NUREG-1806, Vol. 2 

 

Technical Basis for Revision of 

the Pressurized Thermal Shock 

(PTS) Screening Limit in the 

PTS Rule (10 CFR 50.61) 

Appendices 

U.S. Nuclear Regulatory Commission 

Office of Nuclear Regulatory Research 

Washington, DC 20555-0001 
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 NUREG-1806, Vol. 2

 

 

 

ppendices 

anuscript Completed: May 2006 

repared by 

irk, M. Junge, W. Arcieri, 

cher, 

 Woods, 

ivision of Fuel, Engineering and Radiological Research 

Technical Basis for Revision of 

the Pressurized Thermal Shock 

(PTS) Screening Limit in the 

PTS Rule (10 CFR 50.61) 

A

M

Date Published: August 2007 

P

M. EricksonK

B.R. Bass, R. Beaton, D. Bessette, 

T.H.J. Chang, T. Dickson, C.D. Flet

A. Kolaczkowski, S. Malik, T. Mintz, 

C. Pugh, F. Simonen, N. Siu, 

D. Whitehead, P. Williams, R.

S. Yin 

D

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 
ii
iii

Abstract 

During plant operation, the walls of reactor pressure vessels (RPVs) are exposed to neutron radiation, 

resulting in localized embrittlement of the vessel steel and weld materials in the core area. If an 

embrittled RPV had a flaw of critical size and certain severe system transients were to occur, the flaw 

could very rapidly propagate through the vessel, resulting in a through-wall crack and challenging the 

integrity of the RPV. The severe transients of concern, known as pressurized thermal shock (PTS), are 

characterized by a rapid cooling of the internal RPV surface in combination with repressurization of the 

RPV. Advancements in our understanding and knowledge of materials behavior, our ability to 

realistically model plant systems and operational characteristics, and our ability to better evaluate PTS 

transients to estimate loads on vessel walls led the NRC to realize that the earlier analysis, conducted in 

the course of developing the PTS Rule in the 1980s, contained significant conservatisms. 

This report summarizes 21 supporting documents that describe the procedures used and results obtained 

in the probabilistic risk assessment, thermal hydraulic, and probabilistic fracture mechanics studies 

conducted in support of this investigation. Recommendations on toughness-based screening criteria for 

PTS are provided. 
iv
v

Foreword 

The reactor pressure vessel is exposed to neutron radiation during normal operation. Over time, the 

vessel steel becomes progressively more brittle in the region adjacent to the core. If a vessel had a 

preexisting flaw of critical size and certain severe system transients occurred, this flaw could propagate 

rapidly through the vessel, resulting in a through-wall crack. The severe transients of concern, known as 

pressurized thermal shock (PTS), are characterized by rapid cooling (i.e., thermal shock) of the internal 

reactor pressure vessel surface that may be combined with repressurization. The simultaneous occurrence 

of critical-size flaws, embrittled vessel, and a severe PTS transient is a very low probability event. The 

current study shows that U.S. pressurized-water reactors do not approach the levels of embrittlement to 

make them susceptible to PTS failure, even during extended operation well beyond the original 40-year 

design life. 

Advancements in our understanding and knowledge of materials behavior, our ability to realistically 

model plant systems and operational characteristics, and our ability to better evaluate PTS transients to 

estimate loads on vessel walls have shown that earlier analyses, performed some 20 years ago as part of 

the development of the PTS rule, were overly conservative, based on the tools available at the time. 

Consistent with the NRC’s Strategic Plan to use best-estimate analyses combined with uncertainty 

assessments to resolve safety-related issues, the NRC’s Office of Nuclear Regulatory Research undertook 

a project in 1999 to develop a technical basis to support a risk-informed revision of the existing PTS Rule, 

set forth in Title 10, Section 50.61, of the Code of Federal Regulations (10 CFR 50.61). 

Two central features of the current research approach were a focus on the use of realistic input values and 

models and an explicit treatment of uncertainties (using currently available uncertainty analysis tools and 

techniques). This approach improved significantly upon that employed in the past to establish the 

existing 10 CFR 50.61 embrittlement limits. The previous approach included unquantified conservatisms 

in many aspects of the analysis, and uncertainties were treated implicitly by incorporating them into the 

models. 

This report summarizes a series of 21 reports that provide the technical basis that the staff will consider in 

a potential revision of 10 CFR 50.61; it includes a description of analysis procedures and a detailed 

discussion of findings. The risk from PTS was determined from the integrated results of the Fifth Version 

of the Reactor Excursion Leak Analysis Program (RELAP5) thermal-hydraulic analyses, fracture 

mechanics analyses, and probabilistic risk assessment. These calculations demonstrate that, even through 

the period of license extension, the likelihood of vessel failure attributable to PTS is extremely low 

(≈10-8/year) for all domestic pressurized water reactors. Limited analyses are continuing to further 

evaluate this finding. Should the ≈10-8/year value be confirmed, this would provide a basis for significant 

relaxation, or perhaps elimination, of the embrittlement limit established in 10 CFR 50.61. Such changes 

would reduce unnecessary conservatism without affecting safety because the operating reactor fleet has little 

probability of exceeding the limits on the frequency of reactor vessel failure established from NRC guidelines 

on core damage frequency and large early release frequency through the period of license extension. 

 _______________________________ 

 Brian W. Sheron, Director 

 Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 
vi
vii

Contents 

Abstract........................................................................................................................................................iii 

Foreword.......................................................................................................................................................v 

Contents ......................................................................................................................................................vii 

Figures .........................................................................................................................................................xi 

Tables........................................................................................................................................................xvii 

Executive Summary ...................................................................................................................................xix 

Abbreviations...........................................................................................................................................xxix 

Nomenclature.........................................................................................................................................xxxiii 

Glossary .................................................................................................................................................xxxvi 

Appendices...................................................................................................................................................xi 

1 Motivation for and Objective of this Study.......................................................................................1-1 

1.1 Description of Pressurized Thermal Shock ..............................................................................1-1 

1.2 PTS Limits on the Licensable Life of a Commercial Pressurized Water Reactor....................1-1 

1.3 Technical Factors Suggesting Conservatism of the Current PTS Rule ....................................1-3 

1.4 Statement of Objective .............................................................................................................1-3 

1.5 Guide to this Report..................................................................................................................1-3 

2 Pressurized Thermal Shock Background ..........................................................................................2-1 

2.1 General Description of the Progression of a PTS Event...........................................................2-1 

2.1.1 Precursors.........................................................................................................................2-1 

2.1.2 Thermal-Hydraulic Response of the Vessel.....................................................................2-1 

2.1.3 Response of the Vessel to PTS Loading..........................................................................2-4 

2.2 Historical Incidence of PTS......................................................................................................2-5 

2.3 Summary of SECY-82-465 Findings .......................................................................................2-6 

2.4 Current Provisions of 10 CFR 50.61 ........................................................................................2-7 

2.4.1 Index Temperature Approach to Characterize Transition Fracture Toughness in Ferritic 

Steels................................................................................................................................2-7 

2.4.2 Irradiation Effects on Index Temperature........................................................................2-8 

2.4.3 Provisions of the Current Rule.........................................................................................2-8 

2.4.4 Evaluation of Operating Plants Relative to the Current PTS Screening Limits ..............2-9 

3 PTS Reevaluation Project..................................................................................................................3-1 

3.1 Model Used to Evaluate a Revised PTS Screening Limit ........................................................3-1 

3.1.1 Restrictions on the Model ................................................................................................3-1 

3.1.2 Overall Structure of the Model ........................................................................................3-1

3.2 Uncertainty Treatment..............................................................................................................3-5 

3.2.1 Recommended Framework ..............................................................................................3-5 

3.2.2 Implementation ................................................................................................................3-6 

3.3 Fundamental Assumptions and Idealizations .........................................................................3-12 

3.3.1 Probabilistic Risk Assessment .......................................................................................3-12 

3.3.2 Thermal-Hydraulics .......................................................................................................3-14 

3.3.3 Probabilistic Fracture Mechanics...................................................................................3-16 

3.4 Participating Organizations ....................................................................................................3-20 

3.5 External Review Panel............................................................................................................3-20 

4 Structure of this Report, and Changes Relative to Previous Reports ................................................4-1 
viii

4.1 Report Structure........................................................................................................................4-1 

4.2 Changes Relative to Previous Studies ......................................................................................4-1 

4.2.1 Studies Providing the Technical Basis of the Current PTS Rule.....................................4-1 

4.2.2 December 2002 Draft Report...........................................................................................4-4 

5 Probabilistic Risk Assessment and Human Reliability Analysis ......................................................5-1 

5.1 Introduction ..............................................................................................................................5-1 

5.2 Methodology.............................................................................................................................5-2 

5.2.1 Step 1: Collect Information..............................................................................................5-3 

5.2.2 Step 2: Identify the Scope and Features of the PRA Model ............................................5-5 

5.2.3 Step 3: Construct the PTS-PRA Models..........................................................................5-8 

5.2.4 Step 4: Quantify and Bin the PTS-PRA Modeled Sequences........................................5-12 

5.2.5 Step 5: Revise PTS-PRA Models and Quantification....................................................5-13 

5.2.6 Step 6: Perform Uncertainty Analyses...........................................................................5-14 

5.2.7 Step 7: Incorporate Uncertainty and Finalize Results....................................................5-15 

6 Thermal-Hydraulic Analysis.............................................................................................................6-1 

6.1 Introduction and Chapter Structure ..........................................................................................6-1 

6.2 Thermal-Hydraulic Analysis of PTS Transients.......................................................................6-1 

6.3 RELAP5 Code Description ......................................................................................................6-2 

6.3.1 RELAP5 Analysis Process...............................................................................................6-2

6.3.2 RELAP5 Numerics Issues ...............................................................................................6-7 

6.4 Plant Model Development........................................................................................................6-8 

6.5 Transient Event Simulations...................................................................................................6-13 

6.5.1 Loss of Coolant Accidents.............................................................................................6-13 

6.5.2 Reactor/Turbine Trips....................................................................................................6-13 

6.5.3 Main Steam Line Break .................................................................................................6-14 

6.5.4 Operator Actions............................................................................................................6-14 

6.6 RELAP5 Analysis Results......................................................................................................6-15 

6.7 RELAP5 Assessment Against Experimental Data .................................................................6-15 

6.7.1 Separate Effects Tests....................................................................................................6-16 

6.7.2 Integral System Response..............................................................................................6-19 

6.7.3 RELAP5 Assessment Conclusions ................................................................................6-41 

6.8 Sensitivity and Uncertainty Analysis .....................................................................................6-41 

6.8.1 Sensitivity Analysis .......................................................................................................6-41 

6.8.2 Treatment of Uncertainties ............................................................................................6-42 

7 Probabilistic Fracture Mechanics Analysis .......................................................................................7-1 

7.1 Interaction of PFM Model with PRA and TH models..............................................................7-1 

7.2 Components of the PFM Model ...............................................................................................7-1 

7.3 Objectives of this Chapter ........................................................................................................7-2 

7.4 Approach to Model Development and Uncertainty Characterization ......................................7-2 

7.5 Flaw Model...............................................................................................................................7-4 

7.5.1 Buried Flaws in Welds.....................................................................................................7-5 

7.5.2 Buried Flaws in Plates .....................................................................................................7-6 

7.5.3 Surface Flaws in Welds and Plates ..................................................................................7-6 

7.5.4 Comparison of the Current Flaw Distribution with that Proposed by the Marshall 

Committee........................................................................................................................7-7 

7.6 Neutronics Model .....................................................................................................................7-7 

7.6.1 ID Fluence........................................................................................................................7-8 

7.6.2 Through-Wall Fluence Attenuation .................................................................................7-8 

7.7 Crack Initiation Model..............................................................................................................7-8 

7.7.1 Key Features ....................................................................................................................7-8 

7.7.2 Major Changes...............................................................................................................7-10 
ix

7.8 Through-Wall Cracking Model ..............................................................................................7-13 

7.8.1 Key Features ..................................................................................................................7-13 

7.8.2 Major Changes...............................................................................................................7-14 

7.9 Probabilistic Fracture Mechanics Code FAVOR ...................................................................7-16 

7.9.1 Implementation of PFM Model .....................................................................................7-16 

7.9.2 Discretization of the Reactor Pressure Vessel ...............................................................7-18 

7.10 Experimental Validation of Linear Elastic Fracture Mechanics.............................................7-18 

8 Vessel Failure Frequencies Estimated for Oconee Unit 1, Beaver Valley Unit 1, and Palisades .....8-1 

8.1 Chapter Structure......................................................................................................................8-1 

8.2 Plant-Specific Features of Analysis..........................................................................................8-1 

8.2.1 PRA..................................................................................................................................8-1 

8.2.2 TH ....................................................................................................................................8-2 

8.2.3 PFM .................................................................................................................................8-3 

8.3 Estimated Values of FCI and TWCF........................................................................................8-4

8.3.1 Overall Results.................................................................................................................8-5 

8.3.2 Distribution Characteristics .............................................................................................8-5 

8.4 Material Factors Contributing to FCI and TWCF ..................................................................8-17 

8.4.1 Flaws Simulated by FAVOR, and Reference Temperature Metrics..............................8-17 

8.4.2 Effect of RPV Beltline Region on FCI and TWCF Values ...........................................8-24 

8.4.3 Embrittlement Normalization between Different Plants................................................8-26 

8.4.4 Changes in these Results Relative to those Reported in December 2002 .........................8-26 

8.5 Contributions of Different Transients to the Through-Wall Cracking Frequency .................8-26 

8.5.1 Overview........................................................................................................................8-26 

8.5.2 Primary Side Pipe Breaks ..............................................................................................8-35 

8.5.3 Stuck-Open Valves on the Primary Side (SO-1) ...........................................................8-44 

8.5.4 Large Diameter Secondary Side or Main Steam Line Breaks .......................................8-59 

8.5.5 Stuck-Open Valves on the Secondary Side (SO-2) .......................................................8-66 

8.5.6 Other Transient Classes .................................................................................................8-71 

8.6 Summary.................................................................................................................................8-71 

9 Generalization of the Baseline Results to All Pressurized-Water Reactors ......................................9-1 

9.1 Thermal-Hydraulic Sensitivity Studies ....................................................................................9-1 

9.1.1 Introduction......................................................................................................................9-1 

9.1.2 Sensitivity Studies Performed for Uncertainty Analysis .................................................9-2 

9.2 Fracture mechanics sensitivity studies .....................................................................................9-7 

9.2.1 Sensitivity Studies Performed To Assess the Robustness of the PFM Model.................9-7 

9.2.2 Sensitivity Studies Performed to Assess the Applicability of the Results in Chapter 8 to 

PWRs in General ...........................................................................................................9-13 

9.2.3 Summary and Conclusions ............................................................................................9-17 

9.3 Plant-to-Plant Differences in Design/Operational Characteristics that Impact PTS Transient 

Severity .......................................................................................................................................9-18 

9.3.1 Generalization Questionnaire.........................................................................................9-19

9.3.2 Analysis of Collected Information.................................................................................9-19 

9.3.3 Combined Observations and Overall Conclusion..........................................................9-33 

9.4 Consideration of External Events...............................................................................................9-34 

9.4.1 Introduction....................................................................................................................9-34 

9.4.2 Approach........................................................................................................................9-35 

9.4.3 Findings Based on the Reviews .....................................................................................9-35 

9.4.4 Additional Analyses.......................................................................................................9-36 

9.4.5 Overall Findings ............................................................................................................9-38 

9.5 Summary of Generalization Studies .......................................................................................9-39 

10 Risk-Informed Reactor Vessel Failure Frequency Acceptance Criteria .........................................10-1 
x

10.1 Introduction ............................................................................................................................10-1 

10.2 Current Guidance on Risk-Informed Regulation....................................................................10-1 

10.3 Containment Performance During PTS Accidents .................................................................10-2 

10.3.1 Previous Research Results .............................................................................................10-3 

10.3.2 Post-RPV Failure Scenarios Scoping Study ..................................................................10-4 

10.4 Acceptance Criteria Options.................................................................................................10-11 

10.5 Conclusions ..........................................................................................................................10-12 

11 Reference Temperature (RT)-Based PTS Screening Criteria..........................................................11-1 

11.1 Introduction ............................................................................................................................11-1 

11.2 Reference Temperature (RT) Metrics .....................................................................................11-2 

11.3 Relationship between RT Metrics and TWCF........................................................................11-3 

11.3.1 Weighted RT Values ......................................................................................................11-3 

11.3.2 Maximum RT Values .....................................................................................................11-6 

11.4 Proposed RT-Based Screening Limits ....................................................................................11-7 

11.4.1 Plate Vessels ..................................................................................................................11-7 

11.4.2 Forged Vessels.............................................................................................................11-11 

11.4.3 Need for Margin...........................................................................................................11-14 

11.4.4 Non-Best Estimate Aspects of the Model....................................................................11-15 

12 Summary of Findings and Considerations for Rulemaking ............................................................12-1 

12.1 Plant-Specific Baseline Analysis of the PTS Risk at Oconee Unit 1, Beaver Valley 

Unit 1, and Palisades...................................................................................................................12-1 

12.2 Applicability of these Plant Specific Results to Estimating the PTS Risk at PWRs in General.12-5 

12.3 An Anual Per-Plant Limit on Through-Wall Cracking Frequency Consistent with Current 

Regulatory Guidance on Risk-Informed Regulation ..................................................................12-7 

12.4 A Reference Temperature Based PTS Screening Criteria......................................................12-9 

12.5 Considerations for Rulemaking............................................................................................12-13 

13 References .....................................................................................................................................13-14 

13.1 Citations Summarized by this Report...................................................................................13-14 

13.1.1 Probabilistic Risk Assessment .....................................................................................13-14 

13.1.2 Thermal-Hydraulics .....................................................................................................13-14 

13.1.3 Probabilistic Fracture Mechanics.................................................................................13-15 

13.2 Literature Citations...............................................................................................................13-16 
xi

Appendices 

Appendix A Master Transient List and FAVOR 04.1 Results Summary.................................A-1 

Appendix B Peer Review .......................................................................................B-1 

Appendix C Flaw Distribution, Correspondence with Dr. Fredric Simonen of the Pacific Northwest 

National Laboratory...............................................................................C-1 

Appendix D Comparison of Plant-Specific Reference Temperature Values..............................D-1 

Appendix E Detailed Reply to Peer Review Comment 22..................................................E-1 

Appendix F Detailed Reply to Peer Review Comment 75..................................................F-1 

Appendix G Flaw Distributions for Forgings.................................................................G-1 

Figures 

Figure 1.1. Proximity of currently operating PWRs to the 10 CFR 50.61 screening limit for PTS.............1-2 

Figure 2.1. Schematic illustration of the main components of the primary (red) and secondary (blue) 

systems in a pressurized-water reactor ...................................................................................2-2 

Figure 2.2. Figure 2.13 from SECY-82-465, depicting the final temperatures for 32 PTS precursor 

events experienced in commercial reactor service prior to 1982 ...........................................2-6 

Figure 2.3. The empirical data used to establish the ASME KIc curve .....................................................2-8 

Figure 3-1. High-level schematic showing how a probabilistic estimate of through-wall cracking 

frequency (TWCF) is combined with a TWCF acceptance criterion to arrive at a proposed 

revision of the PTS screening limit. .......................................................................................3-2 

Figure 3-2. The three plants analyzed in detail in the PTS reevaluation effort.........................................3-4 

Figure 3.3. Characterization of TH uncertainties......................................................................................3-9 

Figure 3-4. Illustration of the magnitude of Kapplied values that contribute to TWCF because they have 

a conditional probability of crack initiation > 0. The top graph shows all Kapplied values with 

CPI > 0 overlaid on the KIc transition curve from an analysis of Beaver Valley Unit 1 at 60 EFPY. 

