress into the 

SFP racks may be limited by the flow area available into the bottom of the pool and the air 

temperature may be higher than assumed. A sensitivity evaluation may be useful to address 

this issue.” 

NRC Response:

The analysis has shown that during the heatup of the fuel and especially during natural 

circulation once the baseplate clears, the atmosphere of the refueling room becomes very 

hot and the flow is out through the normal leakages and the open hatch region to maintain 

the pressure. If the reactor building is intact, oxygen is eventually depleted that would limit 
E-20

the oxidation of remaining cladding. If ventilation is established in time before significant 

heatup of the fuel, it is possible to prevent a zirconium fire; however, if ventilation comes 

late, then air ingress would supply additional oxygen that would lead to additional oxidation. 

Some of these insights can be inferred from Figures 130 and 131 in Section 9.4 of the 

report. 

48) Comment: 

“The crack in the concrete and the tearing of the liner are assumed to occur at the bottom of 

the SFP. The complete draining of the pool opens the possibility for air cooling, which could 

be reduced if the leak occurred above the bottom of the fuel. The impact of this assumption 

can be estimated where fuel assembly inlet flow is delayed, which results in higher fuel 

temperatures. Further discussion of this particular assumption would be useful.” 

NRC Response:

Please see the response to Comment #18. 

49) Comment:

“The SFPS acknowledges the large uncertainties for several assumptions associated with 

location and size of the breach (e.g., finite element modeling of the concrete structure, 

modeling of the liner strain, modeling of the liner tear and the concrete cracking that would 

allow leakage flow, and calculation of the friction factor associated with the leakage 

pathway). Sensitivity calculations could be used to investigate the possible impact of 

changes to failure location or size. A more probabilistic treatment could produce more 

refined results.” 

NRC Response:

The size and location of the leak were not assumptions made by the staff in the study. As 

discussed in Section 4 of the report, the structural analysis of the reference spent fuel pool 

design predicted the liner tear to be at the bottom of the pool for two leak sizes with a range 

of draindown times. For the SFP studied the uncertainty on the location of the concrete 

cracking and liner tearing is small. The study acknowledges that uncertainties on the water 

leakage rate are high. Consideration of two significantly different leakage rates and 

associated draindown times addresses the consequential range of those uncertainties. 

Please also see the response to Comment # 18.

50) Comment: 

“The study predicts cesium releases following hydrogen combustion (approximately 18 

hours into the accident) in high density loading but not in low density loading. The difference 

in outcome is associated with the higher hydrogen concentration in the base case (high 

density loading).” 

NRC Response:

The NRC staff agrees. None of the low density cases produced enough hydrogen to initiate 

a hydrogen deflagration even though sufficient oxygen was present to sustain combustion.
E-21

51) Comment: 

“The MELCOR calculation does not predict that hydrogen combustion will take place on the 

refueling floor. This result may be influenced by not including the entire reactor building in 

the model. A sensitivity study that accounts for sufficient oxygen from other parts of the 

building may produce additional insights.” 

NRC Response:

For high density cases, MELCOR predicts hydrogen deflagration in the refueling bay (see 

the sensitivity to hydrogen combustion in Section 9.1 on the report).

52) Comment: 

“The SFPS “did not consider hydrogen events originating from a concurrent reactor 

accident”.” 

NRC Response:

Hydrogen combustion is considered as a part of the sensitivity analysis in Section 9.4 of the 

report.

53) Comment: 

“Molten core concrete interactions (MCCI) at the bottom of the pool can result in higher 

release fractions for radionuclides. MCCI can also lead to hydrogen generation. While the 

sequence is very unlikely, some discussion in the analysis may be warranted.” 

NRC Response:

The sensitivity with MCCI in Section 9.5 of the report was focused on the releases and 

chemical form of the radionuclides. This sensitivity assumed that the reactor building has 

already failed as a result of a concurrent reactor accident. The limitations of representing 

MCCI in SFPs using MELCOR are also discussed on page 222 of the report. Nevertheless, 

the large scale fuel damage and relocation occurs mainly for small leak scenarios, and in 

most cases, the reactor building is expected to fail as a result of the hydrogen generated 

from the spent fuel before the start of MCCI. 

54) Comment: 

“If the conditions of the reactor do not warrant prompt declaration of a General Emergency 

(GE) based on the emergency action limits (EALs), then the time to initiate public evacuation 

could be delayed. Generally, EALs do not include guidelines for declaring a GE for events 

involving the SFP. For irradiated fuel or SFP events the highest EAL is alert, although a GE 

is declared by the emergency director.” 

NRC Response:

In general, staff acknowledges that uncertainty can be associated with the timing of 

protective actions (such as sheltering and evacuation). However, staff expects the specific 
E-22

site conditions assumed (those being the loss of all off-site and on-site ac power that is not 

expected to be restored for a prolonged period of time) would promptly warrant declaration 

of a GE, and that this would occur before other potential conditions arose. Also, staff notes 

that the largest releases from the baseline results in this study arise from slow leaks, and 

these releases offer relatively long periods (40 hours or more) prior to the onset of 

release. Because of these reasons and others, the results of this analysis are not expected 

to be very sensitive to the exact time of the GE declaration.

