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Materials Reliability Program:

Safety Evaluation for Boric Acid Wastage ofPWR Reactor Vessel Bottom Heads Due to

Bottom-Mounted Nozzle Leakage (MRP-167)Evaluations Supporting the PWR Bottom-Mounted Nozzle Inspection Plan

Effective October 9, 2012, this report has been made publicly available in accordance

with Section 734.3(b)(3) and published in accordance with Section 734.7 of the U.S.

Export Administration Regulations. As a result of this publication, this report is subject

to only copyright protection and does not require any license agreement from EPRI.

This notice supersedes the export control restrictions and any proprietary licensed material

notices embedded in the document prior to publication.

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Materials Reliability Program:Safety Evaluation for Boric Acid

Wastage of PWR Reactor VesselBottom Heads Due to Bottom-MountedNozzle Leakage (MRP-167)Evaluations Supporting the PWR Bottom-MountedNozzle Inspection Plan

1016591

Final Report, July 2008

EPRI Project ManagersC. HarringtonC. King

ELECTRIC POWER RESEARCH INSTITUTE3420 Hillview Avenue, Palo Alto, California 94304-1338 PO Box 10412, Palo Alto, California 94303-0813 USA

800.313.3774 650.855.2121 [email protected] www.epri.com

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DISCLAIMER OF WARRANTIES AND LIMITATION OF LIABILITIES

THIS DOCUMENT WAS PREPARED BY THE ORGANIZATION(S) NAMED BELOW AS ANACCOUNT OF WORK SPONSORED OR COSPONSORED BY THE ELECTRIC POWER RESEARCHINSTITUTE, INC. (EPRI). NEITHER EPRI, ANY MEMBER OF EPRI, ANY COSPONSOR, THEORGANIZATION(S) BELOW, NOR ANY PERSON ACTING ON BEHALF OF ANY OF THEM:

(A) MAKES ANY WARRANTY OR REPRESENTATION WHATSOEVER, EXPRESS OR IMPLIED, (I)WITH RESPECT TO THE USE OF ANY INFORMATION, APPARATUS, METHOD, PROCESS, ORSIMILAR ITEM DISCLOSED IN THIS DOCUMENT, INCLUDING MERCHANTABILITY AND FITNESSFOR A PARTICULAR PURPOSE, OR (II) THAT SUCH USE DOES NOT INFRINGE ON ORINTERFERE WITH PRIVATELY OWNED RIGHTS, INCLUDING ANY PARTY'S INTELLECTUALPROPERTY, OR (III) THAT THIS DOCUMENT IS SUITABLE TO ANY PARTICULAR USER'SCIRCUMSTANCE; OR

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ORGANIZATION(S) THAT PREPARED THIS DOCUMENT

Dominion Engineering, Inc.

NOTICE: THIS REPORT CONTAINS PROPRIETARY INFORMATION THAT IS THEINTELLECTUAL PROPERTY OF MRP UTILITY MEMBERS AND EPRI.ACCORDINGLY, IT IS AVAILABLE ONLY UNDER LICENSE FROM EPRI ANDMAY NOT BE REPRODUCED OR DISCLOSED, WHOLLY OR IN PART, BY

ANY LICENSEE TO ANY OTHER PERSON OR ORGANIZATION.

NOTE

For further information about EPRI, call the EPRI Customer Assistance Center at 800.313.3774 ore-mail [email protected].

Electric Power Research Institute, EPRI, and TOGETHERSHAPING THE FUTURE OF ELECTRICITYare registered service marks of the Electric Power Research Institute, Inc.

Copyright 2008 Electric Power Research Institute, Inc. All rights reserved.
mailto:[email protected]:[email protected]

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CITATIONS

This report was prepared by

Dominion Engineering, Inc.11730 Plaza America DriveSuite 310Reston, VA 20190

Principal Investigator

G. White

ContributorsS. AhnertD. ArguellesJ. BroussardM. FlemingV. MoroneyN. Nordmann

This report describes research sponsored by the Electric Power Research Institute (EPRI).

The report is a corporate document that should be cited in the literature in the following manner:

Materials Reliability Program: Safety Evaluation for Boric Acid Wastage of PWR ReactorVessel Bottom Heads Due to Bottom-Mounted Nozzle Leakage (MRP-167): EvaluationsSupporting the PWR Bottom-Mounted Nozzle Inspection Plan. EPRI, Palo Alto, CA: 2008.1016591.

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REPORT SUMMARY

This report is one of the principal inputs to the Materials Reliability Program (MRP) bottom-mounted nozzle inspection plan (MRP-206; EPRI 1015003). This program is expected to defineinspection requirements appropriate for the long-term management of the potential for primarywater stress corrosion cracking (PWSCC) of Alloy 600 nozzles and Alloy 82/182 J-groove weldsof the penetrations mounted on the reactor vessel bottom head of pressurized water reactors(PWRs).

BackgroundIn 2003, two leaking bottom-mounted nozzles (BMNs) located on the reactor vessel bottom headwere detected at one PWR unit. Leakage was the result of primary water stress corrosion cracksin the Alloy 600 nozzle material and the Alloy 182 J-groove attachment weld. Although nosignificant wastage resulted from this leakage, boric acid corrosion of the low-alloy steel shellmaterial of the bottom head due to BMN leakage is one of two principal potential safetyconcerns associated with aging degradation of BMNs. The other potential safety concern isejection of a BMN due to circumferential PWSCC in the nozzle tube material below theelevation of the bottom of the J-groove attachment weld. This second concern has been evaluatedby the Westinghouse and B&W Owners Groups.