The bottom graph shows these same results expressed in the form of a cumulative distribution 

function.................................................................................................................................3-18 

Figure 3.5. Distribution of crack initiating depths generated by FAVOR Version 03.1.........................3-19 

Figure 4-1. Structure of documentation summarized by this report. When these reports are cited in the 

text, the citation appears in italicized boldface to distinguish them from literature citations.4-2 

Figure 5.1. Integrated technical analyses comprising the PTS reanalysis project ....................................5-1 

Figure 5.2. Diagrammatic representation of the PRA approach ...............................................................5-3 

Figure 5.3. Functional event tree as the basis for PTS PRA analysis .....................................................5-11 

Figure 6.1. Schematic of RELAP5 Input and Output Processing.............................................................6-5 

Figure 6.2. Schematic of RELAP5 Execution Processing........................................................................6-6

Figure 6.3. Palisades Reactor Vessel Nodalization ................................................................................6-10 

Figure 6.4. Palisades Coolant Loop Nodalization ..................................................................................6-11 

Figure 6.5. Palisades Main Steam System Nodalization ........................................................................6-12 

Figure 6.6. Mass Flow Rate at Nozzle Outlet (Marviken Test 22).........................................................6-17 

Figure 6.7. Pressure Rise (MIT Pressurizer Test ST4) ...........................................................................6-18 
xii

Figure 6.8. Reactor Vessel Downcomer Fluid Temperatures (ROSA-IV Test SB-CL-18)....................6-20 

Figure 6.9. Reactor Vessel Downcomer Temperatures (ROSA-IV Test SB-HL-06).............................6-21 

Figure 6.10. Reactor Vessel Downcomer Fluid Temperatures (ROSA/AP600 Test AP-CL-03).............6-22 

Figure 6.11. Reactor Vessel Downcomer Fluid Temperatures (ROSA/AP600 Test AP-CL-09).............6-23 

Figure 6.12. Reactor Vessel Downcomer Fluid Temperatures (Test APEX-CE-13) ...............................6-25 

Figure 6.13. Pressurizer Pressure (Test APEX-CE-13) ............................................................................6-25 

Figure 6.14. Reactor Vessel Downcomer Fluid Temperatures (Test APEX-CE-05) ...............................6-26 

Figure 6.15. Reactor Vessel Downcomer Fluid Temperatures (LOFT Test L3-7)...................................6-27 

Figure 6.16. Reactor Vessel Downcomer Fluid Temperatures (LOFT Test L2-5)...................................6-28 

Figure 6.17. Reactor Vessel Downcomer Fluid Temperatures (LOFT Test L3-1)...................................6-30 

Figure 6.18. Reactor Vessel Downcomer Fluid Temperatures (MIST Test 360499)...............................6-31 

Figure 6.19. Reactor Vessel Downcomer Fluid Temperatures (MIST Test 3109AA) .............................6-32 

Figure 6.20. Reactor Vessel Downcomer Fluid Temperatures (MIST Test 4100B2) ..............................6-32 

Figure 6.21. UPTF 1-21 DC Velocities at Bottom-Core Elevation Measured, Turbine Meters, 

Clockwise from Cold Leg 2, Filtered...................................................................................6-37 

Figure 6.22. UPTF 1-21 DC Velocities at Bottom-Core Elevation RELAP5 

 (Calculated, Clockwise from Cold Leg 2) .........................................................................6-38 

Figure 6.23. UPTF 1-21 Wall Temperature Minus Fluid Temperature Under Cold Leg 2 

at Core-Top Elevation ..........................................................................................................6-38 

Figure 6.24. UPTF Test 1-21 Wall Temperatures at 25 mm Depth Vessel Wall in Orientation of 

Cold Leg 2............................................................................................................................6-39 

Figure 6.25. APEX–CE–05 Measured and RELAP5 Fluid Temperatures 

4D below and centered on the Cold Leg 4 nozzle................................................................6-39 

Figure 6.26. APEX–CE–05 Measured and RELAP5 Wall Temperatures 

4D below and centered on the Cold Leg 4 nozzle................................................................6-40 

Figure 6.27. Creare Data Compared to Dittus-Boelter .............................................................................6-40 

Figure 6.28. Sensitivity parameter ranking of a 2.8-in. (7.18-cm) surge line break LOCA .....................6-51 

Figure 6.29. Confirmation of the Linearly Additive Assumption for a 2.8-in. (7.18-cm) surge line break LOCA

..............................................................................................................................................6-51 

Figure 6.30. Illustration of the Statistical Results for Downcomer Temperature Distribution.................6-51 

Figure 7.1. Illustration of the interrelationships between PFM model and the TH and PRA models, 

and the four principal sub-models that comprise the PFM model..........................................7-3 

Figure 7.2. Illustration of a root cause diagram showing how uncertainties in input variables (E, F, G, and 

H) propagate through models (nodes), themselves potentially having uncertainty, to produce 

uncertainty in a resultant value (A) ........................................................................................7-4

Figure 7.3. Comparison of the new flaw distribution to the Marshall flaw distribution...........................7-7 

Figure 7.4. Illustration of the impact of the flaw distribution adopted in this study (improved PNNL) with 

that used in previous PTS calculations (Marshall flaw characterization) [Dickson 02] 

(analysis performed on Oconee at 60 EFPY) .........................................................................7-7 

Figure 7.5. Number of flaws simulated that have a conditional probability of through-wall cracking that 

exceeds zero (Oconee at 60 EFPY)......................................................................................7-13 

Figure 7.6. Combined effects of allowing KIa to exceed 200 ksi√in and allowing for ductile crack reinitiation on the upper shelf. Open points show TWCF results when KIa is allowed to exceed 

200 ksi√in and ductile crack re-initiation is permitted. RTNDT* is defined in [Kirk 12-02].7-15 

Figure 7.7. Comparison of temperature separation between crack initiation and crack arrest toughness 

transition curves assumed in our current calculations (blue curve) with the constant separation 

of ~30°C (~86°F) assumed by previous calculations ...........................................................7-15 

Figure 7.8. FAVOR data streams flow through three modules: (1) FAVLoad, (2) FAVPFM, and (3) 

FAVPost ...............................................................................................................................7-16 

Figure 7.9. Flow chart for improved PFM model implemented in FAVPFM showing the four primary 

nested loops – (1) RPV Trial Loop, (2) Flaw Loop, (3) Transient Loop, and (4) Time Loop........7-17 
xiii

Figure 7.10. Rollout diagram of beltline materials and representative fluence maps for Oconee Unit 1.7-18 

Figure 7-11. Test vessels used in the ITV and PTSE test series (top) and in the TSE test series (bottom)......7-20 

Figure 8.1. Rollout diagram of beltline materials and representative fluence maps for Oconee............8-15 

Figure 8.2. Typical distribution of through-wall cracking frequency (as calculated for Beaver Valley at 

32 EFPY (blue circles) and for extended embrittlement conditions (red diamonds) ...........8-17 

Figure 8-3. TWCF distribution percentile corresponding to the mean value..........................................8-17 

Figure 8-4. Relationship between a reference temperature (RT) and various measure of resistance to 

fracture (fracture toughness). This is a schematic illustration of temperature dependence 

only; scatter in fracture toughness is not shown...................................................................8-19 

Figure 8-5. Location and orientation of flaws simulated by FAVOR to exist in different regions 

of the RPV beltline...............................................................................................................8-21 

Figure 8-6. Effect of flaw orientation on the bending experienced by a cylinder 

subjected to a cold thermal shock on the inner diameter......................................................8-21

Figure 8.7. Through-wall variation of crack driving force (KI): axially oriented flaws 

compared to circumferentially oriented flaws 

(Comparison is shown for an 8-inch diameter surge line break in Beaver Valley (Transient #7 

– see top plot) at a time 11 minutes after the start of the transient (see bottom plot).) ........8-22 

Figure 8-8. Contributions of the different transients to the TWCF in Oconee Unit 1 

(Numbers on the abscissa are the TH case numbers, see Appendix A) ...............................8-29 

Figure 8-9. Contributions of the different transients to the TWCF in Beaver Valley Unit 1 

(Numbers on the abscissa are the TH case numbers, see Appendix A) ...............................8-30 

Figure 8-10. Contributions of the different transients to the TWCF in Palisades 

(Numbers on the abscissa are the TH case numbers, see Appendix A) ...............................8-30 

Figure 8-11. Variation in percent contribution to the total TWCF of different transient classes 

with reference temperature (RT) as defined in Table 8.6. The contributions of feed-and-bleed 

LOCAs and steam generator tube ruptures were also assessed. These transient classes made no 

contribution to TWCF, with the exception that feed-and-bleed LOCAs contributed < 0.1% to 

the TWCF of the Palisades RPV. .........................................................................................8-34 

Figure 8-12. Comparison of the annual frequencies of various broad classes of events for full-power conditions

..............................................................................................................................................8-35 

Figure 8.13. Effect of surge line and hot-leg break diameter on the cooldown characteristics of Beaver 

Valley (top) and Oconee (bottom)........................................................................................8-37 

Figure 8.14. Effect of break location on the cooldown characteristics in Oconee (top) and Palisades 

(bottom)................................................................................................................................8-37 

Figure 8.15. Effect of season on the cooldown characteristics of cold leg breaks in Palisades..............8-38 

Figure 8.16. Effect of surge line and hot-leg break diameter on the depressurization characteristics of 

Beaver Valley (top) and Oconee (bottom) ...........................................................................8-38 

Figure 8.17. Effect of surge line and hot-leg break diameter on the heat transfer coefficient in Beaver 

Valley (top) and Oconee (bottom)........................................................................................8-38 

Figure 8.18. Comparison of the cooldown characteristics of the three plants modeled 

for a spectrum of break diameters ........................................................................................8-39 

Figure 8.19. Effect of pipe break diameter and break location on the conditional probability of throughwall cracking. (CPTWC taken at approximately equivalent embrittlement levels between 

plants (Beaver Valley and Palisades at 60 EFPY, Oconee at Ext-Ob)) .................................8-40 

Figure 8.20. Effect of pipe break diameter, break location, and season (S=Summer, W=Winter) on the 

conditional probability of through-wall cracking for Palisades ...........................................8-41 

Figure 8.21. Effect of pipe break diameter and break location on the conditional proportion of initiated 

flaws that propagate through the wall. (CPTWC/CPI ratios taken at approximately equivalent 

embrittlement levels between plants (Beaver Valley and Palisades at 60 EFPY, Oconee at ExtOb).) .....................................................................................................................................8-41 
xiv

Figure 8.22. Effect of LOCA break diameter and break location on the time at which through-wall 

cracking occurs (Break times taken at approximately equivalent embrittlement levels between 

plants (Beaver Valley and Palisades at 60 EFPY, Oconee at Ext-Ob).).................................8-42 

Figure 8.23. The TWCF attributable only to primary side pipe breaks in the three study plants ...........8-43 

Figure 8-24. Radial profile of arrested crack in TSE 6 [Cheverton 85a] (The crack in this experiment 

arrested after propagating 95% of the way through the cylinder wall.) ...............................8-43 

Figure 8-25. Effect of pipe break diameter on the pressure in the primary system at the most likely time 

of failure ...............................................................................................................................8-44 

Figure 8-26. Oconee SO-1 transients where the stuck-open SRVs reclose after 3,000 seconds. 

(Transients in the upper graphs initiate from full power, while transients in the lower graphs 

initiate from hot zero power.)...............................................................................................8-46 

Figure 8-27. Oconee SO-1 transients where the stuck-open SRVs reclose after 6,000 seconds (Transients 

in the upper graphs initiate from full power, while transients in the lower graphs initiate from 

hot zero power.)....................................................................................................................8-48 

Figure 8-28. Beaver Valley SO-1 transients where a single stuck-open SRV recloses after 6,000 seconds 

(Transients in the upper graphs initiate from full power, while transients in the lower graphs 

initiate from hot zero power.)...............................................................................................8-49 

Figure 8-29. Beaver Valley SO-1 transients where a two stuck-open SRVs reclose after 6,000 seconds. 

(Transients in the upper graphs initiate from full power, while transients in the lower graphs 

initiate from hot zero power.)...............................................................................................8-50 

Figure 8-30. Beaver Valley SO-1 transients where stuck-open SRVs recloses after 3,000 seconds 

(All transients initiate from hot zero power conditions. Transients in the upper graphs have one 

stuck-open valve, whereas transients in the lower graphs have two stuck-open valves.) ......8-51 

Figure 8-31. Comparison of SO-1 transients between different plants for transients initiated from 

HZP conditions, valve reclosure after 6,000 sec., and no HPI throttling .............................8-53 

Figure 8.32. The TWCF attributable solely to stuck-open valves on the primary side that later reclose8-58 

Figure 8.33. Power level effects on MSLB transients at Beaver Valley. Both breaks are in containment 

and have AUX feed continuing to the faulted generator for 30 minutes. The operator throttles 

HPSI 30 minutes after allowed.............................................................................................8-61 

Figure 8.34. Power level effects on MSLB transients at Palisades. Both breaks are in containment and 

include failures of both MSIVs to close. The operator takes no actions to either isolate AUX 

feed or to throttle HPI...........................................................................................................8-62 

Figure 8.35. Break location effects on MSLB transients at Beaver Valley. Both breaks include 

continuous AUX feed and are initiated from full-power conditions. In transient 106, the 

operator controls HPSI 30 minutes after allowed.................................................................8-62 

Figure 8.36. Break isolation effects on MSLB transients at Palisades. Both breaks occur inside 

 containment and are initiated from HZP conditions.............................................................8-63 

Figure 8.37. Effect of HHSI control on MSLB transients at Beaver Valley. In both breaks, AUX feed is 

 isolated 30 minutes after the break occurs. Both breaks are initiated from full-power 

conditions. ............................................................................................................................8-63 

Figure 8.38. Percentage of initiated cracks that propagate through-wall for MSLB transients..............8-65 

Figure 8.39. TWCF attributable to MSLB transients..................................................................................8-65 

Figure 8.40. Comparison of TWCF attributable to primary side stuck-open valves, primary side pipe 

 breaks, and MSLBs. Note that the contribution of MSLBs here overrepresents their actual 

 contributions to TWCF because of conservatisms in their modeling. On each graph, an 

upper-bound curve is hand drawn to the data originally presented in Figure 8.23, Figure 8.32, 

and Figure 8.39. On the left hand graph, all three upper-bound curves are placed together for 

easy comparison. ..................................................................................................................8-67 

Figure 8.41. The cooldown rate of various SO-2 transients, graphs (d) through (f), compared to MSLBs, 

graph (c), and primary side transients, graphs (a) and (b)....................................................8-68 
xv

Figure 8.42. Small steam line break simulated by sticking open all MSSVs in steam generator A 

with AFW continuing to feed affected generator for 30 minutes. Beaver Valley transient 111 

occurs at HZP, while Beaver Valley transient 118 occurs at full power..............................8-69 

Figure 8.43. Reactor/turbine trip with loss of MFW and EFW in Oconee. 