55) Comment: 

“The conclusion that there are no calculated early fatalities would benefit from further 

discussion” 

NRC Response:

Staff has identified some of the factors that, in its professional judgment, contribute to the 

modeled result of no early fatalities. Staff believes that this is due in part to the efficacy of 

protective actions such as sheltering, evacuation, and relocation, although staff 

acknowledges that other factors, such as the isotopic composition and duration of the 

release, may play an important role as well. 

Staff believes that the results of the study are consistent with the NUREG-1738. For 

example, Appendix A4B of NUREG-1738 provides computed early fatalities as a function of 

time since shutdown for both high and low ruthenium source terms and for early vs. late 

evacuation. The largest numbers of early fatalities reported in Appendix A4B of NUREG1738 are associated with releases shortly after shutdown, a 75% Ru release, and late 

evacuation of 95% of the population. The NUREG-1738 results for the lower Ru release 

and early evacuation all show substantially less than one early fatality. 

Moreover, with respect to the potential effect of evacuation uncertainty on computed early 

fatalities, staff notes that detailed outputs of SFPS’s consequence simulations include a 

cohort representing the 0.5 percent of the population that is assumed to not evacuate, 

although they are assumed to be subject to dose-dependent relocation following plume 

passage. These detailed results show that even for those who are not assumed to 

evacuate, no early fatalities are computed for any of the releases evaluated. Staff believes 

this result supports the view that uncertainties in the evacuation model are not likely to 

significantly affect the conclusion that the risk of early fatalities from spent fuel pool 

accidents is very low. 

In summary, although a detailed analysis of the factors leading to the result of no early 

fatalities could be instructive, the staff believes that the results reported are reasonably 

robust and are adequate to support the conclusions of the report.

56) Comment: 

“The portable equipment for mitigation is assumed to be available and able to be aligned to 

the SFP despite the potential priority use on the reactor.” 
E-23

NRC Response:

The study analyzed scenarios with and without successful deployment of mitigation, and 

thus it reasonably characterizes the range of possibilities.

57) Comment: 

“The work area to align the equipment is assumed to be accessible despite the draining of 

the SFP, potential high radiation, and potential loss of building structural integrity.” 

NRC Response:

The study analyzed scenarios with and without successful deployment of mitigation that

reasonably characterizes the range of possibilities.

58) Comment:

“The applicability of the HRA approach to the conditions of interest in the main report and 

appendices may merit some additional justification.” 

NRC Response:

The combination of SPAR-H and expert judgment approach to the HRA is appropriate for 

the study. Further refinement to the analysis would require performing a probabilistic risk 

assessment which is outside the scope of this analysis. 

59) Comment:

“The main report and the associated detailed analysis are a very useful compendium of 

methods and associated results for the chosen scenario. However, extrapolating the 

analysis of this single scenario to the risk conclusions drawn in Appendix D introduces large 

uncertainties and additional assumptions. Explaining and quantifying these uncertainties 

should be considered.” 

NRC Response:

The objective of this report is to study the consequences of a beyond design basis 

earthquake on a representative spent fuel pool in two configurations, low and high density at 

the reference plant. This analysis uses information contained within the study, to evaluate 

whether the reference plant in the study will benefit from the change from high- to lowdensity storage configurations in the spent fuel pool.

This analysis calculates the potential benefit per reactor year resulting from expedited fuel 

transfer by comparing the safety of high-density fuel pool storage relative to low density fuel 

pool storage. The comparison uses the initiating frequency and consequences from the 

Spent Fuel Pool Study to indicate whether any changes in the NRC’s understanding of safe 

storage of spent fuel are warranted. Given the uncertainties involved in performing this 

plant-specific regulatory analysis, low and high estimates were used to supplement the best 

estimate. For areas in which there was large uncertainties, a bounding approach was used. 

The staff also used calculated results from previous spent fuel pool studies (i.e., NUREG 
E-24

1353 and NUREG-1738) to extend the applicability of this evaluation to include other 

initiators, which could challenge spent fuel pool cooling or integrity.

Appendix D discusses the methodology, inputs, and assumptions used.

60) Comment:

“With respect to insight #3 on p. 246, the “informed expectation” has not been compared 

against an accepted seismic hazard curve. While the analysis still provides insights, a more 

traceable seismic hazard would improve the study.” 

NRC Response:

The seismic hazard used in the study was derived using the USGS 2008 seismic hazard 

model. The hazard information is provided in Chapter 3 of the report, specifically in Figure 2 

and Table 4. The seismic hazard curve provides hazard information for peak ground 

accelerations up to 1.0 g while Table 4 also provides information on the expected hazard for 

all events with peak ground accelerations greater than 1.0 g. The information in Figure 2 

shows that the seismic hazard curve used in the study predicts higher seismic hazards for 

events with peak ground accelerations greater than about 0.4 g than the hazard curves used 

in past NRC studies. These higher hazards informed the selection of the seismic event for 

the study.

Additional seismic hazard information is expected to be developed by licensees and 

submitted to the NRC for review as part of licensees’ responses to the NRC information 

request for resolution of the Fukushima Near-Term Task Force Recommendation 2.1: 

Seismic. When completed, reviewed and published that new hazard in conjunction with the 

results of the Spent Fuel Pool Study, is expected to provide further insights on seismic 

hazard issues.