ObjectivesTo determine the effect on nuclear safety of the potential for boric acid corrosion of the low-alloy steel material of the reactor vessel bottom head due to potential BMN leakage, given thebenefit of a program of periodic visual and/or volumetric examinations.

ApproachThe project team based the safety evaluations in this study on plant experience with leaking J-groove penetrations in combination with deterministic and probabilistic modeling of the wastageprocess. The approach taken is similar to the approach documented in Section 7 and AppendicesD and E of report MRP-110 to evaluate the potential for wastage of reactor vessel closure headsgiven potential leakage of control rod drive mechanisms (CRDMs) or other closure headpenetrations. The model of the bottom head wastage process includes the main physical

processes that are judged to control the potential for a loss-of-coolant accident (LOCA): crackinitiation, growth of a part-depth crack, leak initiation, growth of a through-wall crack, increasein leak rate and extent of local cooling, increase in resulting wastage rate, increase in area ofunsupported vessel cladding material, and ultimate rupture of unsupported cladding. The modelevaluates the benefit of periodic volumetric examinations in detecting cracking and periodic baremetal or supplemental visual examinations in detecting leakage before a LOCA can occur due towastage. Additional calculations are used to verify tolerance of the head shell to significantmaterial volume loss.

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Wastage models in this report were designed as tools for investigating the benefit of variousprograms of periodic inspections on the likelihood of significant low-alloy steel wastageoccurring on the basis of an assumed linear wastage rate versus leak rate relationship. Thisrelationship was developed on the basis of extensive local cooling that is required for a highlyconcentrated aqueous solution or hydrated slurry to exist over a relatively large (> 2 in

2) surface

of the low-alloy steel shell material. Such conditions are believed to be necessary for thepotential to exist for rapid corrosion over a structurally significant region. These wastage modelsare not intended to reflect physical details of the several complex corrosion mechanisms thatpotentially may be active. Sufficient experimental data are not currently available to supportdevelopment of such explicit models. Statistical inputs to the probabilistic evaluation presentedhere were designed to capture process uncertainties to the extent possible. However, to refineunderstanding of the wastage process, an extensive MRP laboratory boric acid corrosion testprogram is currently ongoing. This program is now beginning to yield results that will allowmore refined models of the individual wastage mechanisms to be developed and the mainassumptions of the models presented in this report to be checked. As additional results aregenerated under the MRP test program and evaluated with regard to the BMN location, it ispossible that the conclusions of this report could be affected.

ResultsPlant experience coupled with deterministic and probabilistic calculations demonstrate that aprogram of periodic visual examinations of the outer surface of the RV bottom head providessufficient assurance against the potential nuclear safety concern of structurally significantwastage of low-alloy steel RV head material. (Visual inspections may be augmented by periodicvolumetric inspections to lessen the visual inspection burden.) Probabilistic risk calculationsbased on a set of input assumptions representing all 58 domestic PWR plants that have BMNsshow acceptable results for the bottom head rupture frequency for several different inspectionprogram options. These results include the program of bare metal visual (BMV) examinationsperformed every other refueling outage that is specified in American Society of Mechanical

Engineers (ASME) Code Case N-722 for the BMN location. Acceptable inspection programs aredefined separately for plants with 18- and 24-month fuel cycles. For example, for 18-monthplants, a program of BMV examinations performed every third refueling outage is concluded tobe acceptable provided that less sensitive supplemental visual examinations are performedduring the other refueling outages.

EPRI PerspectiveAlong with potential for nozzle ejection, boric acid wastage of the low-alloy steel RV head shellresulting from BMN penetration leakage and boric acid concentration is a principal potentialsafety concern associated with PWSCC. This document does not impose any inspectionrequirements or commitments, but rather is an input to the development of appropriate long-termBMN inspection practices.

KeywordsAlloy 600Boric acid corrosionBottom/Incore mounted instrumentation (BMI/IMI) nozzleBottom-mounted nozzle (BMN)Low-alloy steel wastagePrimary water stress corrosion cracking (PWSCC)

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EPRI Proprietary Licensed Material

ABSTRACT

This safety assessment addresses one of the potential safety issues associated with agingdegradation of reactor vessel bottom head penetrations: Bottom-Mounted nozzles (BMNs).Specifically, this report evaluates the concern that BMN leakage due to primary water stresscorrosion cracking (PWSCC) of the Alloy 600 nozzle and/or Alloy 82/182 J-groove attachmentweld could lead to significant wastage of the low-alloy steel head shell material due toconcentration of the boric acid present in the leaking primary coolant. Evaluation of the safetysignificance of potential wastage degradation is based on a review of relevant plant experience incombination with deterministic and probabilistic modeling of the wastage process. The

probabilistic model assumes that wastage progression continues to rupture unless cracking isdetected via a volumetric examination or leakage is detected via a bare metal or supplementalvisual examination or via an increase in the magnitude of the unidentified leak rate to 1 gpm ormore.