 Operator opens all TBVs to depressurize the secondary side. ............................................8-69 

Figure 8.44. Reactor/turbine trip with two stuck-open safety valves in Oconee. ....................................8-69 

Figure 8.45. Reactor/turbine trip with one or two stuck-open ADVs (P-019 and P-055, respectively) 

 in Palisades..........................................................................................................................8-70 

Figure 9.1. CPTWC Behavior for LOCAs of Various Break Diameters..................................................9-6 

Figure 9.2. Flaw dimension and position descriptors adopted in FAVOR ...............................................9-9 

Figure 9.3. Distribution of through-wall position of cracks that initiate...................................................9-9 

Figure 9.4. Flaw depths that contribute to crack initiation probability in Beaver Valley Unit 1 

when subjected to medium- and large-diameter pipe break transients at two different 

embrittlement levels .............................................................................................................9-10 

Figure 9.5. Flaw depths that contribute to crack initiation probability in Beaver Valley Unit 1 

when subjected to stuck-open valve transients at two different embrittlement levels .........9-10 

Figure 9.6. Comparison of embrittlement shift uncertainties simulated by FAVOR (blue line with X 

symbols) with the uncertainties in the experimental embrittlement shift database used by 

Eason to construct the model................................................................................................9-12 

Figure 9.7. Effect of different methods to artificially increase embrittlement on the predicted TWCF 

values....................................................................................................................................9-14 

Figure 9.8. Effect of vessel wall thickness on the variation of applied-KI vs. time for a 16-in. (40.64-cm) 

diameter hot leg break in Beaver Valley. The flaw has the following dimensions: L=0.35-in., 

2a=0.50-in., 2c=1.5-in. (L=8.89-mm, 2a=12.7-mm, 2c=38.1-mm).....................................9-17 

Figure 9.9. Distribution of RPV wall thicknesses for PWRs currently in service [RVID2]...................9-17 

Figure 9.10.Effect of vessel wall thickness on the TWCF of various transients in Beaver Valley (all analyses 

at 60 EFPY)..........................................................................................................................9-17 

Figure 10.1.Post-RPV failure accident progression tree..........................................................................10-6 

Figure 11-1.Correlation of through-wall cracking frequencies with weighted reference temperature 

metrics for the three study plants (°R = °F + 459.69) ..........................................................11-8 

Figure 11-2.Correlation of through-wall cracking frequencies with maximum reference temperature 

metrics for the three study plants (°R = °F + 459.69) ..........................................................11-9 

Figure 11-3. Comparison of FAVOR 04.1 TWCF estimates with TWCF values estimated using weighted 

RT values (Eq. 11-1)...........................................................................................................11-10 

Figure 11-4. Comparison of FAVOR 04.1 TWCF estimates with TWCF values estimated using maximum 

RT values (Eq. 11-2)...........................................................................................................11-10 

Figure 11-5. Weighted RT-based screening criterion for plate vessels based on Eq. 11-1 

(Left: Effect of RTCW value for a fixed TWCFTOTAL value of 1x10-6; 

Right: Effect of TWCFTOTAL for a fixed RTCW value of 300°F (149°C)).............................11-11 

Figure 11-6. Maximum RT-based screening criterion for plate vessels based on Eq. 11-2...................11-12 

Figure 11-7. Comparison of the maximum RT-based screening limit for plate vessels based on Eq. 11-2 

with assessment points for all operating PWRs at EOL (32 EFPY, 40 operating years) (left) 

and EOLE (48 EFPY, 60 operating years) (right) (Plant RT values estimated from 

information in [RVID2]. RTMAX-CW is 300°F (149°C) for both graphs, exceeding the calculated 

RTMAX-CW value for any plant at EOL or EOLE.).....................................................................11-12 

Figure 11-8. TWCF values for operating PWRs estimated using Eq. 11-1 at EOL (left) and EOLE (right) 

(Values for individual plants are reported in Appendix D.) ...............................................11-13 

Figure 11-9. Maximum RT-based screening criterion for forged vessels based on Eq. 11-1, illustrating the 

effect of TWCFTOTAL ...........................................................................................................11-14 
xvi

Figure 11-10. Comparison of maximum RT-based screening limit for forged vessels based on Eq. 11-2 

with operating PWRs at EOL (32 EFPY, 40 operating years) (left) and EOLE (48 EFPY, 60 

operating years) (right) (Plant RT values estimated from information in [RVID2]. RTMAX-PL is 

estimated based on forging properties.)..............................................................................11-14 

Figure 12-1. Comparison of RT-based screening limit (curves) with assessment points for operating 

PWRs at EOL in plate vessels (left) and forged vessels (right) .........................................12-12 
xvii

Tables 

Table 3.1. Summary of uncertainty treatment in the three major technical areas ..................3-10 

Table 3.2. Participating organizations .................................................................................... 3-21 

Table 5.1. Comparison of PRA analyses used in this study 

 with the PRA analyses that supported 10 CFR 50.61.............................................. 5-6 

Table 5.2 General classes of human failures considered in the PTS analyses.........................5-9 

Table 6.1 Phenomena Identification and Ranking Table for Pressurized Thermal Shock in 

PWRs ....................................................................................................................... 6-4 

Table 6.2 Summary of PTS Sensitivity Studies.....................................................................6-43 

Table 6.3 List of Combined Sensitivity Indicators Varied for LAA Verification .................6-49 

Table 6.4 Example of Representative Scenario Selection ..................................................... 6-51 

Table 7.1. Summary of sources of experimental data sources for the flaw distribution...........7-5 

Table 8.1. Summary of Plant Parameters Relevant to the PTS Evaluation............................... 8-7 

Table 8.2. Plant specific material values drawn from the RVID2 database [RVID2]............8-10 

Table 8.3. Summary of vessel specific inputs for the flaw distribution .................................8-12 

Table 8.4. Mean crack initiation and through-wall cracking frequencies estimated for Oconee 

 Unit 1, Beaver Valley Unit 1, using FAVOR Version 04.1 ................................. 8-16 

Table 8.5. Relative contributions of various flaw populations to the FCI and TWCF values 

 estimated by FAVOR Version 04.1....................................................................... 8-25 

Table 8.6. Reference temperature metric used in Section 8.5. ...............................................8-27 

Table 8.7. Transients that contribute most significantly to the estimated TWCF of Oconee 

 Unit 1 .....................................................................................................................8-30 

Table 8.8. Transients that contribute most significantly to the estimated TWCF of Beaver 

 Valley Unit 1...........................................................................................................8-31 

Table 8.9. Transients that contribute most significantly to the estimated TWCF of Palisades 8-32 

Table 8.10. Transient severity indicators and estimated values of CPTWC for Oconee 

 at Ext-Ob embrittlement conditions........................................................................8-54 

Table 8.11. Mean operator action probabilities in our modeling of SO-1 transients................8-57 

Table 8.12. Factors contributing to the severity and risk-dominance of various transient 

 classes ...................................................................................................................8-73

Table 9.1. Summary of Oconee Downcomer Fluid Temperature Sensitivity Results for 

 LOCA.........................................................................................................................9-3 

Table 9.2. Summary of Beaver Valley Downcomer Fluid Temperature Sensitivity Results for 

 LOCA.........................................................................................................................9-3 

Table 9.3. Summary of Palisades Downcomer Fluid Temperature Sensitivity Results for 

 LOCA.........................................................................................................................9-4 

Table 9.4. Summary of Downcomer Fluid Temperature Sensitivity Results 

 for Stuck-Open Primary Side Valves.........................................................................9-5 

Table 9.5. Plant list for generalization study ............................................................................9-21

Table 9.6 Important PTS scenarios and corresponding plant design and operational features9-22 

Table 9.7. Scenarios covered under the external event analyses ..............................................9-37 

Table 9.8. Small-break LOCA internal event results................................................................9-37 
xviii

Table 9.9. Small-break LOCA TWCF comparison ..................................................................9-39 

Table 10.1. Post-RPV-failure technical issues...........................................................................10-5 

Table 10.2. Potentially risk-significant post-RPV failure accident progression scenarios........10-8 

Table 10.3. Key APET scenarios............................................................................................. 10-11 

Table 11.1. Contributions of different flaw populations to the TWCF values estimated by FAVOR 

 Version 04.1 ............................................................................................................. 11-5 

Table 11.2. Non-best estimate aspects of the models used to develop the RT-based screening 

 limits for PTS shown in Figure 11-5 and Figure 11-9..........................................11-16 

Table 12.1. Factors contributing the severity and risk dominance of various transient 

 classes. ....................................................................................................................12-3 
xix

Executive Summary 

This report summarizes the results of a 5-year study conducted by the U.S. Nuclear Regulatory Commission 

(NRC), Office of Nuclear Regulatory Research (RES). The aim of this study was to develop the technical 

basis for revision of the Pressurized Thermal Shock (PTS) Rule, as set forth in Title 10, Section 50.61, 

of the Code of Federal Regulations (10 CFR 50.61), “Fracture Toughness Requirements for Protection 

Against Pressurized Thermal Shock Events,” consistent with the NRC’s current guidelines on risk-informed 

regulation. This report, together with other supporting reports documenting the study details and results, 

provides this basis. 

This executive summary begins with a description of PTS, how it might occur, and its potential consequences 

for the reactor pressure vessel (RPV). This is followed by a summary of the current regulatory approach 

to PTS, which leads directly to a discussion of the motivations for conducting this project. Following this 

introductory information, we describe the approach used to conduct the study, and summarize our key findings 

and recommendations, which include a proposal for revision of the PTS screening limits. We then conclude 

the executive summary with a discussion of the potential impact of this proposal on regulations other than 

10 CFR 50.61. 

Description of PTS 

During the operation of a nuclear power plant, the RPV walls are exposed to neutron radiation, resulting in 

localized embrittlement of the vessel steel and weld materials in the area of the reactor core. If an embrittled 

RPV had an existing flaw of critical size and certain severe system transients were to occur, the flaw 

could propagate very rapidly through the vessel, resulting in a through-wall crack and challenging the integrity 

of the RPV. The severe transients of concern, known as pressurized thermal shock (PTS), are characterized 

by a rapid cooling (i.e., thermal shock) of the internal RPV surface and downcomer, which may be 

followed by repressurization of the RPV. Thus, a PTS event poses a potentially significant challenge to 

the structural integrity of the RPV in a pressurized-water reactor (PWR). 

A number of abnormal events and postulated accidents have the potential to thermally shock the vessel 

(either with or without significant internal pressure). These events include a pipe break or stuck-open valve 

in the primary pressure circuit, a break of the main steam line, etc. During such events, the water level in 

the core drops as a result of the contraction produced by rapid depressurization. In events involving a break 

in the primary pressure circuit, an additional drop in water level occurs as a result of leakage from the break. 

Automatic systems and operators must provide makeup water in the primary system to prevent overheating of 

the fuel in the core. However, the makeup water is much colder than that held in the primary system. As 

a result, the temperature drop produced by rapid depressurization coupled with the near-ambient 

temperature of the makeup water produces significant thermal stresses in the thick section steel wall of the 

RPV. For embrittled RPVs, these stresses could be sufficient to initiate a running crack, which could 

propagate all the way through the vessel wall. Such through-wall cracking of the RPV could precipitate 

core damage or, in rare cases, a large early release of radioactive material to the environment. 

Fortunately, the coincident occurrence of critical-size flaws, embrittled vessel steel and weld material, 

and a severe PTS transient is a very low-probability event. In fact, only a few currently operating PWRs 

are projected to closely approach the current statutory limit on the level of embrittlement during their 

planned operational life. 
xx

Current Regulatory Approach to PTS 

As set forth in 10 CFR 50.61, the PTS Rule requires licensees to monitor the embrittlement of their RPVs 

using a reactor vessel material surveillance program qualified under Appendix H to 10 CFR Part 50, 

“Reactor Vessel Material Survellience Program Requirements.” The surveillance results are then used 

together with the formulae and tables in 10 CFR 50.61 to estimate the fracture toughness transition 

temperature (RTNDT) of the steels in the vessel’s beltline and how those transition temperatures increase 

as a result of irradiation damage throughout the operational life of the vessel. For licensing purposes, 

10 CFR 50.61 provides instructions on how to use these estimates of the effect of irradiation damage 

to estimate the value of RTNDT that will occur at end of license (EOL), a value called RTPTS. 10 CFR 50.61 

also provides “screening limits” (maximum values of RTNDT permitted during the plant’s operational life) 

of +270°F (132°C) for axial welds, plates, and forgings, and +300°F (149°C) for circumferential welds. 

These screening limits correspond to a limit of 5x10-6 events/year on the annual probability of developing 

a through-wall crack [RG 1.154]. Should RTPTS exceed these screening limits, 10 CFR 50.61 requires 

the licensee to either take actions to keep RTPTS below the screening limit (by implementing “reasonably 

practicable” flux reductions to reduce the embrittlement rate, or by deembrittling the vessel by annealing 

[RG 1.162]), or perform plant-specific analyses to demonstrate that operating the plant beyond the 10 CFR 50.61 

screening limit does not pose an undue risk to the public [RG 1.154]. 

While no currently operating PWR has an RTPTS value that exceeds the 10 CFR 50.61screening limit 

before EOL, several plants are close to the limit (3 are within 2°F, while 10 are within 20°F). Those plants 

are likely to exceed the screening limit during the 20-year license renewal period that is currently being 

sought by many operators. Moreover, some plants maintain their RTPTS values below the 10 CFR 50.61 

screening limits by implementing flux reductions (low-leakage cores, ultra-low-leakage cores), which are 

fuel management strategies that can be economically deleterious in a deregulated marketplace. Thus, 

the 10 CFR 50.61 screening limits can restrict both the licensable and economic lifetime of PWRs. 

Motivation for this Project 

It is now widely recognized that the state of knowledge and data limitations in the early 1980s 

necessitated conservative treatment of several key parameters and models used in the probabilistic 

calculations that provided the technical basis for the current PTS Rule. The most prominent of these 

conservatisms include the following factors: 

• highly simplified treatment of plant transients (very coarse grouping of many operational sequences 

(on the order of 105

) into very few groups (≈10), necessitated by limitations in the computational 

resources needed to perform multiple thermal-hydraulic calculations) 

• lack of any significant credit for operator action 

• characterization of fracture toughness using RTNDT, which has an intentional conservative bias 

• use of a flaw distribution that places all flaws on the interior surface of the RPV, and, in general, 

contains larger flaws than those usually detected in service 

• a modeling approach that treated the RPV as if it were made entirely from the most brittle of its 

constituent materials (welds, plates, or forgings) 

• a modeling approach that assessed RPV embrittlement using the peak fluence over the entire interior 

surface of the RPV 
xxi

These factors indicate the high likelihood that the current 10 CFR 50.61 PTS screening limits are 

unnecessarily conservative. Consequently, the NRC staff believed that reexamining the technical basis 

for these screening limits, based on a modern understanding of all the factors that influence PTS, 

would most likely provide strong justification for substantially relaxing these limits. For these reasons, 

RES undertook this study with the objective of developing the technical basis to support a risk-informed 

revision of the PTS Rule and the associated PTS screening limit. 

Approach 

As illustrated in the following figure, three main models (shown as solid blue squares), taken together, 

allow us to estimate the annual frequency of through-wall cracking in an RPV: 

• probabilistic risk assessment (PRA) event sequence analysis 

• thermal-hydraulic (TH) analysis 

• probabilistic fracture mechanics (PFM) analysis 

PRA Event

Sequence

Analysis

(SAPPHIRE)

Thermal

Hydraulic

Analysis

(RELAP)

Probabilistic

Fracture

Analysis

(FAVOR)

Sequence

Definitions

Sequence

Frequencies

freq

Conditional

Probability of

Thru-Wall

Cracking, CPTWC

P(t), T(t), &

HTC(t)

Yearly

Frequency of

Thru-Wall

Cracking

[CPTWC]

x

[freq]

Probabilistic Estimation of Through-Wall Cracking Frequency

Vessel damage, age, 

or operational metric

Yearly Frequency of

Thru-Wall Cracking

Screening 

Limit

Acceptance Criterion 

for TWC Frequency

Established consistent with

• 1986 Commission safety goal 

policy statement

• June 1990 SRM

• RG1.174

Screening Limit 

Development

Schematic showing how a probabilistic estimate of through-wall cracking 

frequency (TWCF) is combined with a TWCF acceptance criterion to arrive 

at a proposed revision of the PTS screening limit 

First, a PRA event sequence analysis is performed to define the sequences of events that are likely to cause 

a PTS challenge to RPV integrity, and estimate the frequency with which such sequences can be expected 

to occur. The event sequence definitions are then passed to a TH model that estimates the temporal variation 

of temperature, pressure, and heat-transfer coefficient in the RPV downcomer, which is characteristic of 

each sequence definition. These temperature, pressure, and heat-transfer coefficient histories are then 

passed to a PFM model that uses the TH output, along with other information concerning plant design 

and construction materials, to estimate the time-dependent “driving force to fracture” produced by a particular 

event sequence. The PFM model then compares this estimate of fracture driving force to the fracture toughness, 

or fracture resistance, of the RPV steel. This comparison allows us to estimate the probability that a crack 

could grow to sufficient size that it would penetrate all the way through the RPV wall if that particular 

sequence of events actually occured. The final step in the analysis involves a simple matrix multiplication 

of the probability of through-wall cracking (from the PFM analysis) with the frequency at which a 

particular event sequence is expected to occur (as defined by the event-tree analysis). This product 

establishes an estimate of the annual frequency of through-wall cracking that can be expected for a 

particular plant after a particular period of operation when subjected to a particular sequence of events. The 
xxii

annual frequency of through-wall cracking is then summed for all event sequences to estimate the total 

annual frequency of through-wall cracking for the vessel. Performance of such analyses for various 

operating lifetimes provides an estimate of how the annual frequency of through-wall cracking can be 

expected to vary over the lifetime of the plant. 

The probabilistic calculations just described are performed to establish the technical basis for a revised 

PTS Rule within an integrated systems analysis framework. Our approach considers a broad range of factors 

that influence the likelihood of vessel failure during a PTS event, while accounting for uncertainties 

in these factors across a breadth of technical disciplines. Two central features of this approach are a focus 

on the use of realistic input values and models (wherever possible), and an explicit treatment of uncertainties 

(using currently available uncertainty analysis tools and techniques). Thus, our current approach 

improves upon that employed in developing SECY-82-465, which included intentional and unquantified 

conservatisms in many aspects of the analysis, and treated uncertainties implicitly by incorporating them 

into the models. 

Key Findings 

The findings from this study are divided into the following five topical areas: (1) the expected magnitude 

of the through-wall cracking frequency (TWCF) for currently anticipated operational lifetimes, 

(2) the material factors that dominate PTS risk, (3) the transient classes that dominate PTS risk, (4) the 

applicability of these findings (based on detailed analyses of three PWRs) to PWRs in general, and (5) the 

annual limit on TWCF established consistent with current guidelines on risk-informed regulation. In this 

summary, the conclusions are presented in boldface italic, while the supporting information is shown in 

regular type. 

TWCF Magnitude for Currently Anticipated Operational Lifetimes 

• The degree of PTS challenge is low for currently anticipated lifetimes and operating conditions. 

o Even at the end of license extension (60 operational years, or 48 effective full-power years (EFPY) 

at an 80% capacity factor), the mean estimated TWCF does not exceed 2x10-8/year for the plants 

analyzed. Considering that the RPVs at the Beaver Valley Power Station and Palisades Nuclear 

Power Plant are constructed from some of the most irradiation-sensitive materials in commercial 

reactor service today, these results suggest that, provided that operating practices do not change 

dramatically in the future, the operating reactor fleet is in little danger of exceeding either 

the TWCF limit of 5x10-6/yr expressed by Regulatory Guide 1.154 [RG 1.154] or the value 

of 1x10-6/yr recommended in Chapter 10 of this report — even after license extension. 

Material Factors and their Contributions to PTS Risk 

• Axial flaws, and the toughness properties that can be associated with such flaws, control nearly all 

of the TWCF. 

o Axial flaws are much more likely than circumferential flaws to propagate through the RPV wall 

because the applied fracture driving force increases continuously with increasing crack depth 

for an axial flaw. Conversely, circumferentially oriented flaws experience a driving force peak 

mid-wall, providing a natural crack arrest mechanism. It should be noted that crack initiation 

from circumferentially oriented flaws is likely; it is only their through-wall propagation that is 

much less likely (relative to axially oriented flaws). 

o It is, therefore, the toughness properties that can be associated with axial flaws that control nearly 

all of the TWCF. These include the toughness properties of plates and axial welds at the flaw locations. 