61) Comment: 

“With respect to insight #5 on p. 246, the case is made that the hottest fuel should be stored 

in 1x8 patterns to minimize the time air cooling would not be effective. This insight was not 

the focus of the analysis and could benefit from further review or additional discussion to 

indicate the expected risk impact.” 

NRC Response:

The 1x8 sensitivity calculations were performed based on the reference plant’s practice of 

fuel loading in this configuration. Conclusion 5 summarizes the advantages associated with 

dispersed fuel loading patterns (both 1x4 and 1x8). As the commenter points out, detailed 

analysis of the risk impact of the 1x8 loading pattern was not the focus of the study and is 

beyond its scope.
E-25

62) Comment: 

“With respect to insight #17 on p. 249, the statement is made that “The risk due to beyond 

design basis accidents in the spent fuel pool analyzed in this study is sufficiently low that the 

added costs involved with expediting the movement of spent fuel from the pool to achieve 

the low-density fuel pool storage alternative are not warranted.” This insight is not well 

supported by the main report because a comprehensive risk assessment has not been 

performed in the main report. Referencing Appendix D would be needed along with all of the 

assumptions and inputs used in Appendix D.” 

NRC Response:

As the comment indicates, Conclusion #17 is supported by the regulatory analysis and 

backfitting discussion in Appendix D of the report. Conclusion #17 is subject to the analysis 

model, data, inputs and assumptions in Section D.3 of Appendix D . This clarification has 

been made in Conclusion #17 the report.

63) Comment:

“To enhance the existing study some additional sensitivity cases and additional 

investigations may be useful, including:

i. Sensitivity of hydrogen combustion (MELCOR)

ii. Sensitivity of 1x fuel assembly pattern (MELCOR)

iii. Sensitivity to a contiguous(uniform) fuel pattern during an outage 

(MELCOR/MACCS2)

iv. Sensitivity to concurrent reactor accident (MELCOR) (loss of Reactor Building)

v. Sensitivity to SFP leakage location and magnitude (MELCOR/MACCS2)

vi. Sensitivity to molten core-concrete interaction (MELCOR/MACCS2)

vii. Sensitivity to radiative heat transfer (MELCOR) (low density case)

viii. Sensitivity to land contamination (MACCS2)

ix. Sensitivity to time truncation (MELCOR/MACCS2)

x. Sensitivity to Reactor Building leakage (MELCOR)

xi. Sensitivity to 50-mile radius assumption (MACCS2)

xii. Sensitivity to scenario duration (>72 hrs)

It seems appropriate that the sensitivities should also address whether these could affect 

the cost-benefit assessment (e.g., would they affect both configurations in a manner that the 

differences in consequences apply equally?).” 

NRC Response:

These sensitivity calculations have been performed and documented in the report. The 

purpose was to show the range of responses that would be expected, and in some cases, 

the results were bounded by the results for the base case. 
E-26

From: State of New York, Office of the Attorney General (ADAMS Accession No. 

ML13217A134):

64) Comment: 

“As an initial matter, the State reiterates its request for more time to provide comments on 

this long-term Draft Study” 

NRC Response:

Because this is a research study we did not extend the comment period. This research 

study does not authorize any licensee action or set regulatory requirements. This study also 

does not establish any Commission policy. The comment period was appropriate for a 

research study and provided sufficient opportunity to receive comments.. 

65) Comment:

“The State is concerned with the study's failure to perform a benchmarking or bounding 

analysis.”

NRC Response:

The study is a best-estimate analysis that did not intentionally bias the results in a 

conservative or non-conservative direction. This study is intended to provide an estimate of 

the consequences from a postulated SFP accident in part to inform a regulatory analysis 

comparing the costs and benefits of expediting the transfer of fuel from the pool to dry cask 

storage. A bounding consequence analysis would have been less useful for this purpose.

66) Comment: 

“The Draft Study relies upon assumptions that underestimate the likelihood of a spent fuel 

pool accident. For example, the Draft Study assumes that liner damage is the only way to 

cause a radiological release. Thus, the Draft Study fails to address other ways cooling water 

could be lost, such as water boil-off. Instead of examining partial boil-off and partial draindown, the study focuses solely on rapid drain-down.”

NRC Response:

The study did not analyze a partial draindown, which would correspond to a leak on the 

walls of the SFP, because the staff’s structural analysis for the reference pool studied 

showed that the walls are not potential leak locations. As discussed in Section 4 of the 

report the structural analysis predicted the earthquake would cause relative motion between 

the pool walls and the pool floor that would have the potential to result in cracking in the 

reinforced concrete structure and a liner tear at the bottom of the walls. The staff found the 

pool structure to be strong enough to prevent liner tearing that could lead to leakage at any 

other location in the pool. Other components, including the refueling cavity gates and the 

piping attached to the SFP, whose failure could have increased the rate of coolant loss,

were evaluated and found to be sufficiently strong and flexible enough to resist the ground 

motion without leakage.