Wastage models presented in this report were designed as tools for investigating the benefit ofvarious programs of periodic inspections on the likelihood of significant low-alloy steel wastageoccurring based on an assumed linear wastage rate versus leak rate relationship. This relationshipwas developed on the basis of extensive local cooling that is required for a highly concentratedaqueous solution or hydrated slurry to exist over a relatively large (> 2 in

2) surface of the low-

alloy steel shell material. Such conditions are believed to be necessary for the potential to existfor rapid corrosion over a structurally significant region. The wastage models are not intended to

reflect physical details of the several complex corrosion mechanisms that potentially may beactive. Sufficient experimental data are currently unavailable to support development of suchexplicit models. To refine understanding of the wastage process, an extensive MaterialsReliability Program (MRP) laboratory boric acid corrosion test program is currently ongoing. Asadditional results are generated under the MRP test program and evaluated with regard to theBMN location, it is possible that the conclusions of this report could be affected.

Plant experience coupled with the deterministic and probabilistic calculations in this reportdemonstrate that a program of periodic visual examinations of the outer surface of the RVbottom head provides sufficient assurance against the potential nuclear safety concern ofstructurally significant wastage of the low-alloy steel head material. (Visual inspections may be

augmented by periodic volumetric inspections to lessen the visual inspection burden.)Probabilistic risk calculations based on a set of input assumptions representing all 58 domesticpressurized water reactor (PWR) plants that have BMNs show acceptable results for the bottomhead rupture frequency for several different inspection program options, including the programof bare metal visual (BMV) examinations performed every other refueling outage that isspecified in American Society of Mechanical Engineers (ASME) Code Case N-722 for the BMNlocation. Acceptable inspection programs are defined separately for plants with 18- and 24-month fuel cycles. For example, for 18-month plants, a program of BMV examinations

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performed every third refueling outage is concluded to be acceptable provided that less sensitivesupplemental visual examinations are performed during the other refueling outages. Thisdocument does not impose any inspection requirements or commitments, but rather is an input tothe development of appropriate long-term BMN inspection practices.

This report is organized as follows, where the appendices provide additional details on thesupporting calculations summarized in Section 4:

Section 1 provides a brief background discussion, describes the basic BMN designconfiguration, defines the scope of this safety evaluation, and outlines the approach used.

Section 2 summarizes the results of this study investigating the benefit of candidate programsof BMN inspection including bare metal visual, supplemental visual, and ultrasonicexamination for reducing the risk associated with the potential for bottom head wastage dueto BMN leakage. Basic inspection programs are identified that result in an acceptably loweffect on nuclear safety.

Section 3 discusses the limited plant experience with leaking BMNs and the more extensive

experience with leakage from other J-groove Alloy 600 penetrations. The experience withleaking Alloy 600 pressurizer heater sleeves is of particular relevance to the inverted BMNgeometry. Section 3 also cites the body of laboratory test data for boric acid corrosion.

Section 4 presents the methodologies and key results of calculations performed to supportmodeling of the wastage process: stress analyses, stress intensity factor fracture mechanicscalculations, crack opening displacement calculations, leak rate calculations, thermalmodeling of leakage, calculations of allowable wastage volume without ASME Codeallowable stress levels being exceeded in the head shell, and a brief review of the claddingburst-strength study performed by Oak Ridge National Laboratory (ORNL) under NuclearRegulatory Commission (NRC) funding.

Section 5 describes the deterministic evaluation of the wastage process that specificallyexamines the benefit of bare metal visual examinations. The deterministic evaluation is basedon the time that is required for the leak rate to increase from the rate that is readily detectablevia bare metal visual examination to the leak rate that is sufficiently high to produce theextensive local cooling that is necessary for rapid wastage to proceed over a relatively largesurface area (> 2 in

2) of the low-alloy steel head material (within the nozzle crevice and/or on

the outer head surface).

Section 6 presents the methodology and results of the probabilistic model of the bottom headwastage process. The model addresses the main physical processes that are judged to controlthe potential for a loss-of-coolant accident (LOCA), and it evaluates the benefit of periodicbare metal or supplemental visual examinations in detecting the leakage before a LOCA can

occur due to wastage. The model also was used to evaluate the benefit of ultrasonicexaminations of the BMN tube in detecting cracking before a through-wall crack and leakageare produced, or after leakage is produced but not yet detected.

Section 7 lists the references cited in the body and appendices of this report.

Appendix A provides background regarding the finite element analysis (FEA) penetrationstress model that was used to support development of several wastage model inputs, such ascrack-tip stress intensity factor computations and crack opening displacement calculations.

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Appendix B presents the methodology for computing the crack-tip stress intensity factor of athrough-wall axial nozzle crack. This crack geometry is assumed in the wastage models toproduce the leakage necessary for boric acid corrosion of the low-alloy steel head to occur.

Appendix C presents the crack opening displacement (and crack opening area) calculationsthat are a key input to the leak rate calculations. Finite element modeling was used to

evaluate crack opening displacements for postulated through-wall axial nozzle cracks ofvarying lengths.

Appendix D summarizes leak rate calculations for postulated through-wall axial nozzlecracks that were performed using two different existing methods: the EPRI PICEP computercode and a methodology developed by Laborelec for steam generator tube leaks. Along withthe empirical plant experience for leaking J-groove penetrations, these leak rate calculationswere applied in the selection of the leak rate curve input to the bottom head wastage models.

Appendix E presents the key aspects of the FEA thermal model used to assess the coolingeffects of a BMN leak on the adjacent region of the bottom head (annulus and outer surface).The extent of cooling is a key parameter controlling the extent of expected wastage.