Conversely, the toughness properties of both circumferential welds and forgings have little effect 

on the TWCF because these can be associated only with circumferentially oriented flaws. 
xxiii

Transients and their Contributions to PTS Risk 

• Transients involving primary side faults are the dominant contributors to TWCF, while transients 

involving secondary side faults play a much smaller role. 

o The severity of a transient is controlled by a combination of three factors: 

 initial cooling rate, which controls the thermal stress in the RPV wall 

 minimum temperature of the transient, which controls the resistance of the vessel to fracture 

 pressure retained in the primary system, which controls the pressure stress in the RPV wall 

o The significance of a transient (i.e., how much it contributes to PTS risk) depends on these three 

factors and the likelihood that the transient will occur. 

o Our analysis considered transients in the following classes (as shown in the following table): 

 primary side pipe breaks 

 stuck-open valves on the primary side 

 main steam line breaks 

 stuck-open valves on the secondary side 

 feed-and-bleed 

 steam generator tube rupture 

 mixed primary and secondary initiators 

Factors contributing to the severity and risk-dominance of various transient classes 

Transient Severity 

Transient Class 

Cooling Rate Minimum 

Temperature Pressure 

Transient 

Likelihood 

TWCF 

Contribution 

Large-Diameter Fast Low Low Low Large 

Medium-Diameter Moderate Low Low Moderate Large Primary Side 

Pipe Breaks 

Small-Diameter Slow High Moderate High ~0 

Stuck-Open Valve Recloses Slow Moderate High High Large 

Valves, 

Primary Side Valve Remains Open Slow Moderate Low High ~0 

Main Steam Line Break Fast Moderate High High Small 

Stuck-Open Valve(s), Secondary Side Moderate High High High ~0 

Feed-and- Bleed Slow Low Low Low ~0 

Steam Generator Tube Rupture Slow High Moderate Low ~0 

Mixed Primary & Secondary Initiators Slow Mixed Very Low ~0 

Color Key Enhances TWCF Contribution Intermediate Diminishes TWCF Contribution 

o The table above provides a qualitative summary our results for these transient classes in terms of 

both transient severity and the likelihood that the transient will occur. The color-coding of table 

entries indicates the contribution (or lack thereof) of these factors to the TWCF of the various 

classes of transients. This summary indicates that the risk-dominant transients (medium- and largediameter primary side pipe breaks, and stuck-open primary side valves that later reclose) all have 

multiple factors that, in combination, result in their significant contributions to TWCF. 

 For medium- to large-diameter primary side pipe breaks, the fast to moderate cooling rates 

and low downcomer temperatures (generated by rapid depressurization and emergency injection 

of low-temperature makeup water directly to the primary) combine to produce a high-severity 

transient. Despite the moderate to low likelihood that these transients will occur, their severity 

(if they do occur) makes them significant contributors to the total TWCF. 
xxiv

 For stuck-open primary side valves that later reclose, the repressurization associated with 

valve reclosure coupled with low temperatures in the primary combine to produce a highseverity transient. This, coupled with a high likelihood of transient occurrence, makes stuckopen primary side valves that later reclose significant contributors to the total TWCF. 

 The small or negligible contribution of all secondary side transients (main steam line break, 

stuck-open secondary valves) results directly from the lack of low temperatures in the primary 

system. For these transients, the minimum temperature of the primary for times of relevance 

is controlled by the boiling point of water in the secondary (212°F (100°C) or above). 

At these temperatures, the fracture toughness of the RPV steel is sufficiently high to resist 

vessel failure in most cases. 

Applicability of These Findings to PWRs in General 

• Credits for operator action, while included in our analysis, do not influence these findings in any 

significant way. Operator action credits can dramatically influence the risk-significance of individual

transients. Therefore, appropriate credits for operator action need to be included as part of a “best estimate” 

analysis because there is no way to establish a priori if a particular transient will make a large contribution 

to the total risk. Nonetheless, the results of our analyses demonstrate that these operator action credits 

have a small overall effect on a plant’s total TWCF, for reasons detailed below.

o Medium- and Large-Diameter Primary Side Pipe Breaks: No operator actions are modeled 

for any break diameter because, for these events, the safety injection systems do not fully refill 

the upper regions of the reactor coolant system (RCS). Consequently, operators would never 

take action to shut off the pumps. 

o Stuck-Open Primary Side Valves that May Later Reclose: Reasonable and appropriate credit 

for operator actions (throttling of the high-pressure injection (HPI) system) has been included 

in the PRA model. However, these credits have a small influence on the estimated values 

of vessel failure probability attributable to transients caused by a stuck-open valve in the primary 

pressure circuit (SO-1 transients) because the credited operator actions only prevent repressurization 

when SO-1 transients initiate from Hot Zero Power (HZP) conditions and when the operators 

act promptly (within 1 minute) to throttle the HPI. Complete removal of operator action credits 

from the model only slightly increases the total risk associated with SO-1 transients. 

o Main Steam Line Breaks: For the overwhelming majority of transients caused by a main steam line 

break (MSLB), vessel failure is predicted to occur between 10 and 15 minutes after transient initiation 

because the thermal stresses associated with the rapid cooldown reach their maximum within this 

timeframe. Thus, all of the long-term effects (isolation of feedwater flow, timing of HPSI control) 

that can be influenced by operator actions have no effect on vessel failure probability because 

such factors influence the progression of the transient after failure has occurred (if it occurs at all). 

Only factors affecting the initial cooling rate (i.e., plant power level at time of transient initiation, 

break location inside or outside of containment) can influence the conditional probability 

of through-wall cracking (CPTWC), and operator actions do not influence such factors in any way. 

• Because the severity of the most significant transients in the dominant transient classes is controlled 

by factors that are common to PWRs in general, the TWCF results presented herein can be used 

with confidence to develop revised PTS screening criteria that apply to the entire fleet of operating 

PWRs. 

o Medium- and Large-Diameter Primary Side Pipe Breaks: For these break diameters, the fluid 

in the primary cools faster than the wall of the RPV. In this situation, only the thermal conductivity 

of the steel and the thickness of the RPV wall control the thermal stresses and, thus, the severity 

of the fracture challenge. Perturbations in the fluid cooldown rate controlled by break diameter, 

break location, and season of the year do not play a role. Thermal conductivity is a physical property, 
xxv

so it is very consistent for all RPV steels, and the thicknesses of the three RPVs analyzed are typical 

of PWRs. Consequently, the TWCF contribution of medium- to large-diameter primary side 

pipe breaks is expected to be consistent from plant-to-plant and can be well represented for all PWRs 

by the analyses reported herein. 

o Stuck-Open Primary Side Valves that May Later Reclose: A major contributor to the risk-significance 

of SO-1 transients is the return to full system pressure once the valve recloses. The operating 

and safety relief valve pressures of all PWRs are similar. Additionally, as previously noted, 

operator action credits only slightly affect the total risk associated with this transient class. 

o Main Steam Line Breaks: Since MSLBs fail early (within 10–15 minutes after transient initiation), 

only factors affecting the initial cooling rate can have any influence on the CPTWC values. 

These factors, which include the plant power level at event initiation and the location of the break 

(inside or outside of containment), are not influenced by operator actions in any way. 

• Sensitivity studies performed on the TH and PFM models to investigate the effect of credible model 

variations on the predicted TWCF values revealed no effects significant enough to recommend 

changes to the baseline RELAP and FAVOR models, or to recommend cautions regarding 

the robustness of those models. 

• An investigation of design and operational characteristics for five additional PWRs revealed 

no differences in sequence progression, sequence frequency, or plant thermal-hydraulic response 

significant enough to call into question the applicability of the TWCF results from the three 

detailed plant analyses to PWRs in general. 

• An investigation of potential external initiating events (e.g., fires, earthquakes, floods) revealed 

that the contribution of those events to the total TWCF can be regarded as negligible. 

Annual Limit on TWCF 

• The current guidance provided by Regulatory Guide 1.174 [RG 1.174] for large early release 

is appropriately applied to setting an acceptable annual TWCF limit of 1x10-6 events/year. 

o While many post-PTS accident progressions led only to core damage (which suggests a TWCF limit 

of 1x10-5 events/year limit in accordance with Regulatory Guide 1.174), uncertainties in 

the accident progression analysis led to our recommendation to adopt the more conservative limit 

of 1x10-6 events/year based on LERF. 

Recommended Revision of the PTS Screening Limits 

We recommend using different reference temperature (RT) metrics to characterize an RPV’s resistance 

to fractures initiating from different flaws at different locations in the vessel. Specifically, we recommend 

a reference temperature for flaws occurring along axial weld fusion lines (RTAW or RTAW-MAX), 

another for flaws occurring in plates or in forgings (RTPL or RTPL-MAX), and a third for flaws occurring 

along circumferential weld fusion lines (RTCW or RTCW-MAX). In each of these reference temperature pairs, 

the first metric is a weighted value that accounts for the differences between plants in weld fusion line 

area or plate volume, while the second metric is a maximum value that can be estimated based only on 

the information in the NRC’s Reactor Vessel Integrity Database (RVID). We also recommend using 

different RT values together to characterize the fracture resistance of the vessel’s beltline region, 

in recognition of the fact that the probability of vessel fracture initiating from different flaw populations 

varies considerably in response to factors that are both understood and predictable. Correlations between 

these RT metrics and the TWCF attributable to axial weld flaws, plate flaws, and circumferential weld flaws 

show little plant-to-plant variability because of the general similarity of PTS challenges among plants. 
xxvi

RT-based screening limits were established by setting the total TWCF (i.e., that attributable to axial weld flaws 

and plate flaws and circumferential weld flaws) equal to the reactor vessel failure frequency acceptance 

criterion of 1x10-6 events per year. The following figures graphically represent these screening limits 

(for the maximum RT metrics), along with an assessment of all operating PWRs relative to these limits. 

In these figures, the region of the graphs between the red locus and the origin has TWCF values below 

the 1x10-6 acceptance criterion, so these combinations of reference temperatures would be considered 

acceptable and require no further analysis. By contrast, the region of the graph outside of the red locus 

has TWCF values above the 1x10-6 acceptance criterion, indicating the need for additional analysis 

or other measures to justify continued plant operation. Clearly, operating PWRs do not closely approach 

the 1x10-6/year limit. At EOL, at least 70°F, and up to 290°F, (39 to 161°C) separate plate-welded PWRs 

from the proposed screening limit; this separation between plant-specific values and the proposed 

screening limit reduces by 10–20°F (5.5 to 11°C) at end of license extension (EOLE, defined as 60 

operating years or 48 EFPY). Additionally, no forged plant is anywhere close to the limit of 1x10-6 events 

per year at either EOL or EOLE. This separation of operating plants from the screening limit contrasts 

markedly with the current situation, where the most embrittled plants are within 1°F (0.5°C) of the 

screening limit set forth in 10 CFR 50.61. These differences in the “proximity” of operating plants to the 

current (10 CFR 50.61) and proposed screening limits are illustrated by the bar graph on the next page. 

0

50

100

150

200

250

300

350

0 50 100 150 200 250 300

RTMAX-AW [oF]

RTMAX-PL [oF]

Plate 

PWRs 

at EOL

1E-8

1E-7

5E-7 1E-6

Forged 

PWRs 

at EOL

0

50

100

150

200

250

300

350

-50 50 150 250 350 450

RTMAX-CW [

o

F]

Forging: RTMAX-PL [oF]

1E-8

1E-7

5E-7 1E-6

Comparison of RT-based screening limits (curves) 

with assessment points for operating PWRs at EOL 

(Left: plate vessels, Right: forged vessels) 
xxvii

0

4

8

12

16

Number of PWRs

Relative to RTPTS limits in 10CFR50.61

Relative to RT-MAX limits proposed in NUREG-1806 

0 120 200 280 320 360 400 40 80 160 240

oF from PTS Screening Limit after 40 Years of Operation

Relative to Current 10CFR50.61 Limits

Relative to NUREG-1806 Proposed Limits

0

4

8

12

16

Number of PWRs

Relative to RTPTS limits in 10CFR50.61

Relative to RT-MAX limits proposed in NUREG-1806 

0 120 200 280 320 360 400 40 80 160 240

oF from PTS Screening Limit after 40 Years of Operation

Relative to Current 10CFR50.61 Limits

Relative to NUREG-1806 Proposed Limits

Difference between the proximity of operating PWRs to the current RTPTS screening limits 

and to the screening limits proposed based on the work presented in this report. 

These RT-based screening limits (and similar limits described in the text for application to weighted 

RT values) apply to PWRs in general, subject only to the following provisos: 

• When assessing a forged vessel where the forging has a very high reference temperature (RTPL above 

225°F (107°C)) and the forging is believed to be susceptible to subclad cracking, a plant-specific 

analysis of the TWCF produced by the subclad cracks should be performed. However, no forging 

is projected to reach this level of embrittlement, even at EOLE. 

• When assessing an RPV having a wall thickness of 7-in. (18-cm) or less (7 vessels), the proposed 

RT limits are conservative. 

• When assessing an RPV having a wall thickness of 11-in. (28-cm) or greater, the proposed RT limits 

may be nonconservative. For the three plants meeting this criterion, either the RT limits would need 

to be reduced or known conservatisms in the current analysis would have to be removed to demonstrate 

compliance with the TWCF limit of 1x10-6 event/year. However, because these three plants 

are Units 1, 2, and 3 of the Palo Verde Nuclear Generating Station, which have vessels with very low 

embrittlement projected at EOL and EOLE, there is little practical need for such plant-specific analysis. 

Aside from relying on different RT metrics than 10 CFR 50.61, this proposed revision of the PTS screening limit 

differs from the current screening limit in the absence of a “margin term.” Use of a margin term is appropriate 

to account (at least approximately) for factors that occur in application but were not considered in the analysis 

upon which the screening limit is based. For example, the 10 CFR 50.61 margin term accounts for uncertainty 

in copper, nickel, and initial RTNDT. However, our model explicitly considers uncertainty in all of these 

variables, and represents these uncertainties as being larger (a conservative representation) than would be 

appropriate in any plant-specific application of the proposed screening limit. Consequently, use of 

the 10 CFR 50.61 margin term with the new screening limits is inappropriate. In general, the following 

additional reasons suggest that use of any margin term with the proposed screening limits is inappropriate: 
xxviii

(1) The TWCF values used to establish the screening limit represent 90th percentile values or greater. 

(2) The results from our three plant-specific analyses apply to PWRs in general, as demonstrated 

in Chapters 8 and 9 of this report. 

(3) Certain aspects of our modeling cannot reasonably be represented as “best estimates.” On balance, 

there is a conservative bias to these non-best-estimate aspects of our analysis because residual 

conservatisms in the model far outweigh residual nonconservatisms. 
xxix

Abbreviations 

1⁄4-T FLAW Surface-breaking flaw defined by ASME Boiler and Pressure Vessel Code 

as having a depth equal to one-quarter of the vessel wall thickness 

and a length equal to six times the flaw depth 

1D One-Dimensional 

ABAQUS Commercial finite element code developed by Hibbett, Karlsson, 

and Sorenson in Pawtucket, Rhode Island 

ACRS Advisory Committee on Reactor Safety (NRC) 

ADV Atmospheric Dump Valve 

AFW Auxiliary Feedwater 

APET Accident Progression Event Tree 

APEX Advanced Plant Experiment 

ASME American Society of Mechanical Engineers 

ASTM American Society for Testing and Materials 

ATWS Anticipated Transient without Scram 

B&W Babcock and Wilcox 

BWOG Babcock and Wilcox Owners’ Group 

BCC Body-Centered Cubic 

BWR Boiling-Water Reactor 

CDF Core Damage Frequency 

CE Combustion Engineering 

CEOG Combustion Engineering Owners’ Group 

CFD Computational Fluid Dynamics 

CL Cold Leg 

CFR Code of Federal Regulations

CFT Core Flood Tank 

CPI Conditional Probability of Crack Initiation 

CPTWC Conditional Probability of Through-Wall Cracking 

CSAU Code Scaling, Applicability, and Uncertainty Methodology 

CSNI Committee on the Safety of Nuclear Installations 

CST Condensate Storage Tank 

CVN Charpy V-Notch 

ECC Emergency Core Cooling 

ECCS Emergency Core Cooling System 

EFPY Effective Full-Power Years 

EFW Emergency Feedwater 

EOL End of License (40 operating years, 32 EFPY) 

EOLE End of License Extension (60 operating years, 48 EFPY) 
xxx

EPRI Electric Power Research Institute 

ESFAS Engineered Safety Features Actuation System 

F&B Feed-and-Bleed 

FAVOR Fracture Analysis of Vessels, Oak Ridge 

FCI Frequency of Crack Initiation 

GMAW Gas Metal Arc Weld 

H2TS Hierarchical, Two-Tiered Scaling 

HCLPF High Confidence of Low Probability of Failure 

HEP Human Error Probability 

HFE Human Failure Event 

HPI High-Pressure Injection 

HPSI High-Pressure Safety Injection 

HRA Human Reliability Analysis 

HSSI Heavy Section Steel Irradiation (Project) 

HZP Hot Zero Power 

IAEA International Atomic Energy Agency 

ID Inner Diameter 

IPE Individual Plant Examination 

IPEEE Individual Plant Examination of External Events 

IPTS Integrated Pressurized Thermal Shock 

ISLOCA Interfacing Systems Loss-of-Coolant Accident 

ITV Intermediate Test Vessel 

IVO Imatran Voima Oy 

LAS Low-Alloy Steel 

LBLOCA Large-Break Loss-of-Coolant Accident (pipe diameters above ~8-in. (~20-cm)) 

LEFM Linear Elastic Fracture Mechanics 

LER Licensee Event Report 

LERF Large Early Release Frequency 

LOCA Loss-of-Coolant Accident 

LOF Lack of Inter-Run Fusion 

LOFT Loss-of-Fluid Test facility 

LPI Low-Pressure Injection 

LPSI Low-Pressure Safety Injection 

MBLOCA Medium-Break Loss-of-Coolant Accident (pipe diameters of ~4 to 8-in. 