 
E-27

The study investigated loss of water from boil-off, which did not lead to any releases in 72 

hours after the earthquake. For the complete drain down, the study analyzed two cases, the 

moderate leak case, which leads to drain down in a few hours, and the small leak case, 

which leads to drain down in about 2 days. Based on the structural analysis performed, 

partial drain downs are not considered credible mechanisms for the reference SFP used in 

this study under severe earthquake conditions. 

67) Comment:

“the Draft Study fails to address the impact of a prolonged loss of power. As the events at 

the Fukushima facilities demonstrated, natural disasters often cause prolonged power loss 

and equipment failures, and such events at multi-unit sites can have synergistic and 

cascading consequences; therefore, it is important for NRC to conduct an in-depth study of 

the consequences of such occurrences.”

NRC Response:

The study analyzed scenarios with and without successful deployment of mitigation. All the 

scenarios without successful deployment of mitigation assumed a prolonged loss of power.

68) Comment:

“Additionally, the offsite consequence analysis in section 7 of the Draft Study is flawed.

In particular, the Draft Study fails to use realistic input values for its MACCS2 analysis. For 

example, the Draft Study unreasonably relies upon "Sample Problem A" generic values 

developed decades ago for the Surry site in rural Virginia. Instead, Staff, in drafting the 

study, should have developed site-specific MACCS2 input values. The Draft Study 

underestimates land contamination, land interdiction, and displaced individuals.”

NRC Response:

This study used input values specific for the analyzed reference plant.. The input 

parameters for the consequence analyses are based on those developed for Peach Bottom 

for the recently completed “State-of-the-Art Reactor Consequence Analyses” research 

project (NUREG-1935), as discussed in Chapter 7. Examples of site specific input 

parameters used include site specific population distributions, meteorology, and evacuation 

timing parameters including the effects of seismic events on local road networks.

69) Comment:

“The State recommends that the NRC clarify the differences between Peach Bottom and 

other U.S. nuclear reactors, and that it make clear that the Draft Study is limited to the 

reference plant and not necessarily indicative of conditions at other plants.”

NRC Response:

Section 1.3 of the SFPS report is devoted to explaining the use of the reference plant. In 

addition, the report makes clear in the Executive Summary and Conclusions that “The 

regulatory analysis for this study indicates that expediting movement of spent fuel from the 

pool does not provide a substantial safety enhancement for the reference plant. The NRC 

plans to use the insights from this analysis to inform a broader regulatory analysis of the 
E-28

spent fuel pools at all US operating nuclear reactors as part of its Japan Lessons-learned 

Tier 3 plan.”

This study is intentionally based on plant-specific information for a particular site, as 

opposed to attempting to define a generic site that might bound a set of plants. This 

approach was taken because it provides the best context for examining SFP accident 

progression and release phenomenology in a realistic fashion, for the purpose of providing a 

better understanding of the factors that affect the characterization of SFP beyond-design 

basis accidents.

70) Comment: 

“Finally, the State submits that the Draft Study's conclusion-that "expediting movement

of spent fuel from the pool does not provide a substantial safety enhancement for the 

reference plant" is not adequately supported or explained in the Draft Study.”

“The Draft Study does not give proper weight to the substantial benefits associated with low 

density spent fuel pool storage.”

NRC Response:

Section D.5 of Appendix D, "Decision Rationale," explains that the low-density spent fuel 

storage alternative evaluated does not achieve a cost-beneficial increase in public health 

and safety for the reference plant using the current regulatory framework when all event 

initiators, which may challenge spent fuel cooling or pool integrity, are considered. 

Moreover, the three sensitivity studies provided in Section D.4.1.4 also showed that the lowdensity spent fuel storage alternative was not cost-justified for any of the discounted 

sensitivity cases.

The NRC also conducted a backfit analysis for the reference plant relative to the backfit 

requirements in 10 CFR 50.109. The NRC does not conclude that this alternative results in 

a cost-justified substantial safety enhancement for the reference plant. First, the risk of a 

spent fuel pool accident at the reference plant appears to meet the Safety Goal Policy 

Statement public health objectives. The estimated spent fuel pool accident release 

frequency also is less than the 1x10-5 per reactor-year Large Early Release Frequency

guideline. Second, the cost-justified criteria are not met when evaluating the averted 

accident consequences within 50 miles of the site consistent with the regulatory framework. 

Sensitivity analyses that extend the analyses beyond 50 miles also show that the lowdensity spent fuel storage alternative was not cost-justified for any of the discounted 

sensitivity cases. Therefore, the Regulatory Baseline is justified for the alternative described 

in Section D.2.2 as evaluated for the reference plant.

In light of the findings above, the NRC concludes that the quantified safety benefits of the 

potential regulatory action provisions that qualify as backfits, considered in the aggregate, 

would not constitute a substantial increase in protection to public health or safety or the 

common defense and security. In addition the costs of this potential action would not be 

justified in view of the increase in protection to safety and security provided by the backfits 

embodied in the potential regulatory action.
E-29

From: Nuclear Energy Institute (ADAMS Accession No. ML13217A129):

71) Comment:

August 1, 2013 letter from Steven Kraft on behalf of NEI.