Appendix F presents calculations used to verify the tolerance of the low-alloy steel shell tosignificant material volume loss. In the probabilistic wastage model, rupture of anunsupported area of the stainless steel head cladding is assumed to be the ultimate failuremechanism rather than rupture of the low-alloy steel shell material.

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ACKNOWLEDGMENTS

This report is a product of the Assessment Issue Task Group (ITG) of the Materials ReliabilityProgram (MRP). Mel Arey of Duke Energy is the leader of the BMN Core Team within theAssessment ITG. Craig Harrington is the EPRI Project Manager for the BMN Core Team, andChristine King is the EPRI Program Manager for the MRP.

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ACRONYMS

The following acronyms are used in this report:

B&W Babcock & Wilcox

BAC boric acid corrosion

BMI (reactor vessel) bottom-mounted instrumentation (Westinghouse design)

BMN bottom-mounted nozzle (i.e., BMI/IMI nozzle)

BMV bare metal visual

CCDP conditional core damage probability

CDF core damage frequency

CE Combustion Engineering

CEDM control element drive mechanism

CGR crack growth rate

COA crack opening area

COD crack opening displacement

CRDM control rod drive mechanism

FEA finite element analysis

ICI incore instrumentation (CE design)

IMI incore monitoring instrumentation (B&W design)

LWR light water reactor

MRP Materials Reliability Program

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NDE non-destructive examination

NSSS nuclear steam supply system

PWR pressurized water reactor

PWSCC primary water stress corrosion cracking

RCS reactor coolant system

RTD resistance temperature detector

SCC stress corrosion cracking

SV supplemental visual

UT ultrasonic testing

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CONTENTS

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

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

1.2 Basic BMN Configuration ................................................................................................1-1

1.3 Purpose...........................................................................................................................1-2

1.4 Scope ..............................................................................................................................1-2

1.5 Approach.........................................................................................................................1-21.6 MRP Efforts to Refine Understanding of the Wastage Process ......................................1-3

1.7 Report Structure..............................................................................................................1-4

2SUMMARY AND CONCLUSIONS .........................................................................................2-1

2.1 Summary.........................................................................................................................2-1

2.1.1 Review of Relevant Plant Experience .....................................................................2-1

2.1.2 Deterministic Wastage Model..................................................................................2-4

2.1.3 Probabilistic Wastage Model ...................................................................................2-6

2.2 Conclusions.....................................................................................................................2-7

3RELEVANT PLANT EXPERIENCE AND LABORATORY TESTING....................................3-1

3.1 Introduction .....................................................................................................................3-1

3.2 Plant Experience .............................................................................................................3-1

3.2.1 South Texas Project Unit 1 BMNs (2003)................................................................3-2

3.2.2 Arkansas Nuclear One Unit 2 Pressurizer Heater Sleeve (1987)............................3-3

3.2.3 Calvert Cliffs Unit 2 Pressurizer Heater Sleeve H3 (1989) ......................................3-4

3.2.4 RCS Piping Instrumentation Nozzles and Pressurizer Heater Sleeves...................3-53.3 Laboratory TestingABB-CE Tests Using Inverted Penetration Geometry ...................3-6

3.4 Conclusions.....................................................................................................................3-7

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4SUPPORTING CALCULATIONS...........................................................................................4-1

4.1 Introduction .....................................................................................................................4-1

4.2 Stress Analyses...............................................................................................................4-1

4.2.1 Finite Element Model...............................................................................................4-1

4.2.2 Cases/Configurations Considered...........................................................................4-2

4.3 Stress Intensity Factor Calculations Based on Finite Element Analysis .........................4-3

4.4 Crack Opening Displacement Calculations Based on Finite Element Analysis ..............4-4

4.5 Leak Rate Calculations ...................................................................................................4-5

4.5.1 Introduction..............................................................................................................4-5

4.5.2 PICEP Code Analyses.............................................................................................4-5

4.5.3 Calculations Using Laborelec Approach for Steam Generator Tubes.....................4-6

4.5.4 Assumed Bottom Head Wastage Model Curve .......................................................4-64.6 Thermal Modeling of Leaks.............................................................................................4-7

4.7 Allowable Wastage Volume Calculations ........................................................................4-7

4.8 Review of ORNL Cladding Burst-Strength Evaluation Program......................................4-8

4.9 Modeling of Part-Depth Cracks for Purpose of Considering Benefit of VolumetricExaminations.........................................................................................................................4-9

5DETERMINISTIC WASTAGE EVALUATION ........................................................................5-1

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

5.2 Volume of Boric Acid Deposits Detectable by Bare Metal Visual Examination ...............5-2

5.3 Volume of Boric Acid Deposits versus Leak Rate ...........................................................5-2

5.4 Leak Rate to Produce Rapid Corrosion...........................................................................5-3

5.5 Ongoing MRP Boric Acid Corrosion Test Program .........................................................5-4

5.6 Deterministic Evaluation..................................................................................................5-7

5.7 Conclusions.....................................................................................................................5-8

6PROBABILISTIC WASTAGE MODELING ............................................................................6-1

6.1 Introduction .....................................................................................................................6-1