(~10 to 20-cm)) 

MFIV Main Feedwater Isolation Valve 

MFW Main Feedwater 

MIST Multi-loop Integral System Test 

MRJ Materials Reliability Project 

MSIV Main Steam Isolation Valve 

MSLB Main Steam Line Break 
xxxi

NDT Nil-Ductility Temperature 

NEA Nuclear Energy Agency (OECD) 

NRC U.S. Nuclear Regulatory Commission 

NRR Office of Nuclear Reactor Regulation (NRC) 

NUREG/CR NRC Technical Report Designator (Contractor-prepared Report 

published by the U.S. Nuclear Regulatory Commission) 

OD Outer Diameter 

OECD Organization for Economic Cooperation and Development 

ORNL Oak Ridge National Laboratory 

PFM Probabilistic Fracture Mechanics 

PIRT Phenomena Identification and Ranking Table 

PNNL Pacific Northwest National Laboratories 

PORV Power-Operated Relief Valve 

Ppb Parts per Billion 

PRA Probabilistic Risk Assessment 

PRODIGAL Probability of Defect Initiation and Growth Analysis 

PTS Pressurized Thermal Shock 

PTSE Pressurized Thermal Shock Experiment 

PVRUF Pressure Vessel Research Users’ Facility 

PWR Pressurized-Water Reactor 

QHO Quantitative Health Objective, as defined by the Commission’s Safety Goal 

Policy Statement [NRC FR 86] 

RCP Reactor Coolant Pump 

RCS Reactor Coolant System 

RELAP Reactor Leak and Power excursion code 

REMIX a computer program used to determine the temperature of a plume 

in the downcomer when the flow in the loops is stagnant 

RES Office of Nuclear Regulatory Research (NRC) 

RG Regulatory Guide 

RLE Review-Level Earthquake 

ROSA Rig of Safety Assessment 

RPS Reactor Protection System 

RPV Reactor Pressure Vessel 

RT Reference Temperature 

RVFF Reactor Vessel Failure Frequency 

RVID Reactor Vessel Integrity Database 

RWST Refueling Water Storage Tank 

SAPHIRE Systems Analysis Programs for Hands-on Integrated Reliability Evaluations 

SAW Submerged Arc Weld 

SBLOCA Small-Break Loss-of-Coolant Accident (pipe diameters below ~4-in. (~10-cm)) 

SCC Stress Corrosion Cracking 

SECY Secretary of the (U.S. Nuclear Regulatory) Commission 
xxxii

SEMISCALE a 1:1705 scaled experimental facility that simulates the primary system 

of a 4-loop PWR plant 

SG Steam Generator 

SGTR Steam Generator Tube Rupture 

SIAS Safety Injection Actuation Signal 

SIT Safety Injection Tank 

SMAW Submerged Metal Arc Weld 

SO-1 Stuck-open valve in the primary pressure circuit 

SO-2 Stuck-open valve in the secondary pressure circuit 

SQA Software Quality Assurance 

SRM Staff Requirements Memorandum 

SRV Safety/Relief Valve 

SSC System, Structure, or Component 

SSE Safe-Shutdown Earthquake 

SSRV Secondary System Relief Valve 

TBV Turbine Bypass Valve 

TH Thermal-Hydraulics 

TMI Three Mile Island 

TSE Thermal Shock Experiment 

TWCF Through-Wall Cracking Frequency 

UMD University of Maryland 

UPTF Upper Plenum Test Facility 

USE Charpy V-Notch Upper-Shelf Energy 

V&V Verification and Validation 

VCIF Vessel Crack Initiation Frequency 

(W) Westinghouse 

WOG Westinghouse Owners’ Group 

WPS Warm Pre-Stress 
xxxiii

Nomenclature 

Symbols Used in Thermal-Hydraulics 

α thermal diffusivity, m2

/s 

β bulk coefficient of expansion, 1/C

μ viscosity, kg/m-s 

ν kinematic viscosity, m2

/s 

ρ density, kg/m3

σ stress, kg/s2

τ characteristic time 

Cp heat capacity, m2

/s2-C 

g gravitational acceleration, m/s2

Gr Grashof Number 

h convective heat transfer coefficient, W/m2

-C

D diameter, m 

J joules, kg-m2

/s2

k conductivity, W/m-C

l length, m 

Nu Nusselt Number 

Pr Prandtl Number 

P pressure, kg/m-s2

q heat flux, W/m2

Re Reynolds Number 

Ri Richardson Number 

s seconds 

t thickness, m 

t time, s 

u velocity, m/s 

T temperature, C 

W watts, kg-m2

/s3
xxxiv

Symbols Used in Fracture Mechanics 

2a Flaw depth measured through the vessel wall thickness 

2c Flaw length measured parallel to the axial or circumferential direction 

of the vessel 

Cu Copper content, weight% 

JIc A fracture toughness measure defined by ASTM E1820, which quantifies 

the resistance of metals to crack initiation by the initiation, growth, 

and coalescence of microvoids 

J-R A fracture toughness measure defined by ASTM E1820, which quantifies 

the resistance of metals to ductile tearing 

KJc A fracture toughness measure defined by ASTM E1921, which quantifies 

the resistance of metals to crack initiation by cleavage mechanisms 

KIa A fracture toughness measure defined by ASTM E1221, which quantifies 

the ability of metals to arrest (stop) a running cleavage crack 

KIc A fracture toughness measure defined by ASTM E399, which quantifies 

the resistance of metals to crack initiation under plane strain conditions 

KIc(min) The minimum KIc fracture toughness possible at a particular temperature 

KAPPLIED Linear elastic crack driving force 

L For a buried defect, distance from the wetted clad surface on the vessel ID 

to the inner crack tip 

l The length of the fusion line of an axial weld 

Ni Nickel content, weight% 

P Phosphorus content, weight% 

RTAW A fracture toughness reference temperature, which characterizes the RPV’s 

resistance to fractures initiating from flaws found along the axial weld fusion 

lines. It corresponds to the maximum RTNDT of the plates/welds that lie 

to either side of the weld fusion lines, and is weighted to account for differences 

in weld fusion line length (and, therefore, number of simulated flaws) 

between vessel courses. 

RTPL A fracture toughness reference temperature, which characterizes the RPV’s 

resistance to fractures initiating from flaws found in plates that are not 

associated with welds. It corresponds to the maximum RTNDT occurring 

anywhere in the plate. 

RTCW A fracture toughness reference temperature, which characterizes the RPV’s 

resistance to fractures initiating from flaws found along the circumferential 

weld fusion lines. It corresponds to the maximum RTNDT of the plates/welds 

that lie to either side of the weld fusion lines. 

RTNDT Transition fracture toughness reference temperature defined by 

ASME NB-2331 

RTNDT(u) Unirradiated value of RTNDT

RTPTS RTNDT projected end of license to account for the effects of irradiation 

(defined in 10 CFR 50.61) 

tWALL Vessel wall thickness 

tCLAD Stainless steel cladding thickness 
xxxv

T30 The temperature at which the mean CVN energy is 30 ft-lbs (41J) 

T35/50 Charpy V-notch energy transition temperature defined as the temperature 

at which the CVN energy is at least 50 ft-lbs (68J) and the lateral expansion 

of the specimen is at least 0.035-in. (0.89-mm) [See the definition on page 2-

7]

TNDT Nil-ductility temperature defined by ASTM E-208 

ΔT30 The shift in the CVN 30 ft-lb (41J) transition temperature produced by 

radiation damage 

σflow Flow strength, average of tensile yield and tensile ultimate strength 

φt Fluence 
xxxvi

Glossary 

Terms Used in Probabilistic Risk Assessment 

Abnormal operating procedure A procedure (i.e., list of actions) used to address unique or special plant 

circumstances identified while using emergency operating procedures (EOPs). 

These abnormal operating procedures are usually called by EOPs, but may be 

indicated directly by some plant conditions. 

Accident progression event tree The event tree used to model the part of the accident sequence that follows 

the onset of core damage, including containment response to severe accident 

conditions, equipment availability, and operator performance. 

Binning The process of taking a large number of sequences and combining then into 

a smaller number of groups, that are expected to have similar characteristics 

(e.g., TH conditions), to allow effective utilization of limited resources. 

Core damage Uncovery and heatup of the reactor core to the point at which prolonged oxidation 

and severe fuel damage is anticipated and involving enough of the core to cause 

a significant release. 

Dominant scenario An accident sequence (scenario) that is usually represented by the top 10 or 20 events 

or groups of events modeled in a PRA, which accounts for a large fraction 

of the specified end state. 

Emergency operating procedure The primary procedure (i.e., list of actions) used to respond to a plant disturbance 

resulting from an initiating event. 

Event tree A logic diagram that begins with an initiating event or condition and progresses 

through a series of branches that represent expected system or operator performance 

that either succeeds or fails and arrives at either a successful or failed end state. 

Fault tree A deductive logic diagram that depicts how a particular undesired event can occur 

as a logical combination of other undesired events. 

Large Early Release The rapid, unmitigated release of airborne fission products from the containment 

to the environment occurring before the effective implementation of offsite 

emergency response and protective actions, such that there is a potential for 

early health effects. 

Latin Hypercube sampling A stratified sampling technique, in which the random variable distributions 

are divided into equal probability intervals, and probabilities are then randomly 

selected from within each interval. 

Mitigating equipment Systems or components, used to respond to an initiating event, of which 

successful operation prevents the occurrence of an undesired event or state. 

Pre-initiator human failure event Human failure events that represent the impact of human errors committed 

during actions performed prior to the initiation of an accident (e.g., during 

maintenance or the use of calibration procedures). 

Post-initiator human failure event Human failure events that represent the impact of human errors committed 

during actions performed in response to an accident initiator. 
xxxvii

Prompt fatality A fatality that results from substantial radiation exposures incurred during 

short time periods (usually within weeks, though up to 1 year for pulmonary 

effects). 

PTS bin A group of sequences that are expected to have similar TH characteristics 

and are represented by one unique set of TH characteristics during a FAVOR 

calculation. 

Risk-informed An approach to analyzing and evaluating activities, which bases decisions 

on the results of traditional engineering evaluations, supported by insights 

derived from the use of PRA methods. 

Scenario See Sequence. 

Screening The process of eliminating items from further consideration based on their 

negligible contribution to the probability of an undesired end state or its 

consequences. 

Sequence A representation in terms of an initiating event followed by a sequence 

of failures or successes of events (i.e., system, function, or operator performance) 

that can lead to undesired consequences, with a specified end state 

(e.g., potential for PTS). 

Terms Used in Thermal-Hydraulics 

Blowdown Rapid depressurization of a system in response to a break. 

Break flow Flow of water (liquid and vapor) out a pipe break or a valve. 

Break energy Energy content of the fluid flow out a break. 

Bottom-up To break up a complex system into its subsystems, and then break up each subsystem 

into its components, examine individual local phenomena and processes that 

most affect each component, and build up the total complex system from these 

individual pieces (like manufacturing a car). 

Coast down Time required for a pump to stop rotating once power is shut off due to inertia. 

Decay heat Heat generated from radioactive decay of fission products. 

Enthalpy Sum of internal energy and volume multiplied by pressure. 

Flash Change of phase from saturated liquid to vapor resulting from decrease in pressure. 

Flow quality Mass fraction of flow stream that is steam. Higher quality flow would have 

a high mass fraction of steam. 

Forced flow Flow driven by a pump. 

Inventory Mass of water. 

Loop flow Mass flow rate of coolant in a circuit. 

Makeup water Water reservoir available for inventory control. 

Natural circulation Flow driven by buoyancy (gravity). 

Pressure drop Change in pressure due to conversion of mechanical energy to internal energy. 

Protection system Electrical controls to actuate engineering safety features. 

Quality Mass fraction of steam in a two-phase steam-water mixture. 
xxxviii

Saturation temperature A temperature corresponding to phase change from liquid to vapor. 

Sensible heat The product of specific heat and temperature change of subcooled liquid. 

Subcooled A system is subcooled if it exists entirely in a liquid state. The degree 

of subcooling is the number of degrees that the temperature of the system 

would have to be raised to cause boiling. 

Throttled Operation of a control valve to regulate flow. 

Top-down To characterize a complex system by establishing the governing behavior, 

or phenomenon, that is most important, and then proceed from that starting point 

to successive lower levels, by identifying the processes that have the greatest 

influence on the top-level phenomenon. 

Trip A “trip” occurs when a breaker opens in response to its trip mechanism 

(an arm that holds the breaker closed moves to allow the breaker to open). 

When a reactor trips, all of the breakers that provide power to the rod control 

system open, causing the rods to be inserted in the core and stopping the nuclear 

reaction. When a pump trips, the breaker opens, thereby disconnecting power 

and causing the pump to stop. 

Water solid A situation in which there is no steam in the system (i.e., it is all liquid). 

A “water solid” system is subcooled. 

Terms Used in Fracture Mechanics 

Brittle Fracture occurring without noticeable macroscopic plastic deformation 

(stretching) of the material. 

Cleavage fracture Microscopically, cleavage is a fracture mode that occurs preferentially along 

certain atomic planes through the grains of the material. Cleavage can only 

occur in ferritic steels (i.e., steels having a body-centered cubic lattice structure). 

Macroscopically, cleavage fracture is often called “brittle” fracture because 

little noticeable plastic deformation (stretching) of the material occurs. 

(Note, however, that plastic flow at the micro-scale is a necessary precursor 

to cleavage.) Macroscopically, cleavage fracture is also characterized as being 

a sudden event, with cracks of very large dimensions developing over durations 

measured in fractional seconds. A useful, although inexact, analogue 

for cleavage fracture in common experience is the breaking of glass. 

Ductile fracture Microscopically, ductile fracture occurs through the initiation, growth, 

and eventual coalescence of micro-voids in the material into a macroscopic crack. 

These micro-voids tend to initiate at local heterogeneities in the material 

(e.g., inclusions, carbides, clusters of dislocations). Macroscopically, ductile fracture 

is associated with considerable plastic deformation (stretching) of the material. 

Relative to cleavage fracture, ductile fracture occurs very slowly, with crack 

growth rates measured in seconds rather than in micro-seconds (for cleavage). 

Fracture toughness A general term referring to a material’s resistance to fracture. The term may be 

modified to refer to fractures by different mechanisms: 

Arrest fracture toughness measures a material’s ability to stop a running 

cleavage crack. 

Cleavage fracture toughness measures a material’s ability to resist 

crack initiation in cleavage. 

Ductile fracture toughness measures a material’s ability to resist crack initiation 

attributable to ductile mechanisms on the upper shelf. 
xxxix

Lower shelf At low temperatures, the toughness behavior of steels occurs by transgranular 

cleavage and is said to be on the lower shelf. On the lower shelf, a fracture is 

unstable, and is often referred to as a “brittle” fracture. 

Reference temperature A characteristic temperature used to locate the transition curve of a ferritic steel 

on the temperature axis. 

Transition (or transition curve) Between lower shelf and upper shelf temperatures, the fracture behavior 

of a ferritic material is said to be in “transition.” At low temperatures in transition, 

fracture occurs by cleavage. As temperature increases through the transition regime, 

fracture occurs by ductile crack initiation and growth, a process which is terminated 

by cleavage. At still higher temperatures, cleavage cannot occur, and upper shelf 

conditions exist. 

Upper shelf At high temperatures, the toughness behavior of steels occurs by ductile mechanisms 

(micro-void initiation, growth, and coalescence) and is said to be on the upper shelf. 

On the upper shelf, afracture is stable and dissipates considerable amounts of energy. 

Terms Used in Uncertainty Analysis 

Aleatory Aleatory uncertainties arise as a result of the randomness inherent in a physical 

or human process. Consequently, aleatory uncertainties are fundamentally 

irreducible. If the uncertainty in a variable is characterized as being aleatory, 

the entire distribution of the variable is carried through each simulation run. 

Epistemic Epistemic uncertainties are caused by limitations in our current state of knowledge 

(or understanding) of a given process. Epistemic uncertainties can, in principle, 

be reduced by an increased state of knowledge. If the uncertainty in a variable 

is characterized as being epistemic in a probabilistic simulation, individual values 

of the variable are randomly selected from a distribution and propagated through 

the calculation. This procedure models the understanding that the “correct” value 

of the variable is knowable, at least in principal. Thus, for epistemic uncertainties, 

individual simulation runs are deterministic, while the totality of all simulation runs 

captures the uncertainty characteristic of the epistemic variable. 
A-1 

Appendix A – Master Transient List and FAVOR 04.1 

Results Summary

Table A.1. Transient descriptions and FAVOR 04.1 results for medium- and large-diameter 

pipe break (LOCA) transients 

Percent 

Contribution to 

Total Frequency of 

Crack Initiation 

(FCI) 

Percent 

Contribution to 

Total Through Wall 

Cracking 

Frequency (TWCF) 

Mean CPI Mean CPTWC 

# 

Dia [in] System Failure 

IEF 

32 

60 

Ext-A 

Ext-B 

32 

60 

Ext-A 

Ext-B 

32 

60 

Ext-A 

Ext-B 

32 

60 

Ext-A 

Ext-B 

Beaver Valley Unit 1 

9 

16 2.54-cm [16-in.] hot leg break 

6.99E-06 

5.0 

5.8 

6.5 

7.6 

1.5 

3.3 

10.5 

10.2 

6.40E-04 

3.02E-03 

1.21E-02 

7.44E-02 

8.63E-07 

1.43E-05 

1.58E-04 

2.66E-03 

7 

8 2.54-cm [8-in.] surge line break 

2.11E-05 

14.4 

15.7 

17.2 

17.4 

17.2 

28.6 

19.4 

14.9 

6.16E-04 

2.58E-03 

9.06E-03 

4.66E-02 

2.01E-06 

2.08E-05 

1.44E-04 

1.52E-03 

117 

5.7 

14.366-cm [5.657-in.] cold leg break, summer conditions (HHSI, LHSI temp = 55°F, Accumulator Temp = 105°F) 

2.11E-05 

0.2 

0.2 

0.2 

0.2 

0.0 

0.0 

0.0 

0.0 

1.27E-05 

4.66E-05 

1.36E-04 

6.45E-04 

8.29E-10 

2.29E-08 

9.46E-08 

1.14E-06 

116 

5.7 14.366-cm [5.657-in.] cold leg break with break area 

increased 30% 

1.81E-05 

0.0 

0.0 

0.1 

0.1 

0.0 

0.0 

0.0 

0.0 

1.27E-06 

9.46E-06 

4.85E-05 

4.33E-04 

2.15E-10 

1.27E-08 

6.23E-08 

2.16E-06 

56 

4 

10.16-cm [4.0-in.] surge line break (see Note at end of table) 

1.23E-04 

79.4 

77.1 

74.2 

68.9 

13.2 

30.8 

51.4 

43.2 

7.43E-04 

2.84E-03 

9.36E-03 

4.46E-02 

8.76E-07 

1.07E-05 

8.29E-05 

9.82E-04 
A-2 

Percent 

Contribution to 

Total Frequency of 

Crack Initiation 

(FCI) 

Percent 

Contribution to 

Total Through Wall 

Cracking 

Frequency (TWCF) 

Mean CPI Mean CPTWC 

# 

Dia [in] System Failure 

IEF 

32 

60 

Ext-A 

Ext-B 

32 

60 

Ext-A 

Ext-B 

32 

60 

Ext-A 

Ext-B 

32 

60 

Ext-A 

Ext-B 

Oconee Unit 1 

156 

16 

40.64-cm [16-in.] hot leg break. ECC suction switch to the containment sump included in the analysis. 