NRC Response:

This letter reflects on the study from NEI’s perspective, but does not contain any request for 

changes or clarifications. Thus, no response is provided here.

From: State of Vermont, Department of Public Service (ADAMS Accession No. ML13217A131):

72) Comment:

“...the comment period be extended to allow for an adequate review of a complicated study.” 

NRC Response:

Because this is a research study we did not extend the comment period. This research 

study does not authorize any licensee action or set regulatory requirements. This study also 

does not establish any Commission policy. The comment period was appropriate for a 

research study and provided sufficient opportunity to receive comments.. 

73) Comment:

“First, the study finds that there is an increased risk of a radiological release from a spent 

fuel pool within the first few months after fuel is moved from the reactor into the spent fuel 

pool. The Department thus requests that the NRC research all measures that can be taken 

by nuclear power plants to reduce the risks of a radiological release during refueling when 

"hot" fuel is present in the spent fuel pool. The goal of the NRC should be to prevent any 

radiological release.”

“Second, the study explicitly found that there are beneficial effects in the form of a reduction 

in offsite consequences, such as land contamination when a nuclear power plant moves 

from a high-density spent fuel pool loading scenario to a low-density scenario. Because 

measures can be taken to reduce the potential of land contamination which would also 

result in a reduction of radiation dose to the public and subsequent health effects, the NRC 

should consider immediately ordering reductions of spent fuel assemblies stored in a pool.”

NRC Response:

NRC’s regulatory framework is based on providing adequate protection of public health and 

safety. Land contamination is considered in the consequences and cost-benefit analysis. 

The results of this study and previous studies show that high density SFPs protect public 

health and safety in that the risk of an accident is low, and in the unlikely event of a SFP 

accident, no early fatalities are predicted and latent cancer fatality risk is low. The 

regulatory analysis in Appendix D of the report indicates that expediting movement of spent 

fuel from the pool does not provide a substantial safety enhancement for the reference 

plant. However, the NRC plans to use the insights from this analysis to inform a broader 
E-30

regulatory analysis of the spent fuel pools at all US operating nuclear reactors as part of its 

Japan Lessons-learned Tier 3 plan. In addition, Orders EA-12-051 and EA-12-049 currently 

being implemented by all operating U.S. nuclear power plants should serve to further reduce 

spent fuel pool accident risk by increasing the capability of nuclear power plants to mitigate 

beyond-design-basis external events. 

From: Mark Kelly (ADAMS Accession No. ML13219B156):

74) Comment:

“The NRC draft study depends on information that was generated using MELCOR in 

inappropriate ways. Even with input of reliable data, MELCOR produces accurate 

projections for systems only under circumstances and conditions that are understood and 

modeled well by the program. The NRC descriptions of its use of MELCOR indicate that the 

NRC attempted to use MELCOR to predict outcomes under circumstances and conditions 

that are beyond its current capabilities. MELCOR may be the best currently-available 

program for this type of study, but MELCOR capabilities don’t appear to be up to 

performance standards required for this type of NRC study of accidents.”

“NRC relies on technical information from the nuclear industry, which sometimes fails to 

meet scientific standards necessary for studies to reliably inform decision makers. The NRC 

relied heavily upon industry data for the study. The NRC has failed to require the nuclear 

industry to maintain technical information on nuclear fuel rod cladding alloys and other 

materials.”

“The draft NRC report relies excessively on NUREGs and other information sources which

themselves rely on information that may not have been adequately maintained.”

NRC Response:

MELCOR is the NRC's best estimate tool for severe accident analysis and has been 

validated against experimental data. Section 6.1.1 of the report details the modeling 

approach used for this analysis. The study relied on specific SFP models that have been 

integrated into MELCOR over the past 10 years. These models are supported by 

experimental data (e.g. new air oxidation kinetics as documented in NUREG/CR-6846). The 

comment states that NRC has relied only on technical information from the nuclear industry. 

This statement in not correct. For example MELCOR was validated against NRC-sponsored 

zirconium fire experiments conducted at Sandia National Laboratory (see NUREG/CR7143). Please see the response to Comment #83 for additional discussion on the use of 

MELCOR in the study. 

Plant specific design information for the study was provided by the licensee. In addition, 

NRC staff visited the reference plant to gather and verify plant information including 

interviews with plant staff regarding plant equipment and procedures.

NUREGs are NRC or NRC contractor reports that have been reviewed and approved for 

publication. Only credible information sources were used in the study.
E-31

From: Harmon, Curran, Spielberg + Eisenberg LLC (ADAMS Accession No. ML13225A397)

75) Comment: 

“Prior to publication of the Draft Consequence Study, NRC’s technical credibility on the poolfire issue was low. Over a period exceeding three decades, NRC had published bad 

analysis and hidden other analysis behind a veil of secrecy. Moreover, NRC failed to 

conduct sophisticated modeling and supporting experiments that could have resolved 

technical issues central to pool-fire risk, despite having an appropriate capability prior to 

1990.”

NRC Response:

This study provides updated, publicly available consequence estimates of a representative, 

postulated spent fuel pool severe accident from a beyond-design-basis earthquake under 

high-density and low-density loading conditions. MELCOR is the NRC's best estimate tool 

for severe accident analysis, and it has been validated against experimental data. Section 

6.1.1 of the report details the modeling approach used for this analysis. The study relied on 

specific SFP models that have been integrated into MELCOR over the past 10 years. 