6.2 Assumed Wastage Progression......................................................................................6-2

6.3 Description of the Probabilistic Wastage Model ..............................................................6-2

6.3.1 Probabilistic Wastage Model Basics........................................................................6-3

6.3.2 Statistical Distributions Used for Inputs ...................................................................6-7

6.3.3 Additional Inputs to Investigate Benefit of Supplemental Visual andVolumetric Examinations ................................................................................................6-12

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6.3.4 Monte Carlo Calculations ......................................................................................6-14

6.4 Base and Sensitivity Cases Investigated ......................................................................6-15

6.5 Results of the Probabilistic Wastage Model ..................................................................6-16

6.5.1 Benefit of BMV Exam Strategy ..............................................................................6-17

6.5.2 Benefit of SV Exams..............................................................................................6-17

6.5.3 Benefit of Volumetric Exams .................................................................................6-17

6.5.4 Distribution Parameter Value Sensitivity Cases ....................................................6-18

6.5.5 Distribution Type Sensitivity Cases .......................................................................6-20

6.5.6 Sensitivity of Case 13 (BMV Every Other Outage for 24-month Cycle) andCase 14 (BMV Every Other Outage plus Volumetric Examination Every 10 Yearsfor 24-month Cycle) to 38% Reduction in the Wastage Rate Multiplicative FactorParameter.......................................................................................................................6-22

6.6 Conclusions of the Probabilistic Wastage Model ..........................................................6-23

6.7 Refinement of Modeling Assumptions and Inputs .........................................................6-25

7 REFERENCES .......................................................................................................................7-1

ABOTTOM HEAD PENETRATION FEA STRESS MODEL ...................................................A-1

A.1 Introduction .................................................................................................................... A-1

A.2 Finite Element Model ..................................................................................................... A-1

BCRACK-TIP STRESS INTENSITY FACTOR COMPUTATIONS ......................................... B-1

B.1 Introduction .................................................................................................................... B-1B.2 Analysis Overview.......................................................................................................... B-1

B.3 Assumptions .................................................................................................................. B-3

B.4 Designs Analyzed .......................................................................................................... B-4

CCRACK OPENING DISPLACEMENT CALCULATIONS.....................................................C-1

C.1 Introduction.................................................................................................................... C-1

C.2 Crack Opening Displacement Analysis..........................................................................C-1

C.3 Crack Opening Area (COA) and Crack Width (CW) ......................................................C-2

C.4 Assumptions .................................................................................................................. C-3

DLEAK RATE CALCULATIONS ............................................................................................D-1

D.1 Introduction.................................................................................................................... D-1

D.2 PICEP Code Analysis.................................................................................................... D-1

D.3 Analysis Based on Laborelec Methodology................................................................... D-3

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ETHERMAL MODELING OF LEAKS ..................................................................................... E-1

E.1 Introduction .................................................................................................................... E-1

E.2 Methodology .................................................................................................................. E-1

E.3 Comparison to Corresponding Reactor Vessel Closure Head Calculation.................... E-2

FALLOWABLE WASTAGE VOLUME CALCULATIONS........................................................F-1

F.1 Allowable Low-Alloy Steel Wastage Volume...................................................................F-1

F.1.1 Allowable Corrosion Volume Based on Finite Element Analysis.............................F-1

F.1.2 Allowable Corrosion Volume for Other Head Designs ............................................F-3

F.2 Summary.........................................................................................................................F-3

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LIST OF FIGURES

Figure 1-1 Location of Bottom-Mounted Nozzles (BMNs) on Bottom Head of PWRReactor Vessel and Typical Basic BMN Configuration ......................................................1-7

Figure 1-2 BMN Configuration for B&W Design Plants..............................................................1-8

Figure 3-1 STP 1 Bottom-Mounted Nozzle Leakage Nozzle 1 ...............................................3-9

Figure 3-2 STP 1 Bottom-Mounted Nozzle Leakage Nozzle 46 .............................................3-9

Figure 3-3 1989 Pressurizer Heater Sleeve Experience at Calvert Cliffs Unit 2: Core

Sample H3 Side Views [10]..............................................................................................3-10Figure 3-4 1989 Pressurizer Heater Sleeve Experience at Calvert Cliffs Unit 2: Core

Sample H3 Bottom Views [10] .........................................................................................3-11

Figure 3-5 1989 Pressurizer Heater Sleeve Experience at Calvert Cliffs Unit 2: SleeveH3 Scraped Areas [10].....................................................................................................3-12

Figure 3-6 Pressurizer Bottom Head Leakage at Calvert Cliffs 2 in 1989 [27] ........................3-13

Figure 3-7 Schematic of CE Leaking Annulus Corrosion Test.................................................3-14

Figure 4-1 Nozzle Hoop Stress Drop-off with Distance Below the Bottom of the Weld forSelect Cases ....................................................................................................................4-11

Figure 4-2 FEA Model for Stress intensity Factor Fracture Mechanics Calculations...............4-12

Figure 4-3 Crack-Tip Stress Intensity Factor Curve (for Westinghouse 3-Loop Plant)Used as a Base Case Input to the Deterministic and Probabilistic Bottom HeadWastage Models ..............................................................................................................4-13

Figure 4-4 Leak Rate versus Axial Crack Length Comparison of Models/Results andthe Base Case Curve Assumed in the Deterministic and Probabilistic Bottom HeadWastage Models ..............................................................................................................4-14

Figure 4-5 Results of ORNL Cladding Burst-Strength Study: Failure Pressure versusCrack Depth (Normalized to Clad Thickness) (reproduced from Reference [16])............4-15