7.03E-06 

6.8 

13.2 

26.9 

28.5 

0.0 

0.1 

4.6 

6.3 

1.63E-06 

1.74E-05 

2.81E-03 

1.55E-02 

2.88E-10 

6.82E-09 

6.21E-06 

7.83E-05 

164 

8 

20.32-cm [8 in.] surge line break. ECC suction switch to the containment sump included in the analysis. 

2.12E-05 

41.2 

38.9 

36.1 

39.7 

0.1 

1.2 

17.1 

29.6 

1.20E-06 

1.09E-05 

1.29E-03 

7.06E-03 

7.78E-11 

9.05E-09 

6.85E-06 

8.05E-05 

160 

5.7 

14.37-cm [5.656-in.] surge line break. ECC suction switch to the containment sump included in the analysis. 

1.82E-05 

33.3 

39.8 

32.4 

26.3 

0.1 

0.6 

24.6 

32.8 

2.77E-06 

1.97E-05 

1.32E-03 

5.84E-03 

1.36E-09 

4.01E-08 

1.33E-05 

1.21E-04 

172 

4 

10.16-cm [4-in.] cold leg break. ECC suction switch to the containment sump included in the analysis. 

1.06E-04 

0.0 

0.1 

0.8 

1.5 

- 

- 

0.5 

1.2 

4.90E-10 

8.71E-09 

6.86E-06 

6.29E-05 

- 

- 

9.55E-08 

2.27E-06 

Palisades 

40 

16 

 40.64-cm (16-in.) hot leg break. Containment sump recirculation included in the analysis. 

3.22E-05 

65.3 

63.7 

58.2 

58.7 

14.4 

20.5 

27.7 

30.5 

7.60E-04 

1.76E-03 

1.03E-02 

6.42E-02 

1.36E-05 

7.11E-05 

1.21E-03 

9.57E-03 

62 

8 

20.32-cm (8-in.) cold leg break. Winter conditions assumed (HPI and LPI injection temp = 40 F, Accumulator temp = 60 F) 

7.07E-06 

6.9 

6.6 

6.6 

6.9 

2.1 

3.2 

4.2 

4.7 

3.73E-04 

8.85E-04 

5.31E-03 

3.37E-02 

1.08E-05 

5.50E-05 

8.57E-04 

6.59E-03 

63 

5.7 

14.37-cm (5.656-in.) cold leg break. Winter conditions assumed (HPI and LPI injection temp = 40 F, Accumulator temp = 60 F) 

6.06E-06 

2.0 

2.0 

2.3 

2.3 

0.8 

1.4 

2.0 

2.1 

1.41E-04 

3.37E-04 

2.08E-03 

1.30E-02 

5.00E-06 

2.72E-05 

4.53E-04 

3.30E-03 
A-3 

Percent 

Contribution to 

Total Frequency of 

Crack Initiation 

(FCI) 

Percent 

Contribution to 

Total Through Wall 

Cracking 

Frequency (TWCF) 

Mean CPI Mean CPTWC 

# 

Dia [in] System Failure 

IEF 

32 

60 

Ext-A 

Ext-B 

32 

60 

Ext-A 

Ext-B 

32 

60 

Ext-A 

Ext-B 

32 

60 

Ext-A 

Ext-B 

58 

4 

10.16-cm (4-in.) cold leg break. Winter conditions assumed (HPI and LPI injection temp = 40 F, Accumulator temp = 60 F) 

2.66E-04 

11.5 

11.7 

11.8 

11.3 

4.9 

8.8 

12.6 

12.4 

2.01E-05 

4.79E-05 

3.05E-04 

1.78E-03 

8.11E-07 

4.47E-06 

7.88E-05 

5.34E-04 

64 

4 

10.16-cm (4-in.) surge line break. Summer conditions assumed (HPI and LPI injection temp = 100 F, Accumulator temp = 90 F) 

7.07E-06 

4.0 

3.8 

3.4 

3.3 

2.2 

3.5 

3.7 

3.4 

2.35E-04 

5.28E-04 

2.95E-03 

1.68E-02 

1.21E-05 

6.03E-05 

7.92E-04 

4.88E-03 

59 

4 

10.16-cm (4-in.) cold leg break. Summer conditions assumed (HPI and LPI injection temp = 100 F, Accumulator temp = 90 F) 

2.09E-04 

0.6 

0.8 

1.0 

1.2 

0.1 

0.3 

0.7 

1.0 

1.34E-06 

3.94E-06 

3.32E-05 

2.15E-04 

2.00E-08 

1.72E-07 

5.73E-06 

5.37E-05 

Note: There are no operator actions for any of these transients, and all transients initiate from full power conditions except for Beaver 

Valley 56, which initiates from hot zero power conditions. However, Beaver Valley 56 is used to represent full power conditions in this analysis. 
A-4 

Table A.2. Transient descriptions and FAVOR 04.1 results for small-diameter pipe break (LOCA) transients 

Percent 

Contributio

n to Total 

Frequency 

of Crack 

Initiation 

(FCI) 

Percent 

Contributio

n to Total 

Through 

Wall 

Cracking 

Frequency 

(TWCF) 

Mean CPI Mean CPTWC 

# 

Dia [in] System Failure 

IEF 

32 

60 

Ext-A 

Ext-B 

32 

60 

Ext-A 

Ext-B 

32 

60 

Ext-A 

Ext-B 

32 

60 

Ext-A 

Ext-B 

Beaver Valley Unit 1 

114 

2.8 

7.18-cm [2.828-in.] surge line break, summer conditions (HHSI, LHSI temp = 55°F, Accumulator Temp = 105°F), heat transfer coefficient increased 30% (modeled by increasing heat transfer surface area by 30% in passive heat structures). 

9.76E-05 

0.24 

0.35 

0.54 

0.70 

0.00 

0.01 

0.03 

0.08 

2.55E-06 

1.52E-05 

6.70E-05 

4.57E-04 

6.63E-11 

8.69E-09 

7.25E-08 

2.45E-06 

115 

2.8 7.18-cm [2.828-in.] cold leg break 

9.76E-05 

- 

- 

- 

- 

- 

- 

- 

- 

- 

- 

- 

- 

- 

- 

- 

- 

3 

2 5.08-cm [2-in.] surge line break 

9.76E-05 

0.00 

0.02 

0.03 

0.08 

- 

0.00 

0.00 

0.01 

3.81E-08 

6.85E-07 

4.01E-06 

5.79E-05 

- 

2.28E-10 

2.18E-09 

3.57E-07 

2 

1.4 3.59-cm [1.414-in.] surge line break 

1.23E-04 

- 

- 

- 

- 

- 

- 

- 

- 

- 

- 

- 

- 

- 

- 

- 

- 

Oconee Unit 1 

154 

3.4 

8.53-cm [3.36-in.] surge line break [Break flow area reduced by 30% from 10.16-cm [4-in.] break]. Vent valves do not function. ECC suction switch to the containment sump included in the analysis. 

1.34E-04 

0.00 

0.00 

0.15 

0.36 

- 

- 

0.10 

0.79 

0.00E+00 

1.56E-10 

1.18E-06 

1.44E-05 

- 

- 

1.44E-08 

1.08E-06 

178 

3.4 

8.53-cm [3.36-in.] surge line break [Break flow area reduced by 30% from 10.16-cm [4-in.] break]. Vent valves do not function. ECC suction switch to the containment sump included in the analysis. 

2.12E-05 

0.00 

0.00 

0.04 

0.07 

- 

- 

0.03 

0.21 

0.00E+00 

1.56E-10 

1.18E-06 

1.44E-05 

- 

- 

1.44E-08 

1.08E-06 

141 

3.2 8.19-cm [3.22-in.] surge line break [Break flow area increased by 30% from 7.18-cm [2.828-in.] break]. 

1.06E-04 

0.64 

1.40 

1.71 

1.45 

0.00 

0.06 

1.92 

2.66 

2.00E-08 

2.05E-07 

1.54E-05 

6.44E-05 

2.40E-13 

7.57E-10 

3.77E-07 

4.21E-06 
A-5 

Percent 

Contributio

n to Total 

Frequency 

of Crack 

Initiation 

(FCI) 

Percent 

Contributio

n to Total 

Through 

Wall 

Cracking 

Frequency 

(TWCF) 

Mean CPI Mean CPTWC 

# 

Dia [in] System Failure 

IEF 

32 

60 

Ext-A 

Ext-B 

32 

60 

Ext-A 

Ext-B 

32 

60 

Ext-A 

Ext-B 

32 

60 

Ext-A 

Ext-B 

142 

2.4 6.01-cm [2.37-in.] surge line break [Break flow area decreased by 30% from 7.18-cm [2.828-in.] break]. 

1.06E-04 

- 

- 

- 

0.00 

- 

- 

- 

- 

- 

- 

- 

2.24E-09 

- 

- 

- 

- 

145 

1.7 

4.34-cm [1.71-in.] surge line break [Break flow area increased by 30% from 3.81-cm [1.5-in.] break]. Winter conditions assumed [HPI, LPI temp = 277 K [40° F] and CFT temp = 294 K [70° F]]. 

1.34E-04 

- 

- 

- 

- 

- 

- 

- 

- 

- 

- 

- 

- 

- 

- 

- 

- 

Palisades 

61 

2.8 7.18-cm (2.8-in.) cold leg break. Summer conditions assumed (HPI and LPI injection temp = 100 F, 

Accumulator temp = 90 F) 

2.09E-04 

0.01 

0.03 

0.11 

0.16 

0.00 

0.01 

0.04 

0.11 

2.98E-08 

1.89E-07 

3.87E-06 

3.06E-05 

3.57E-10 

3.32E-09 

3.71E-07 

6.55E-06 

60 

2 

5.08-cm (2-in.) surge line break. Winter conditions assumed (HPI and LPI injection temp = 40 F, Accumulator temp = 60 F) 

2.09E-04 

1.67 

1.94 

2.60 

2.56 

0.89 

1.69 

3.00 

3.51 

3.82E-06 

1.07E-05 

8.22E-05 

5.02E-04 

1.81E-07 

1.13E-06 

2.41E-05 

1.88E-04 

2 

1.4 3.59-cm (1.414-in.) surge line break. Containment sump recirculation included in the analysis. 

2.66E-04 

- 

- 

- 

0.00 

- 

- 

- 

0.00 

- 

- 

- 

1.41E-14 

- 

- 

- 

2.82E-15 

Note: There are no operator actions for any of these transients, and all transients initiate from full power 

conditions.
A-6 

Table A.3. Transient descriptions and FAVOR 04.1 results for stuck-open primary valve transients 

(including value reclosure) 

Percent 

Contributio

n to Total 

Frequency 

of Crack 

Initiation 

(FCI) 

Percent 

Contributio

n to Total 

Through 

Wall 

Cracking 

Frequency 

(TWCF) 

TH # Transients Including Valve Reclosure System Failure Operator Action 

HZP 

IEF 

32 

60 

Ext-A 

Ext-B 

32 

60 

Ext-A 

Ext-B 

32 

Beaver Valley Unit 1 

126 Reactor/turbine trip w/one stuck-open pressurizer SRV, which recloses at 6,000 s Operator controls HHSI 10 minutes after allowed. N 

1.87E-04 

0.2 

0.1 

0.0 

0.0 

23.6 

14.0 

3.1 

0.77 

8.24E-07 

60 Reactor/turbine trip w/one stuck-open pressurizer SRV, which recloses at 6,000 s. None. N 

2.15E-05 

0.1 

0.1 

0.0 

0.0 

16.8 

7.5 

1.8 

0.41 

4.69E-06 

130 Reactor/turbine trip w/one stuck-open pressurizer SRV, which recloses at 3,000 s at HZP Operator controls HHSI 10 minutes after allowed. Y 

3.09E-05 

0.1 

0.1 

0.0 

0.0 

16.5 

7.1 

1.0 

0.24 

4.14E-06 

97 Reactor/turbine trip w/one stuck-open pressurizer SRV, which recloses at 3,000 s. None. Y 

3.74E-06 

0.0 

0.0 

0.0 

0.0 

4.0 

2.0 

0.6 

0.13 

7.68E-06 

129 Reactor/turbine trip w/one stuck-open pressurizer SRV, which recloses at 6,000 s at HZP Operator controls HHSI 10 minutes after allowed. Y 

3.09E-05 

0.0 

0.0 

0.0 

0.0 

2.9 

1.2 

0.2 

0.03 

9.54E-07 

123 Reactor/turbine trip w/two stuck-open pressurizer SRVs, which reclose at 3,000 s at HZP Operator controls HHSI 10 minutes after allowed. Y 

1.65E-07 

0.0 

0.0 

0.0 

0.0 

1.2 

0.6 

0.3 

0.09 

6.32E-05 

71 Reactor/turbine trip w/one stuck-open pressurizer SRV, which recloses at 6,000 s. None. Y 

3.74E-06 

0.0 

0.0 

0.0 

0.0 

1.2 

0.5 

0.1 

0.02 

1.96E-06 
A-7 

Percent 

Contributio

n to Total 

Frequency 

of Crack 

Initiation 

(FCI) 

Percent 

Contributio

n to Total 

Through 

Wall 

Cracking 

Frequency 

(TWCF) 

TH # Transients Including Valve Reclosure System Failure Operator Action 

HZP 

IEF 

32 

60 

Ext-A 

Ext-B 

32 

60 

Ext-A 

Ext-B 

32 

61 Reactor/turbine trip w/two stuck-open pressurizer SRV, which recloses at 3,000 s. None. N 

1.79E-06 

0.0 

0.0 

0.0 

0.0 

0.2 

0.3 

0.4 

0.26 

5.67E-06 

69 Reactor/turbine trip w/two stuck-open pressurizer SRVs, which reclose at 3,000 s. None. Y 

2.09E-08 

0.0 

0.0 

0.0 

0.0 

0.3 

0.2 

0.1 

0.04 

1.32E-04 

120 Reactor/turbine trip w/two stuck-open pressurizer SRVs, which recloses at 6,000 s Operator controls HHSI 10 minutes after allowed. N 

9.98E-07 

0.0 

0.0 

0.0 

0.0 

0.5 

0.2 

0.0 

0.01 

3.62E-06 

124 Reactor/turbine trip w/two stuck-open pressurizer SRVs, which reclose at 6,000 s at HZP Operator controls HHSI 10 minutes after allowed. Y 

1.65E-07 

0.0 

0.0 

0.0 

0.0 

0.1 

0.0 

0.0 

0.01 

1.39E-05 

92 Reactor/turbine trip w/two stuck-open pressurizer SRVs, one recloses at 3000 s. None. Y 

2.13E-07 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.01 

5.15E-05 

93 

Reactor/turbine trip w/two stuck-open pressurizer SRVs. One valve recloses at 6000 seconds, while the other valve remains open. 

None. Y 

2.13E-07 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.01 

5.15E-05 

70 Reactor/turbine trip w/two stuck-open pressurizer SRVs, which reclose at 6,000 s. None. Y 

2.09E-08 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.00 

1.35E-05 

66 Reactor/turbine trip w/two stuck-open pressurizer SRVs. One valve recloses at 3000 seconds, while the other valve remains 

open.

None. N 

1.18E-06 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.01 

5.25E-06 

62 Reactor/turbine trip w/two stuck-open pressurizer SRV, which recloses at 6,000 s. None. N 

1.08E-07 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.00 

1.20E-06 
A-8 

Percent 

Contributio

n to Total 

Frequency 

of Crack 

Initiation 

(FCI) 

Percent 

Contributio

n to Total 

Through 

Wall 

Cracking 

Frequency 

(TWCF) 

TH # Transients Including Valve Reclosure System Failure Operator Action 

HZP 

IEF 

32 

60 

Ext-A 

Ext-B 

32 

60 

Ext-A 

Ext-B 

32 

59 Reactor/turbine trip w/one stuck-open pressurizer SRV, which recloses at 3,000 s. None. N 

3.46E-04 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.00 

0.00E+00 

67 

Reactor/turbine trip w/two stuck-open pressurizer SRVs. One valve recloses at 6000 seconds, while the other valve remains open. 

None. N 

1.18E-06 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.00 

2.07E-07 

119 Reactor/turbine trip w/two stuck-open pressurizer SRV, which recloses at 6,000 s Operator controls HHSI 1 minute after allowed. N 

6.84E-07 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.00 

5.60E-07 

121 Reactor/turbine trip w/two stuck-open pressurizer SRV, which recloses at 3,000 s at HZP Operator controls HHSI 1 minute after allowed Y 

1.33E-07 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.00 

9.79E-06 

122 Reactor/turbine trip w/two stuck-open pressurizer SRVs, which reclose at 6,000 s at HZP Operator controls HHSI 1 minute after allowed. Y 

1.33E-07 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.00 

9.79E-06 

125 Reactor/turbine trip w/one stuck-open pressurizer SRV, which recloses at 6,000 s Operator controls HHSI 1 minute after allowed. N 

1.34E-04 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.00 

2.39E-09 

127 Reactor/turbine trip w/one stuck-open pressurizer SRV, which recloses at 6,000 s at HZP Operator controls HHSI 1 minute after allowed. Y 

2.59E-05 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.00 

4.37E-18 

128 Reactor/turbine trip w/one stuck-open pressurizer SRV, which recloses at 3,000 s at HZP Operator controls HHSI 1 minute after allowed. Y 

2.59E-05 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.00 

4.37E-18 

Oconee Unit 1 
A-9 

Percent 

Contributio

n to Total 

Frequency 

of Crack 

Initiation 

(FCI) 

Percent 

Contributio

n to Total 

Through 

Wall 

Cracking 

Frequency 

(TWCF) 

TH # Transients Including Valve Reclosure System Failure Operator Action 

HZP 

IEF 

32 

60 

Ext-A 

Ext-B 

32 

60 

Ext-A 

Ext-B 

32 

122 Stuck-open pressurizer safety valve. Valve recloses at 6000 secs. 

Operator throttles HPI at 10 minutes after 2.7 K [5°F] subcooling and 254-cm [100”] pressurizer level is reached (throttling criteria is 27.8 K [50°F] subcooling). 