These models are supported by experimental data (e.g. new air oxidation kinetics as 

documented in NUREG/CR-6846). In addition, MELCOR was validated against NRCsponsored zirconium fire experiments conducted at Sandia National Laboratory (see 

NUREG/CR-7143). Please see the response to Comment #83 for additional discussion on 

the use of MELCOR in the study.

The NRC conducted the SFP security assessments in a manner consistent with their 

sensitivity. The sensitive unclassified and classified analyses were made available to the 

Advisory Committee on Reactor Safeguards, the Government Accountability Office, and the 

National Academy of Sciences. None of those entities expressed concern with the 

classification of the SFP security assessments.

76) Comment:

“NRC’s Draft Consequence Study seeks to create the appearance of being a 

comprehensive assessment of the risk of a pool fire. That image is conveyed by the 

structure of the Study, by the way the Study is described in its Foreword, Abstract, and 

Executive Summary, and by unequivocal statements that high-density spent-fuel pools 

protect public health and safety. In fact, the Study’s scope is narrow. As a result, the Study’s 

examination of pool-fire risk is incomplete, and cannot support the broad, unequivocal 

findings that the Study presents. This disjunction between the apparent and actual scope of 

the Study is misleading. Moreover, in specific instances, the Study is misleading and is 

designed to support pre-determined conclusions.”

NRC Response:

The staff made clear throughout the report, (including in its Foreword, Abstract and

Executive Summary), that the analysis and results are specific to a beyond-design-basis 

seismic event at a reference plant spent fuel pool connected to a boiling water reactor with a 

Mark I containment. For example, Chapter 2 of the report lists, in detail, the key limitations 
E-32

and assumptions of the study, and states that the study is a limited-scope consequence 

study rather than a full PRA. 

77) Comment:

“Pretence of considering low-density storage: The Study does not consider the risk 

implications of reverting to low-density, open-frame racks. Instead, NRC misuses the phrase 

“low density” in order to create a false impression of the Study’s scope. This pretence 

reflects pre-determined conclusions based on a “feeling”.”

NRC Response:

The comment does not specify how the phrase “low density” was misused, and the staff had 

no intention of creating a false impression of the study’s scope. Section 6.2 of the report 

describes in text and diagrams the staff’s modeling of high density and low density racking. 

The comment correctly states that the study did not look at open-frame racks. This was 

made clear in the report at page 23, that "to get the full benefit of low-density racking, BWR 

fuel would likely need to have the channel boxes removed." Note that, for the reference 

plant, the BWR fuel assemblies channel boxes would impede crossflow even with openframe racks.

78) Comment:

“Limited consideration of water-loss scenarios: The Study focuses its analysis exclusively on 

water-loss scenarios involving total drainage. By so doing, the Study ignores a substantial 

part of the pool-fire risk. For example, the Study makes no effort to determine how the 

presence of residual water could affect fuel ignition. Extrapolation of Study findings indicates 

that consideration of this issue would substantially increase the estimated risk.”

NRC Response:

The study considered the consequences of a very large earthquake at the reference plant. 

As discussed in Section 4 of the report the structural analysis for this event predicted the 

liner tear to be at the bottom of the pool for two leak sizes with a range of draindown times. 

This provided the boundary condition for the accident progression analysis.

Please also see response to Comment #66.
E-33

79) Comment:

“Limited consideration of initiating events: The Study considers only one type of initiating 

event – an earthquake. That narrow focus reflects a pre-determined conclusion that 

earthquake is the dominant contributor to the risk of a pool fire.”

NRC Response:

The study is focused on an earthquake because it is the event shown by past studies to 

dominate spent fuel pool risk. (See NUREG-1353 and NUREG-1738)

80) Comment:

“No consideration of attack: The Study ignores the potential for an attack on a pool and/or 

adjacent reactor to initiate a pool fire. Yet, the probability of an attack with a substantial 

likelihood of success is at least equal to the probability of the severe earthquake that the 

Study does consider. Thus, the Study significantly under-estimates pool-fire risk.”

NRC Response:

Please see NRC response to Comment #17

81) Comment:

“No analysis of risk linkages among pools and reactors: The Study identifies the potential for 

risk linkages, but does not properly analyze them. For example, the Study does not analyze 

a situation in which onsite radioactive contamination and other impacts of a reactor core 

melt would preclude mitigating actions that might prevent a pool fire. Yet, the probability of a 

core melt at an adjacent reactor is at least equal to the probability of the severe earthquake 

that the Study does consider. Thus, the Study significantly under-estimates pool-fire risk.”

NRC Response:

The possibility that a concurrent reactor event may preclude operator actions is part of the 

motivation for the "unmitigated" cases analyzed in the study. Section 9.4 presents 

sensitivity calculations to show the importance of the reactor building in the progression of 

accidents in the SFP and the source term from a concurrent reactor accident. More explicit 

treatment of multi-unit issues between SFPs and reactors is being pursued under the NRC’s 

comprehensive site Level 3 PRA project discussed in SECY-11-0089. The NRC staff does 

not agree that the study significantly under-estimates pool-fire risk.