Figure 4-6 Assessment of Wastage Length Producing an Area of Unsupported CladdingPotentially Resulting in Cladding Rupture Based on ORNL Burst Study .........................4-16

Figure 4-7 Crack Growth Calculation for Axial Inside Surface Flaws for the BoundingWestinghouse and CE Plant [5] .......................................................................................4-17

Figure 4-8 Crack Growth Calculation for Axial Outside Surface Flaws for the BoundingWestinghouse and CE Plant [5] .......................................................................................4-17

Figure 4-9 Example Plant with Relatively Thick BMNs and a Tcold of 545F: CrackGrowth Calculation for Axial ID Flaws in Nozzles with a: a) 6 Angle, b) 27 Angle,and c) 44 Angle...............................................................................................................4-18

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Figure 4-10 Example Plant with Relatively Thick BMNs and a Tcold of 545F: CrackGrowth Calculation for Axial OD Flaws on the Downhill Side in Nozzles with a: a)6Angle, b) 27Angle, and c) 44Angle .........................................................................4-19

Figure 4-11 Example Plant with Relatively Thick BMNs and a Tcold of 545F: CrackGrowth Calculation for Axial OD Flaws on the Uphill Side in Nozzles with a: a) 6Angle, b) 27Angle, and c) 44 Angle..............................................................................4-20

Figure 4-12 Example Plant with Relatively Thick BMNs and a Tcold of 545F: CrackGrowth Calculation for Axial OD Flaws on the Downhill Side in Nozzles with a: a)6Angle, b) 27Angle, and c) 44Angle .........................................................................4-21

Figure 4-13 Example Plant with Relatively Thick BMNs and a Tcold of 545F: CrackGrowth Calculation for Axial OD Flaws on the Uphill Side in Nozzles with a: a) 6Angle, b) 27Angle, and c) 44 Angle..............................................................................4-22

Figure 5-1 Volume of Boric Acid Deposits as a Function of Leak Rate ...................................5-10

Figure 5-2 Time for Leak Rate to Increase from 110-6gpm to the Critical Leak Rate that

May Lead to Rapid Corrosion (estimated to be 0.05 gpm)...............................................5-11

Figure 5-3 Volume of Boric Acid Deposits and Bottom Head Corrosion Length

Progression as a Function of Time after Leakage Begins for the Deterministic BMNModel Using a) the 50

thPercentile of the CGR Distribution Reported in MRP-55; b)

the 75thPercentile of the CGR Distribution Reported in MRP-55 (i.e., the

deterministic MRP-55 equation).......................................................................................5-12

Figure 6-1 Simplified Flow Chart for the Probabilistic Wastage Model for BMN Leakage .......6-33

Figure 6-2 Probability of Cracking Weibull Plot Based on Alloy 600 Small-Bore NozzleExperience for CE Design Plants and Reactor Vessel Closure Head Vent Nozzlesin Westinghouse Design Plants .......................................................................................6-34

Figure 6-3 Assumed Dependence of Linear Wastage Rate on Leak Rate Based onAvailable Data: a) Log Scale for Leak Rate; b) Linear Scale for Leak Rate.....................6-35

Figure 6-4 Assumed Probability of Detection (POD) Curves for Detection of Leakage Via

Bare Metal Visual (BMV) Examination for Presence of Boric Acid Deposits: a) BaseCase POD Curve; b) Conservatively Low POD Curve (Case 22s) ..................................6-36

Figure 6-5 Assumed Probability of Detection (POD) Curve for Detection of Cracks ViaUltrasonic Examination ....................................................................................................6-37

Figure 6-6 Weibull Characteristic Time Distributions ...............................................................6-38

Figure 6-7 Weibull Slope Distributions.....................................................................................6-39

Figure 6-8 Log-Triangular Fit on a Heat Basis to Laboratory Crack Growth Rate DataPresented in MRP-55.......................................................................................................6-40

Figure 6-9 Distributions for Leak Rate for Crack Extending 1.3" Below Bottom of Weld .........6-40

Figure 6-10 Distributions for Upper-Shelf Wastage Rate for Leak Rates Greater than

LRcrit ..................................................................................................................................6-41

Figure 6-11 Time-Averaged (a) and Calendar-Year Peak (b) Head Rupture Frequenciesfor the Fictive Plant Bottom Head: Base Cases for Inspection Programs withPeriodic BMV Examinations.............................................................................................6-42

Figure 6-12 Rupture Frequency for the Fictive Plant Bottom Head: Base Cases forInspection Programs with Periodic BMV Examinations on a 1.5-Year Interval................6-43

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Figure 6-15 Rupture Frequency for the Fictive Plant Bottom Head: Base Cases for

Inspection Programs with Periodic BMV Examinations on a 4.0-Year Interval................6-44 Figure 6-16 Rupture Frequency for the Fictive Plant Bottom Head: Base Cases for

Inspection Programs with Periodic BMV Examinations on a 4.5-Year Interval................6-45

Figure 6-17 Leakage Frequency for the Fictive Plant Bottom Head: Base Cases forInspection Programs with Periodic BMV Examinations....................................................6-45

Figure 6-18 Comparison of Time-Averaged and Calendar-Year Peak RuptureFrequencies for the Fictive Plant Bottom Head: Effect of SV Sensitivity for BMVEvery Third Outage for 18-month Cycle (Base Case 22).................................................6-46