Y 

7.57E-06 

13.7 

4.7 

0.4 

0.3 

76.8 

74.5 

26.9 

11.27 

3.01E-06 

165 Stuck-open pressurizer safety valve. Valve recloses at 6000 secs [RCS low pressure point]. None Y 

1.76E-06 

4.0 

1.3 

0.1 

0.1 

22.4 

20.3 

7.8 

2.26 

2.75E-06 

124 Stuck-open pressurizer safety valve. Valve recloses at 3000 secs. 

Operator throttles HPI at 10 minutes after 2.7 K [5°F] subcooling and 254-cm [100”] pressurizer level is reached (throttling criteria is 27.8 K [50°F] subcooling). 

Y 

7.57E-06 

0.1 

0.2 

0.2 

0.2 

0.5 

3.0 

12.8 

9.02 

3.61E-08 

168 TT/RT with stuck-open pzr SRV. SRV assumed to reclose at 3000 secs. None Y 

1.76E-06 

0.0 

0.0 

0.0 

0.0 

0.1 

0.4 

2.4 

1.33 

5.48E-08 

113 Stuck-open pressurizer safety valve. Valve recloses at 6000 secs. 

After valve recloses, operator throttles HPI 10 minutes after 2.7 K [5°F] subcooling and 254-cm [100”] pressurizer level is reached (throttling criteria is 27.8 K [50°F] subcooling) 

N 

5.07E-05 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.30 

0.00E+00 

109 Stuck-open pressurizer safety valve. Valve recloses at 6000 secs [RCS low pressure point]. None N 

9.58E-06 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.00 

0.00E+00 

112 Stuck-open pressurizer safety valve. Valve recloses at 6000 secs. 

After valve recloses, operator throttles HPI 1 minute after 2.7 K [5°F] subcooling and 254-cm [100”] pressurizer level is reached (throttling criteria is 27 K [50°F] subcooling) 

N 

1.25E-04 

- 

- 

- 

- 

- 

- 

- 

- 

- 

114 Stuck-open pressurizer safety valve. Valve recloses at 3000 secs. 

After valve recloses, operator throttles HPI 1 minute after 2.7 K [5°F] subcooling and 254-cm [100”] pressurizer level is reached (throttling criteria is 50°F subcooling) 

N 

1.25E-04 

- 

- 

- 

- 

- 

- 

- 

- 

- 
A-10

Percent 

Contributio

n to Total 

Frequency 

of Crack 

Initiation 

(FCI) 

Percent 

Contributio

n to Total 

Through 

Wall 

Cracking 

Frequency 

(TWCF) 

TH # Transients Including Valve Reclosure System Failure Operator Action 

HZP 

IEF 

32 

60 

Ext-A 

Ext-B 

32 

60 

Ext-A 

Ext-B 

32 

115 Stuck-open pressurizer Safety Valve. Valve recloses at 3000 secs. 

After valve recloses, operator throttles HPI 10 minutes after 2.7 K [5°F] subcooling and 254-cm [100”] pressurizer level is reached (throttling criteria is 50°F subcooling) 

N 

5.07E-05 

- 

- 

- 

- 

- 

- 

- 

- 

- 

121 Stuck-open pressurizer safety valve. Valve recloses at 6000 secs . 

Operator throttles HPI at 1 minute after 2.7 K [5°F] subcooling and 254-cm [100”] pressurizer level is reached [throttling criteria is 27.8 K [50°F] subcooling]. 

Y 

2.28E-05 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.00 

0.00E+00 

123 Stuck-open pressurizer safety valve. Valve recloses at 3000 secs. 

Operator throttles HPI at 1 minute after 2.7 K [5°F] subcooling and 254-cm [100”] pressurizer level is reached (throttling criteria is 27.8 K [50°F] subcooling). 

Y 

2.28E-05 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.00 

0.00E+00 

149 TT/RT with stuck-open pzr SRV. SRV assumed to reclose at 3000 secs. None N 

9.58E-06 

- 

- 

- 

- 

- 

- 

- 

- 

- 

Palisades 

65 One stuck-open pressurizer SRV that recloses at 6000 sec after initiation. Containment spray is assumed not to actuate. None Y 

1.24E-04 

6.5 

5.8 

4.4 

2.4 

67.2 

45.4 

17.5 

8.40 

2.60E-05 

48 Two stuck-open pressurizer SRVs that reclose at 6000 sec after initiation. Containment spray is assumed not to actuate. None Y 

7.67E-07 

0.1 

0.1 

0.1 

0.0 

1.4 

0.9 

0.3 

0.12 

8.57E-05 

53 

Turbine/reactor trip with two stuck-open pressurizer SRVs that reclose at 6000 sec after initiation. Containment spray is assumed not to actuate. 

None N 

1.09E-03 

0.0 

0.0 

0.3 

0.4 

0.0 

0.2 

0.8 

1.27 

9.91E-10 
A-11

Percent 

Contributio

n to Total 

Frequency 

of Crack 

Initiation 

(FCI) 

Percent 

Contributio

n to Total 

Through 

Wall 

Cracking 

Frequency 

(TWCF) 

TH # Transients Including Valve Reclosure System Failure Operator Action 

HZP 

IEF 

32 

60 

Ext-A 

Ext-B 

32 

60 

Ext-A 

Ext-B 

32 

42 Turbine/reactor trip with two stuck-open pressurizer SRVs. Containment spray is assumed not to actuate. 

Operator assumed to throttle HPI if auxiliary feedwater is running with SG wide range level > -84% and RCS subcooling > 25 F. HPI is throttled to maintain pressurizer level between 40 and 60%. 

N 

7.67E-07 

- 

- 

- 

- 

- 

- 

- 

- 

- 
A-12

Table A.4. Transient descriptions and FAVOR 04.1 results for stuck-open primary valve transients (no value reclosure) 

Percent 

Contributio

n to Total 

Frequency 

of Crack 

Initiation 

(FCI) 

Percent 

Contributio

n to Total 

Through 

Wall 

Cracking 

Frequency 

(TWCF) 

TH # Transients Without Valve Reclosure System Failure Operator Action 

HZP 

IEF 

32 

60 

Ext-A 

Ext-B 

32 

60 

Ext-A 

Ext-B 

32 

Beaver Valley Unit 1 

14 Reactor/turbine trip w/one stuck-open pressurizer SRV None. N 

2.23E-04 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.00 

1.56E-11 

34 Reactor/turbine trip w/two stuck-open pressurizer SRVs None. N 

4.95E-07 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.00 

2.39E-07 

64 Reactor/turbine trip w/two stuck-open pressurizer SRVs None. Y 

8.67E-08 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.00 

7.51E-06 

94 Reactor/turbine trip w/one stuck-open pressurizer SRV. None. Y 

4.10E-05 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.00 

0.00E+00 

Oconee Unit 1 

169 

TT/RT with stuck-open pzr SRV [valve flow area reduced by 30%]. Summer conditions assumed [HPI, LPI temp = 302 K [85° F] and CFT temp = 310 K [100° F]]. Vent valves do not function. 

None Y 

7.33E-06 

0.0 

0.0 

0.1 

0.2 

0.0 

0.0 

0.6 

0.95 

1.90E-10 

146 

TT/RT with stuck-open pzr SRV [valve flow area reduced by 30%]. Summer conditions assumed [HPI, LPI temp = 302 K [85° F] and CFT temp = 310 K [100° F]]. Vent valves do not function. 

None N 

4.23E-05 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.17 

0.00E+00 

147 TT/RT with stuck-open pzr SRV. Summer conditions assumed [HPI, LPI temp = 302 K [85° F] and CFT temp = 310 K [100° 

F]]. 

None N 

3.63E-05 

- 

- 

- 

- 

- 

- 

- 

- 

- 
A-13

Percent 

Contributio

n to Total 

Frequency 

of Crack 

Initiation 

(FCI) 

Percent 

Contributio

n to Total 

Through 

Wall 

Cracking 

Frequency 

(TWCF) 

TH # Transients Without Valve Reclosure System Failure Operator Action 

HZP 

IEF 

32 

60 

Ext-A 

Ext-B 

32 

60 

Ext-A 

Ext-B 

32 

148 TT/RT with partially stuck-open pzr SRV [flow area equivalent to 1.5 in diameter opening]. HTC coefficients increased by 

1.3. 

None N 

4.23E-05 

- 

- 

- 

- 

- 

- 

- 

- 

- 

170 TT/RT with stuck-open pzr SRV. Summer conditions assumed [HPI, LPI temp = 302 K [85° F] and CFT temp = 310 K [100° 

F]]. 

None Y 

6.28E-06 

- 

- 

0.0 

0.0 

- 

- 

- 

0.00 

- 

171 TT/RT with partially stuck-open pzr SRV [flow area equivalent to 1.5 in diameter opening]. HTC coefficients increased by 

1.3. 

None Y 

7.33E-06 

- 

- 

- 

- 

- 

- 

- 

- 

- 
A-14

Table A.5. Transient descriptions and FAVOR 04.1 results for large-diameter steam line break transients 

Percent 

Contributio

n to Total 

Frequency 

of Crack 

Initiation 

(FCI) 

Percent 

Contributio

n to Total 

Through 

Wall 

Cracking 

Frequency 

(TWCF) 

TH Transient # System Failure Operator Action 

HZP 

IEF 

32 

60 

Ext-A 

Ext-B 

32 

60 

Ext-A 

Ext-B 

32 

Beaver Valley Unit 1 

103 Main steam line break with AFW continuing to feed affected generator for 30 minutes. 

Operator controls HHSI 30 minutes after allowed. Break is assumed to occur inside containment so that the operator trips the RCPs due to adverse containment conditions. 

Y 

1.1E-05 

0.07 

0.17 

0.30 

0.63 

0.74 

2.54 

4.78 

5.17 

736E06 

104 Main steam line break with AFW continuing to feed affected generator for 30 minutes. 

Operator controls HHSI 60 minutes after allowed. Break is assumed to occur inside containment so that the operator trips the RCPs due to adverse containment conditions. 

N 

1.1E-04 

0.07 

0.25 

0.36 

1.72 

0.01 

0.55 

2.67 

10.24 

363E07 

102 Main steam line break with AFW continuing to feed affected generator for 30 minutes. 

Operator controls HHSI 30 minutes after allowed. Break is assumed to occur inside containment so that the operator trips the RCPs due to adverse containment conditions. 

N 

1.0E-04 

0.03 

0.11 

0.36 

1.62 

0.00 

0.32 

3.05 

363E07 

107 Main steam line break with AFW continuing to feed affected generator. 

Operator controls HHSI 30 minutes after allowed. Break is assumed to occur inside containment so that the operator trips the RCPs due to adverse containment conditions. 

Y 

4.3E-07 

0.00 

0.00 

0.01 

0.02 

0.03 

0.12 

0.18 

0.22 

616E06 

105 Main steam line break with AFW continuing to feed affected generator for 30 minutes. 

Operator controls HHSI 60 minutes after allowed. Break is assumed to occur inside containment so that the operator trips the RCPs due to adverse containment conditions. 

Y 

1.1E-05 

0 

0 

0.01 

0.02 

0 

0.01 

0.04 

0.14 

862E09 

106 Main steam line break with AFW continuing to feed affected generator. 

Operator controls HHSI 30 minutes after allowed. Break is assumed to occur inside containment so that the operator trips the RCPs due to adverse containment conditions. 

N 

2.2E-06 

0 

0 

0.01 

0.04 

0 

0 

0.08 

0.21 

352E07 

74 Main steam line break with AFW continuing to feed affected generator None. N 

1.5E-06 

0 

0 

0 

0.01 

0 

0 

0 

0.02 

146E08 
A-15

Percent 

Contributio

n to Total 

Frequency 

of Crack 

Initiation 

(FCI) 

Percent 

Contributio

n to Total 

Through 

Wall 

Cracking 

Frequency 

(TWCF) 

TH Transient # System Failure Operator Action 

HZP 

IEF 

32 

60 

Ext-A 

Ext-B 

32 

60 

Ext-A 

Ext-B 

32 

81 

Main Steam Line Break with AFW continuing to feed affected generator and with HHSI failure initially. 

Operator opens ADVs (on intact generators). HHSI is restored after CFTs discharge 50%. N 

2.7E-06 

0 

0 

0 

0 

0 

0 

0 

0 

000E00 

27 MSLB without trip of turbine-driven emergency feedwater. Operator throttles HPI to maintain 27.8 K [50° F] subcooling margin. N 

2.1E-06 

0 

0 

0 

0 

0 

0 

0 

0.01 

000E00 

99 MSLB with trip of turbine-driven EFW by MSLB Circuitry 

HPI is throttled 20 minutes after 2.7 K [5°F] subcooling and 254-cm [100”] pressurizer level is reached (throttling criteria is 27.8 K [50°F] subcooling). 

N 

2.4E-07 

0 

0 

0 

0 

0 

0 

0 

0 

000E00 

100 MSLB with trip of turbine-driven EFW by MSLB Circuitry 

Operator throttles HPI 20 minutes after 2.7 K [5°F] subcooling and 254-cm [100”] pressurizer level is reached (throttling criteria is 27.8 K [50°F] subcooling). 

Y 

5.1E-08 

0 

0 

0 

0 

0 

0 

0 

0 

000E00 

101 MSLB without trip of turbine-driven EFW by MSLB Circuitry Operator throttles HPI to maintain 27.8 K [50° F] subcooling margin (throttling criteria is 27.8 K [50°F] subcooling). Y 

3.9E-07 

0 

0 

0 

0 

0 

0 

0 

0 

000E00 

54 

Main steam line break with failure of both MSIVs to close. Break assumed to be inside containment causing containment spray actuation. 

Operator does not isolate AFW on affected SG. Operator does not throttle HPI. N 

4.3E-06 

0.44 

0.51 

0.66 

0.69 

1.66 

1.88 

1.62 

1.51 

516E05 

27 

Main steam line break with controller failure resulting in the flow from two AFW pumps into affected steam generator. Break assumed to be inside containment causing containment spray actuation. 

Operator starts second AFW pump. N 

3.7E-05 

0.13 

0.2 

0.37 

0.41 

0.26 

0.41 

0.72 

0.92 

156E06 

Oconee Unit 1 

Palisades 
A-16

26 

Main steam line break with the break assumed to be inside containment causing containment spray actuation. 

Operator isolates AFW to affected SG at 30 minutes after initiation. N 

5.7E-04 

0.04 

0.11 

0.41 

0.7 

0.02 

0.05 

0.29 

0.98 

498E08 

50 

Main steam line break with controller failure resulting in the flow from two AFW pumps into affected steam generator. Break assumed to be inside containment causing containment spray actuation. 

Operator starts second AFW pump. Operator does not throttle HPI. Y 

5.8E-07 

0.01 

0.01 

0.01 

0.01 

0.02 

0.02 

0.03 

0.03 

445E06 

51 

Main steam line break with failure of both MSIVs to close. Break assumed to be inside containment causing containment spray actuation. 

Operator does not isolate AFW on affected SG. Operator does not throttle HPI. Y 

7.5E-08 

0 

0 

0.01 

0.01 

0.02 

0.02 

0.02 

0.02 

255E05 

34 Main steam line break concurrent with a single tube failure in SG-A due to MSLB vibration. 

Operator isolates AFW to affected SG at 15 minutes after initiation. Operator trips RCPs assuming that they do not trip as a result of the event. Operator assumed to throttle HPI if auxiliary feedwater is running with SG wide range level > -84% and RCS 

N 

1.5E-05 

0 

0.01 

0.03 

0.04 

0 

0.01 

0.03 

0.06 

143E07 

49 

Main steam line break with the break assumed to be inside containment causing containment spray actuation. 

Operator isolates AFW to affected SG at 30 minutes after initiation. Operator does not throttle HPI. Y 

1.0E-05 

0 

0 

0 

0.01 

0 

0 

0 

0.01 

902E09 

24 

Main steam line break with the break assumed to be inside containment causing containment spray actuation. 

None N 

2.4E-06 

0 

0 

0 

0 

0 

0 

0 

0 

498E08 

29 

Main steam line break with break assumed to be inside containment causing containment spray actuation. 

None. Operator does not throttle HPI. Y 

4.2E-08 

0 

0 

0 

0 

0 

0 

0 

0 

202E07 
A-17

Table A.6. Transient descriptions and FAVOR 04.1 results for SO-2 transients involving all 

(or a very large number of) stuck-open valves 

Percent 

Contributio

n to Total 

Frequency 

of Crack 

Initiation 

(FCI) 

Percent 

Contributio

n to Total 

Through 

Wall 

Cracking 

Frequency 

(TWCF) 

TH Transient # System Failure Operator Action 

HZP 

IEF 

32 

60 

Ext-A 

Ext-B 

32 

60 

Ext-A 

Ext-B 

32 

Beaver Valley Unit 1 

110 Small steam line break (simulated by sticking open all SG-A SRVs) with AFW continuing to 

feed affected generator for 30 minutes 

Operator controls HHSI 60 minutes after allowed. N 

6.9E-04 

0 

0 

0.02 

0.32 

0 

0 

0.02 

1.43 

175E11 

108 Small steam line break (simulated by sticking open all SG-A SRVs) with AFW continuing to 

feed affected generator for 30 minutes. 

Operator controls HHSI 30 minutes after allowed. Y 

6.5E-04 

0 

0.01 

0.01 

0.2 

0 

0 

0.02 

0.64 

175E11 

111 Small steam line break (simulated by sticking open all SG-A SRVs) with AFW continuing to 

feed affected generator for 30 minutes. 

Operator controls HHSI 60 minutes after allowed. Break is assumed to occur inside containment so that the operator trips the RCPs due to adverse containment conditions. 

Y 

6.8E-05 

0 

0.01 

0.02 

0.11 

0 

0 

0.07 

0.51 

308E09 

109 Small steam line break (simulated by sticking open all SG-A SRVs) with AFW continuing to 

feed affected generator for 30 minutes. 