82) Comment:

“Misleading statements regarding mitigating actions: The Study concedes that its analysis of 

the feasibility of mitigating actions is very limited. Yet, the Study makes unequivocal 

statements about this feasibility. Some of those statements are misleading, and reflect predetermined conclusions.”
E-34

NRC Response:

The NRC staff had no intention of making misleading statements regarding the feasibility of 

mitigating actions. The uncertainty with the feasibility of mitigating actions is part of the 

motivation for the “unmitigated” cases analyzed in the study. The study reports in Section 6 

both "mitigated" and "unmitigated" scenarios to show the effectiveness of mitigation. The 

assumptions regarding mitigation are discussed in Section 5.3 and Section 5.3.2 details the 

rationale for reporting the unmitigated results. The study treats situations where deployment 

of mitigation is either successful or not successful, considering only portable pumping 

equipment since the study assumes that the normal ac-powered systems are likely to be 

unavailable. As Chapter 8 explains, the Human Reliability Analysis (HRA) studies the 

human performance elements of mitigation along with binning scenarios based on ac and dc 

power assumptions. But the HRA does not include a system failure analysis for the 

equipment (such as water tanks or the fire header). These types of failures are subsumed 

by the cases without successful deployment of mitigation.

83) Comment:

“In the Study, NRC employs the MELCOR code to model phenomena related to a pool fire –

including heat transfer, cladding ignition, and fire dynamics. MELCOR findings are 

significant to NRC’s estimation of pool-fire risk. Yet, the validity of MELCOR in this context, 

and the appropriateness of NRC’s input assumptions, have not been tested through a 

process of open scientific inquiry. There are significant issues that should be addressed 

through such a process, including MELCOR’s simplified treatment of radiative heat transfer.”

NRC Response:

This response discusses both the validity of the MELCOR code for SFP analysis, and some 

of the input assumptions and results of the accident progression in the reviewer detailed 

comments (see ADAMS Accession No. ML13225A397) .

MELCOR is the most widely-used regulatory severe accident code in the world. It 

represents the current state of the art in severe accident analysis, and has been developed 

through NRC and international research performed over the past three decades. In addition 

to domestic use, the code is provided by NRC to international organizations through bilateral 

agreements under the Cooperative Severe Accident Research Program (CSARP). 

MELCOR has been benchmarked against many experiments including separate and integral 

effects tests for a wide range of phenomena. Any new application of MELCOR requires 

targeted assessment of the code. The SFP models in MELCOR have been developed over 

the past 10 years, and are supported by experimental validation. The air oxidation kinetics 

models in MELCOR for zirconium-based alloys (including Zirlo and M5) are based on the 

research sponsored by NRC and documented in NUREG/CR-6846. The MELCOR model 

does not make assumptions about the temperature at which a zirconium fire is initiated, but 

rather relies on experimental data (see Section 6.1.1) to correlate the timing of breakaway 

oxidation. The generalized thermal radiation model (between any groups of assemblies 

modeled as individual rings) was developed to have flexibility in the modeling of the spent 

fuel assemblies, so no need exists for the assumption of a “cylindrical” geometry. MELCOR 

provides a best estimate prediction of fuel heat up considering all the heat transfer 

mechanisms such as radial radiation between assemblies, convective cooling by steam and 
E-35

air, radiation and convective heat transfer to the pool liner, and heat transfer from the 

exposed fuel assemblies to an underlying water pool. 

MELCOR was used in the zirconium fire experiments (see NUREG/CR-7143) and the 

predictions showed good agreement with data for the initiation and propagation of zirconium 

fire. The publication of experimental results in NUREG/CR-7143 (including code to code 

comparisons) as well as the present study and review by the Advisory Committee on 

Reactor Safeguards supports the adequacy of MELCOR’s use for this purpose.

The main focus of the present report was on progression of severe accidents following a 

loss of coolant inventory in a SFP, resulting in high temperature oxidation of cladding (and in 

some instances large scale melting and relocation of the assemblies) and associated 

release of fission products. Uncertainties in modeling of the accident progression (e.g., 

hydrogen combustion, radiative heat transfer) have been considered as part of sensitivity 

studies showing a range of fission product release. The model limitations (e.g., lack of a 

cladding deformation model) are documented in the report. In some instances, these

limitations are not expected to change the results significantly. For example, in OCP3, the 

code calculates large scale zirconium fire and fuel degradation, and release from the fuel. 

However, in OCP4, temperatures remain below 700 C (see Figures 52-55 in the report 

reproduced as Figures VII-2 and VII-3 by the reviewer).