Figure 6-19 Rupture Frequency for the Fictive Plant Bottom Head: Effect of SVSensitivity for BMV Every Third Outage for 18-month Cycle (Base Case 22)..................6-46

Figure 6-20 Leakage Frequency for the Fictive Plant Bottom Head: Effect of SVSensitivity for BMV Every Third Outage for 18-month Cycle (Base Case 22)..................6-47

Figure 6-21 Comparison of Time-Averaged and Calendar-Year Peak RuptureFrequencies for the Fictive Plant Bottom Head: Effect of SV Sensitivity for BMVEvery Other Outage for 24-month Cycle (Base Case 16) ................................................6-47

Figure 6-22 Rupture Frequency for the Fictive Plant Bottom Head: Effect of SVSensitivity for BMV Every Other Outage for 24-month Cycle (Base Case 16).................6-48

Figure 6-23 Leakage Frequency for the Fictive Plant Bottom Head: Effect of SVSensitivity for BMV Every Other Outage for 24-month Cycle (Base Case 16).................6-48

Figure 6-24 Time-Averaged (a) and Calendar-Year Peak (b) Head Rupture Frequenciesfor the Fictive Plant Bottom Head: Cases for Inspection Programs without Future

Periodic Bare Metal Visual Examinations for 18-month Cycle .........................................6-49Figure 6-25 Rupture Frequency for the Fictive Plant Bottom Head: Cases for Inspection

Programs without Future Periodic Bare Metal Visual Examinations for 18-monthCycle (Cases 25, 26, 27, 31, 32, 32b, and 33) .................................................................6-50

Figure 6-26 Leakage Frequency for the Fictive Plant Bottom Head: Cases for InspectionPrograms without Future Periodic Bare Metal Visual Examinations for 18-monthCycle (Cases 25, 26, 27, 31, 32, 32b, and 33) .................................................................6-50

Figure 6-27 Time-Averaged (a) and calendar-Year Peak (b) Head Rupture Frequenciesfor the Fictive Plant Bottom Head: Cases for Inspection Programs without FuturePeriodic Bare Metal Visual Examinations for 24-month Cycle .........................................6-51

Figure 6-28 Rupture Frequency for the Fictive Plant Bottom Head: Cases for Inspection

Programs without Future Periodic Bare Metal Visual Examinations for 24-monthCycle (Cases 28, 29, 30, 34, 35, 35b, and 36) .................................................................6-52

Figure 6-29 Leakage Frequency for the Fictive Plant Bottom Head: Cases for InspectionPrograms without Future Periodic Bare Metal Visual Examinations for 24-monthCycle (Cases 28, 29, 30, 34, 35, 35b, and 36) .................................................................6-52

Figure 6-30 Time-Averaged (a) and Calendar-Year Peak (b) Head Rupture Frequenciesfor the Fictive Plant Bottom Head: Sensitivity of Results to Volumetric ExamCoverage Assumption......................................................................................................6-53

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Figure 6-31 Rupture Frequency for the Fictive Plant Bottom Head: Sensitivity of Resultsto Volumetric Exam Coverage Assumption (Cases 5, 5b, 11 and 11b) ...........................6-54

Figure 6-32 Rupture Frequency for the Fictive Plant Bottom Head: Sensitivity of Resultsto Volumetric Exam Coverage Assumption (Cases 17, 17b, 23 and 23b) .......................6-54

Figure 6-33 Leakage Frequency for the Fictive Plant Bottom Head: Sensitivity of Results

to Volumetric Exam Coverage Assumption (Cases 1, 2, 2b, and 3) ................................6-55Figure 6-34 Leakage Frequency for the Fictive Plant Bottom Head: Sensitivity of Results

to Volumetric Exam Coverage Assumption (Cases 5, 5b, 11 and 11b) ...........................6-55

Figure 6-35 Leakage Frequency for the Fictive Plant Bottom Head: Sensitivity of Resultsto Volumetric Exam Coverage Assumption (Cases 17, 17b, 23, and 23b) ......................6-56

Figure 6-36 Time-Averaged (a) and Calendar-Year Peak (b) Head Rupture Frequenciesfor the Fictive Plant Bottom Head: Distribution Parameter Value Sensitivity CasesBased on Case 22 (BMV Every Third Outage with SV during Other Outages for 18-month Cycle) ....................................................................................................................6-57

Figure 6-37 Rupture Frequency for the Fictive Plant Bottom Head: DistributionParameter Value Sensitivity Cases Based on Case 22 (BMV Every Third Outage

with SV during Other Outages for 18-month Cycle) (Cases 22, 22j, 22k, and 22l)..........6-58 Figure 6-38 Rupture Frequency for the Fictive Plant Bottom Head: Distribution

Parameter Value Sensitivity Cases Based on Case 22 (BMV Every Third Outagewith SV during Other Outages for 18-month Cycle) (Cases 22, 22m, 22n, and 22o) ......6-58

Figure 6-39 Rupture Frequency for the Fictive Plant Bottom Head: DistributionParameter Value Sensitivity Cases Based on Case 22 (BMV Every Third Outagewith SV during Other Outages for 18-month Cycle) (Cases 22, 22p, 22q, and 22r) ........6-59