Operator controls HHSI 30 minutes after allowed. Break is assumed to occur inside containment so that the operator trips the RCPs due to adverse containment conditions. 

Y 

6.8E-05 

0 

0 

0.02 

0.12 

0 

0 

0.03 

0.43 

308E09 

112 Small steam line break (simulated by sticking open all SG-A SRVs) with AFW continuing to 

feed affected generator. 

Operator controls HHSI 30 minutes after allowed. Break is assumed to occur inside containment so that the operator trips the RCPs due to adverse containment conditions. 

N 

1.4E-05 

0 

0 

0 

0.01 

0 

0 

0 

0.02 

186E11 

118 Small steam line break (simulated by sticking open all SG-A SRVs) with AFW continuing to 

feed affected generator 

None. N 

9.3E-06 

0 

0 

0 

0.01 

0 

0 

0 

0.02 

314E10 
A-18

Percent 

Contributio

n to Total 

Frequency 

of Crack 

Initiation 

(FCI) 

Percent 

Contributio

n to Total 

Through 

Wall 

Cracking 

Frequency 

(TWCF) 

TH Transient # System Failure Operator Action 

HZP 

IEF 

32 

60 

Ext-A 

Ext-B 

32 

60 

Ext-A 

Ext-B 

32 

113 Small steam line break (simulated by sticking open all SG-A SRVs) with AFW continuing to 

feed affected generator. 

Operator controls HHSI 30 minutes after allowed. Break is assumed to occur inside containment so that the operator trips the RCPs due to adverse containment conditions. 

Y 

2.7E-06 

0 

0 

0 

0 

0 

0 

0 

0.01 

407E10 

78 Reactor/turbine trip with failure of MFW and AFW. Operator opens all ASDVs to let condensate fill SGs. N 

3.3E-8 

0 

0 

0 

0 

0 

0 

0 

0.00 

0 

89 Reactor/turbine trip with loss of MFW and EFW. 

Operator opens all TBVs to depressurize the secondary side to below the condensate booster pump shutoff head so that these pumps feed the steam generators. Booster pumps are assumed to be initially uncontrolled so that the steam generators are overfilled 

N 

5.4E-07 

0 

0 

0 

0 

0 

0 

0 

0 

000E00 

98 Reactor/turbine trip with loss of MFW and EFW 

Operator opens all TBVs to depressurize the secondary side to below the condensate booster pump shutoff head so that these pumps feed the steam generators. Booster pumps are assumed to be initially uncontrolled so that the steam generators are overfilled 

Y 

1.0E-07 

0 

0 

0 

0 

0 

0 

0 

0 

000E00 

22 Turbine/reactor trip with loss of MFW and AFW. 

Operator depressurizes through ADVs and feeds SGs using condensate booster pumps. Operators maintain a cooldown rate within technical specification limits and throttle condensate flow at 84% level in the steam generator. 

N 

6.7E-05 

0 

0 

0 

0 

0 

0 

0 

0.01 

000E00 

Oconee Unit 1 

Palisades 
A-19

Table A.7. Transient descriptions and FAVOR 04.1 results for SO-2 transients involving just a few 

(one or two) stuck-open secondary valves 

Percent 

Contribution to 

Total Frequency of 

Crack Initiation 

(FCI) 

Percent 

Contribution to 

Total Through Wall 

Cracking 

Frequency (TWCF) 

Mean CPI 

TH Transient # System Failure Operator Action 

HZP 

IEF 

32 

60 

Ext-A 

Ext-B 

32 

60 

Ext-A 

Ext-B 

32 

60 

Ext-A 

Beaver Valley Unit 1 

No transients of this type were analyzed 

Oconee Unit 1 

28 Reactor/turbine trip with 1 stuck- open safety valve in SG-A. None N 

7.5E-08 

0 

0 

0 

0 

0 

0 

0 

0 

0 

0 

0 

29 

Reactor/turbine trip with 1 stuckopen safety valve in SG-A and a second stuck-open safety valve in SG-B. 

None N 

3.1E-07 

0 

0 

0 

0 

0 

0 

0 

0 

0 

0 

0 

30 Reactor/turbine trip with 1 stuck- open safety valve in SG-A. None Y 

1.5E-07 

0 

0 

0 

0 

0 

0 

0 

0 

0 

0 

0 

31 

Reactor/turbine trip with 1 stuckopen safety valve in SG-A and a second stuck-open safety valve in SG-B. 

None Y 

8.4E-09 

0 

0 

0 

0 

0 

0 

0 

0 

0 

0 

0 

36 

Reactor/turbine trip with 1 stuckopen safety valve in SG-A and a second stuck-open safety valve in SG-B. 

Operator throttles HPI to maintain 27.8 K [50° F] subcooling and 304.8-cm [120-in.] pressurizer level. 

N 

1.4E-05 

0 

0 

0 

0 

0 

0 

0 

0 

0 

0 

0 

37 Reactor/turbine trip with 1 stuck- open safety valve in SG-A. 

Operator throttles HPI to maintain 27.8 K [50° F] subcooling and 304.8-cm [120-in.] pressurizer level. 

Y 

1.4E-06 

0 

0 

0 

0 

0 

0 

0 

0 

0 

0 

0 

38 

Reactor/turbine trip with 1 stuckopen safety valve in SG-A and a second stuck-open safety valve in SG-B. 

Operator throttles HPI to maintain 27.8 K [50° F] subcooling and 304.8-cm [120-in.] pressurizer level. 

Y 

2.7E-06 

0 

0 

0 

0 

0 

0 

0 

0 

0 

0 

0 
A-20

Percent 

Contribution to 

Total Frequency of 

Crack Initiation 

(FCI) 

Percent 

Contribution to 

Total Through Wall 

Cracking 

Frequency (TWCF) 

Mean CPI 

TH Transient # System Failure Operator Action 

HZP 

IEF 

32 

60 

Ext-A 

Ext-B 

32 

60 

Ext-A 

Ext-B 

32 

60 

Ext-A 

90 Reactor/turbine trip with 2 stuck- open safety valves in SG-A. 

Operator throttles HPI 20 minutes after 2.7 K [5°F] subcooling and 254-cm [100”] pressurizer level is reached [throttling criteria is 27.8 K [50°F] subcooling]. 

N 

6.3E-07 

0 

0 

0 

0 

0 

0 

0 

0 

0 

0 

0 

102 Reactor/turbine trip with 2 stuck- open safety valves in SG-A. 

Operator throttles HPI 20 minutes after 2.77 K [5°F] subcooling and 254-cm [100-in.] pressurizer level is reached (throttling criteria is 27 K [50°F] subcooling). 

Y 

2.0E-07 

0 

0 

0 

0 

0 

0 

0 

0 

0 

0 

0 

Palisades 

P55 

Turbine/reactor trip with 2 stuckopen ADVs on SG-A combined with controller failure resulting in the flow from two AFW pumps into affected steam generator. 

Operator starts second AFW pump. N 

2.7E-03 

0.6 

1.7 

5.0 

5.4 

3.0 

8.4 

16.8 

18.0 

3.5E-08 

3.0E-07 

5.9E-06 

19 Reactor trip with 1 stuck-open ADV on SG-A. None. Operator does not throttle HPI. Y 

2.3E-03 

0.2 

0.8 

2.5 

3.1 

0.9 

3.2 

7.4 

9.7 

9.1E-09 

1.1E-07 

2.4E-06 

52 

Reactor trip with 1 stuck-open ADV on SG-A. Failure of both MSIVs (SG-A and SG-B) to close. 

Operator does not isolate AFW on affected SG. Normal AFW flow assumed (200 gpm). Operator does not throttle HPI. 

Y 

6.4E-04 

0 

0.0 

0.1 

0.4 

0.0 

0.1 

0.4 

1.2 

1.7E-08 

1.5E-07 

3.0E-06 

16 

Turbine/reactor trip with 2 stuckopen ADVs on SG-A combined with controller failure resulting in the flow from two AFW pumps into affected steam generator. 

Operator starts second AFW pump. Operator isolates AFW to affected SG at 30 minutes after initiation. Operator assumed to throttle HPI if auxiliary feedwater is running with SG wide range level > -84% and RCS subcooling > 25 F. HPI is throttled to main 

N 

1.2E-04 

0 

0 

0 

0 

0 

0 

0 

0 

0 

9.4E-12 

2.3E-08 
A-21

Percent 

Contribution to 

Total Frequency of 

Crack Initiation 

(FCI) 

Percent 

Contribution to 

Total Through Wall 

Cracking 

Frequency (TWCF) 

Mean CPI 

TH Transient # System Failure Operator Action 

HZP 

IEF 

32 

60 

Ext-A 

Ext-B 

32 

60 

Ext-A 

Ext-B 

32 

60 

Ext-A 

18 

Turbine/reactor trip with 1 stuckopen ADV on SG-A. Failure of both MSIVs (SG-A and SG-B) to close. 

Operator does not isolate AFW on affected SG. Normal AFW flow assumed (200 gpm). Operator assumed to throttle HPI if auxiliary feedwater is running with SG wide range level > -84% and RCS subcooling > 25 F. HPI is throttled to maintain pressurizer level 

N 

4.7E-03 

0 

0 

0 

0 

0 

0 

0 

0 

0 

2.1E-12 

3.0E-11 
A-22 

Table A.8. Transient descriptions and FAVOR 04.1 results for feed and bleed, overfeed, 

and steam generator tube rupture transients 

Percent 

Contributio

n to Total 

Frequency 

of Crack 

Initiation 

(FCI) 

Percent 

Contributio

n to Total 

Through 

Wall 

Cracking 

Frequency 

(TWCF) 

Plant 

Class 

TH # System Failure Operator Action 

HZP 

IEF 

32 

60 

Ext-A 

Ext-B 

32 

60 

Ext-A 

Ext-B 

32 

31 

Turbine/reactor trip with failure of MFW and AFW. Containment spray actuation assumed due to PORV discharge. 

Operator maintains core cooling by “feed and bleed” using HPI to feed and two PORVs to bleed. 

N 

1.29E-05 

0.05 

0.05 

0.07 

0.07 

0.06 

0.08 

0.10 

0.09 

209E06 

Palisades 32 Turbine/reactor trip with failure of MFW and AFW. Containment spray actuation 

assumed due to PORV discharge. 

Operator maintains core cooling by “feed and bleed” using HPI to feed and two PORV to bleed. AFW is recovered 15 minutes after initiation of “feed and bleed” cooling. Operator closes PORVs when SG level reaches 60%. 

N 

1.08E-06 

0.00 

0.00 

0.00 

0.00 

0.00 

0.00 

0.00 

0.00 

387E09 

F&B 

31 Reactor/turbine trip w/ feed & bleed Operator opens all pzr PORVs & uses all charging/HHSI pumps N 

3.10E-7 

0.00 

0.00 

0.00 

0.00 

0.00 

0.00 

0.00 

0.00 

344E7 

Beaver 

Overfeed 76 Reactor/turbine trip w/full MFW to all 3 SGs (MFW maintains SG level near top). Operator trips reactor coolant pumps. Y 

1.1E-04 

0 

0 

0.00 

0.00 

0 

0 

0.00 

0.00 

0 

Oconee 

SGTR 127 

SGTR with a stuck-open SRV in SG-B. A reactor trip is assumed to occur at the time of the tube rupture. Stuck safety relief valve is assumed to reclose 10 minutes after initiation. 

Operator trips RCPs 1 minute after initiation. Operator also throttles HPI 10 minutes after 2.77 K [5° F] subcooling and 254-cm [100-in.] pressurizer level is reached (assumed throttling criteria is 27 K [50°F] subcooling). 

Y 

1.25E-07 

- 

- 

- 

- 

- 

- 

- 

- 
A-23 

Table A.9. Transient descriptions and FAVOR 04.1 results for mixed primary and secondary initiator transients 

Percent 

Contributio

n to Total 

Frequency 

of Crack 

Initiation 

(FCI) 

Percent 

Contributio

n to Total 

Through 

Wall 

Cracking 

Frequency 

(TWCF) 

M

# System Failure Operator Action 

HZP 

IEF 

32 

60 

Ext-A 

Ext-B 

32 

60 

Ext-A 

Ext-B 

32 

Beaver Valley Unit 1 

83 

2.54-cm [1.0-in.] surge line break with HHSI failure and motor driven AFW failure. MFW is tripped. Level control failure causes all steam generators to be overfed with turbine AFW, with the level maintained at top of SGs. 

Operator trips RCPs. Operator opens all ASDVs 5 minutes after HHSI would have come on. 

N 

3.51E-06 

- 

- 

- 

- 

- 

- 

- 

- 

- 

B-65 Reactor/turbine trip w/two stuck-open pressurizer SRVs and HHSI failure Operator opens all ASDVs 5 minutes after HHSI would have come on. N 

1.04E-09 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.00 

1.71E-09 

68 Reactor/turbine trip w/two stuck-open pressurizer SRVs that reclose at 6000 s with HHSI failure. Operator opens all ASDVs 5 minutes after HHSI would have come on. N 

1.33E-08 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.00 

1.41E-07 

72 Reactor/turbine trip w/one stuck-open pressurizer SRV with HHSI failure. Operator opens all ASDVs 5 minutes after HHSI would have come on. N 

5.14E-07 

- 

- 

- 

- 

- 

- 

- 

- 

- 

73 Reactor/turbine trip w/one stuck-open pressurizer SRV with HHSI failure Operator open all ASDVs 5 minutes after HHSI would have come on. Y 

6.55E-08 

- 

- 

- 

- 

- 

- 

- 

- 

- 

82 Reactor/turbine trip w/one stuck-open pressurizer SRV (recloses at 6000 s) and with HHSI failure. Operator opens all ASDVs 5 minutes after HHSI would have started. N 

1.51E-06 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.00 

0.00E+00 
A-24 

Percent 

Contributio

n to Total 

Frequency 

of Crack 

Initiation 

(FCI) 

Percent 

Contributio

n to Total 

Through 

Wall 

Cracking 

Frequency 

(TWCF) 

M

# System Failure Operator Action 

HZP 

IEF 

32 

60 

Ext-A 

Ext-B 

32 

60 

Ext-A 

Ext-B 

32 

Oconee Unit 1 

O-110 5.08-cm [2 in.] surge line break with HPI failure At 15 minutes after transient initiation, operator opens both TBV to lower primary system pressure and allow CFT and LPI injection. N 

3.42E-06 

0.18 

0.46 

1.18 

1.48 

0.00 

0.00 

0.64 

1.16 

7.19E-08 

120 2.54-cm [1-in.] surge line break with HPI Failure 

At 15 minutes after sequence initiation, operators open all TBVs to depressurize the system to the CFT setpoint. When the CFTs are 50% discharged, HPI is assumed to be recovered. The TBVs are assumed remain opened for the duration of the transient 

Y 

4.22E-08 

0.00 

0.00 

0.00 

0.00 

0.00 

0.00 

0.01 

0.01 

1.69E-12 

44 2.54-cm [1-in.] surge line break with HPI Failure 

At 15 minutes after initiation, operators open all TBVs to depressurize the system to the CFT setpoint. When the CFTs are 50% discharged, HPI is assumed to be recovered. The TBVs are assumed remain open for the duration of the transient. 

N 

2.69E-07 

0.00 

0.00 

0.00 

0.00 

0.00 

- 

- 

0.01 

0.00E+00 

119 2.54-cm [1-in.] surge line break with HPI Failure 

At 15 minutes after transient initiation, the operator opens all turbine bypass valves to lower primary system pressure and allow core flood tank and LPI injection. 

Y 

4.41E-07 

0.00 

0.00 

0.00 

0.01 

0.00 

- 

0.00 

0.01 

2.83E-10 

8 2.54-cm [1-in.] surge line break with 1 stuck-open safety valve in SG-A. None N 

9.68E-08 

- 

- 

- 

- 

- 

- 

- 

- 

- 

12 2.54-cm [1-in.] surge line break with 1 stuck-open safety valve in SG-A. HPI throttled to maintain 27.8 K [50° F] subcooling margin N 

9.24E-07 

- 

- 

- 

- 

- 

- 

- 

- 

- 
A-25 

15 2.54-cm [1-in.] surge line break with HPI Failure 

At 15 minutes after transient initiation, operator opens all TBVs to lower primary system pressure and allow CFT and LPI injection. 

N 

3.39E-08 

- 

- 

0.00 

0.00 

- 

- 

- 

0.00 

- 

111 2.54-cm [1-in.] surge line break with HPI failure 

At 15 minutes after initiation, operator opens all TBVs to lower primary pressure and allow CFT and LPI injection. When the CFTs are 50% discharged, HPI is recovered. At 3000 seconds after initiation, operator starts throttling HPI to 55 K [100°F] subcooling 

N 

4.16E-07 

- 

- 

0.00 

0.00 

- 

- 

- 

0.00 

- 

117 Stuck-open pressurizer safety valve and HPI failure 

At 15 minutes after initiation, operator opens all TBV to lower primary pressure and allow CFT and LPI injection. When the CFTs are 50% discharged, HPI is recovered. The SRV is closed 5 minutes after HPI recovered. HPI is throttled at 1 minute after 2. 

N 

5.38E-07 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.01 

2.12E-11 

116 Stuck-open pressurizer safety valve and HPI failure 

At 15 minutes after initiation, operator opens all TBVs to lower primary pressure and allow CFT and LPI injection. When the CFTs are 50% discharged, HPI is recovered. The HPI is throttled 20 minutes after 2.7 K [5°F] subcooling and 254-cm [100”] pressure 

N 

2.60E-07 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.00 

0.00E+00 

125 Stuck-open pressurizer safety valve and HPI Failure 

At 15 minutes after initiation, operator opens all TBVs to lower primary pressure and allow CFT and LPI injection. When the CFTs are 50% discharged, HPI is recovered. HPI is throttled 20 minutes after 2.7 K [5°F] subcooling and 254-cm [100”] pressurize 

Y 

4.61E-08 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.00 

1.44E-10 

126 Stuck-open pressurizer safety valve and HPI Failure 

At 15 minutes after initiation, operator opens all TBVs to lower primary pressure and allow CFT and LPI injection. When the CFTs are 50% discharged, HPI is recovered. SRV is closed at 5 minutes after HPI is recovered. HPI is throttled at 1 minute after 

Y 

8.41E-08 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.0 

0.00 

0.00E+00