For the layout of the assemblies in the pool (see Figure 44 in the report reproduced as 

Figure VII-1 by the reviewer), this is a stylized assumption regarding the 1x4 loading pattern 

for both the high density and low density pool to compare the consequences. In fact, as 

stated in the report, the reference plant employs a 1x8 arrangement. This is a valid 

assumption since there is relatively less variation in the decay heat of the older assemblies 

and it was appropriate to group all the older assemblies in a few computational rings since 

the risk from the zirconium fire is dominated by the hot fuel. For the study, individual 

assembly data were provided by the licensee that were subsequently used for an inventory 

calculation using the computer code SCALE. The breaking up of the freshly discharged fuel 

(284 assemblies) into two rings was to account for the variations in decay heat of the 

individual assemblies with different burnups based on the actual assembly data as 

discussed in Section 6.1.5. The calculations in the present report were for a specific SFP 

with pre-arranged fuel loading pattern. This fact is acknowledged in the report in Table 15 

and in Section 9.3. Even though the specific time requirement for achieving the 1x4 

arrangement is not publicly available information, the sensitivity calculations presented in 

Section 9.3 show that somewhat higher releases are expected in OCP1 and OCP2 for a 

contiguous fuel pattern, and the large releases for 1x4 in OCP3 are comparable to OCP2. 

The objective of this study was not to provide a detailed analyses for spent fuel pools, but to 

provide a detailed accident progression analysis for a specific pool with additional sensitivity 

calculations to offer insights on releases from different loading patterns.
E-36

84) Comment:

“In the Study, NRC has erected an elaborate superstructure of analysis on a weak 

foundation of basic knowledge about pool-fire phenomena. This superstructure culminates 

in a regulatory analysis. As discussed in paragraph VIII-2, above, the findings emanating 

from this superstructure lack scientific credibility and are misleading. Thus, the design of the 

Study is fundamentally and irredeemably flawed.”

NRC Response:

As summarized in Chapter 1 of the report, the NRC has been studying SFP risk for many 

years, including the performance of zirconium fire experiments and MELCOR benchmarking 

(NUREG/CR-7143). The NRC team conducting the study included senior engineers in each 

of the key technical areas (seismic, structural, severe accident, consequence analysis, and 

probabilistic considerations).

The statements above that the study lacks “scientific credibility” and that its design is 

“fundamentally and irredeemably flawed: are not supported by the Advisory Committee on 

Reactor Safeguards (ACRS), which the NRC views as an independent body. In the ACRS

letter recommending that the study be published (ML13198A433), the ACRS stated in part:

• The SFPS has been performed in a thorough and systematic manner, and provides 

a state-of-the-practice assessment of the consequences of a beyond-design-basis 

seismic event on the spent fuel pool in a reference boiling water reactor containing 

either high-density or low-density fuel loading.

• The SFPS provides sound approaches, tools, and insights for a broader evaluation 

of the consequences of severe seismic events on spent fuel pools of different 

design and will be valuable in determining whether expedited transfers to dry cask 

storage systems (DCSSs) produce substantial safety benefit for U.S. BWRs and 

PWRs.

85) Comment:

“The Study addresses an issue that is significant in terms of public health and safety. This 

significance is illustrated by the Study’s finding that a pool fire could lead to long-term 

displacement from their homes of more than 4 million people. Thus, citizens deserve a much 

better analysis of pool-fire risk than the incomplete, misleading work presented in NRC’s 

Draft Consequence Study.”

NRC Response:

The possibility of a zirconium fire does not contradict a finding of low risk. Risk is the product 

of likelihood and consequence, and can only be evaluated by considering both of these 

components. Moreover, the potential that beyond design-basis events might lead to a 

radiological release does not contradict a finding of reasonable assurance of adequate 

protection. 

Five Japanese nuclear power plant sites with a combined total of 20 reactors and 20 SFPs 

were subjected to severe ground motions from two major earthquakes in the past 6 years 

(i.e., March 11, 2011 Tohoku and July 16, 2007 Niigataken Chuetsu-Oki earthquakes). The 
E-37

operators of these 20 SFPs reported no water leakage from the pools after these 

earthquakes.

86) Comment: 

NRC’s Draft Consequence Study should be scrapped.

NRC Response:

See response to Comment 84.

87) Comment: 

“In addressing the pool-fire issue, NRC should focus its initial attention exclusively on 

establishing a solid technical understanding of phenomena directly related to a potential 

pool fire. To do this, NRC would start with a clean slate and use the best available modeling 

capability backed up by experiment. This modeling and experimental work would be done 

according to scientific principles.”

NRC Response:

Starting with a “clean slate” would ignore 35 years of accrued knowledge as summarized in 

Chapter 1 of the report. Please also see the response to Comment #83.

88) Comment:

“the Study proceeds to make unequivocal statements about the feasibility of mitigation. For 

example, in addressing the potential for a boil-off scenario of water loss, the Study says that 

the probability of mitigation failure extending for 7 days is “negligible”.43 That statement is 

based on no analysis, and reflects a pre-determined conclusion. NRC ignores, for example, 

the possibility that radiation fields and other onsite impacts of a reactor core melt could 

preclude mitigation for an extended period.”

Footnote 43: Barto et al, 2013, page 175.

NRC Response:

The study analyzed scenarios with and without successful deployment of mitigation that

reasonably characterizes the range of possibilities. Please also see the NRC response to 

Comment #6




UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, DC 20555-0001

------------------

OFFICIAL BUSINESS
 

 

NUREG-2161 Consequence Study of a Beyond-Design-Basis Earthquake Affecting the 

Spent Fuel Pool for a U.S. Mark I Boiling Water Reactor