Figure 6-40 Rupture Frequency for the Fictive Plant Bottom Head: DistributionParameter Value Sensitivity Cases Based on Case 22 (BMV Every Third Outagewith SV during Other Outages for 18-month Cycle) (Cases 22, 22s, and 22t).................6-59

Figure 6-41 Leakage Frequency for the Fictive Plant Bottom Head: Distribution

Parameter Value Sensitivity Cases Based on Case 22 (BMV Every Third Outagewith SV during Other Outages for 18-month Cycle) (Cases 22 and 22j through 22t) ......6-60

Figure 6-42 Time-Averaged (a) and Calendar-Year Peak (b) Head Rupture Frequenciesfor the Fictive Plant Bottom Head: Distribution Type Sensitivity Cases Based onCase 22 (BMV Every Third Outage with SV during Other Outages for 18-monthCycle) ...............................................................................................................................6-61

Figure 6-43 Rupture Frequency for the Fictive Plant Bottom Head: Distribution TypeSensitivity Cases Based on Case 22 (BMV Every Third Outage with SV duringOther Outages for 18-month Cycle) (Cases 22 and 22u through 22y).............................6-62

Figure 6-44 Leakage Frequency for the Fictive Plant Bottom Head: Distribution TypeSensitivity Cases Based on Case 22 (BMV Every Third Outage with SV during

Other Outages for 18-month Cycle) (Cases 22 and 22u through 22y).............................6-62

Figure 6-45 Time-Averaged (a) and Calendar-Year Peak (b) Head Rupture Frequenciesfor the Fictive Plant Bottom Head: Sensitivity of Case 13 (BMV Every Other Outagefor 24-month Cycle) and Case 14 (BMV Every Other Outage plus VolumetricExamination Every 10 Years for 24-month Cycle) to 38% Reduction in the WastageRate Multiplicative Factor Parameter ...............................................................................6-63

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Figure 6-46 Rupture Frequency for the Fictive Plant Bottom Head: Sensitivity of Case13 (BMV Every Other Outage for 24-month Cycle) and Case 14 (BMV Every OtherOutage plus Volumetric Examination Every 10 Years for 24-month Cycle) to 38%Reduction in the Wastage Rate Multiplicative Factor Parameter .....................................6-64

Figure 6-47 Leakage Frequency for the Fictive Plant Bottom Head: Sensitivity of Case

13 (BMV Every Other Outage for 24-month Cycle) and Case 14 (BMV Every OtherOutage plus Volumetric Examination Every 10 Years for 24-month Cycle) to 38%Reduction in the Wastage Rate Multiplicative Factor Parameter .....................................6-64

Figure A-1 Bottom-Head Nozzle FEA Models ......................................................................... A-4

Figure B-1 Fracture Mechanics FEA Model............................................................................. B-5

Figure B-2 Fracture Mechanics ModelCrack Mesh Detail.................................................... B-6

Figure B-3 Crack-Tip Stress Intensity Factor CE System 80 .............................................. B-7

Figure C-1 Crack Opening Displacement Versus Axial Position: Innermost Penetration,CE-Fabricated Westinghouse 4-Loop ............................................................................... C-4

Figure C-2 Crack Opening Displacement Versus Axial Position: Outermost Penetration,CE-Fabricated Westinghouse 4-Loop ............................................................................... C-5

Figure C-3 Crack Opening Displacement Versus Axial Position: Innermost Penetration,Westinghouse 3-Loop ....................................................................................................... C-6

Figure C-4 Crack Opening Displacement Versus Axial Position: Outermost Penetration,Westinghouse 3-Loop ....................................................................................................... C-7

Figure E-1 Expansion Cooling Heat Sink Rate Versus Leak Rate Assuming InitialCoolant at 550F ............................................................................................................... E-3

Figure E-2 Example Thermal Analysis Results: Temperature Contours (F) for aUniform 3000 Btu/h Heat Sink on 45 Total Arc Surface Corresponding to CompleteVaporization of a 0.008 gpm Leak .................................................................................... E-4

Figure E-3 Average Metal Temperature Along Annulus Leak Path Versus Heat SinkMagnitude ......................................................................................................................... E-5

Figure F-1 Finite Element Model of Bottom Head for Allowable Wastage VolumeCalculation (B&W Design)..................................................................................................F-4

Figure F-2 Finite Element Model of Bottom Head for Allowable Wastage VolumeCalculation (CE-Fabricated Westinghouse 4-Loop Design)...............................................F-5

Figure F-3 Finite Element ModelWastage Between Adjacent Nozzles (B&W Design) .........F-6

Figure F-4 Finite Element ModelWastage Between Adjacent Nozzles (CE-FabricatedWestinghouse 4-Loop) .......................................................................................................F-7

Figure F-5 Finite Element Analysis ResultsWastage Between Adjacent Nozzles (B&W

Design)...............................................................................................................................F-8Figure F-6 Finite Element Analysis ResultsWastage Between Adjacent Nozzles (CE-

Fabricated Westinghouse 4-Loop) .....................................................................................F-9

Figure F-7 Finite Element ModelWastage Distributed Around Single Nozzle (B&WNozzle) .............................................................................................................................F-10

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Figure F-8 Finite Element ModelWastage Distributed Around Single Nozzle (CE-Fabricated Westinghouse 4-Loop) ...................................................................................F-11

Figure F-9 Finite Element Analysis ResultsWastage Distributed Around Single Nozzle(B&W Design) ..................................................................................................................F-12

Figure F-10 Finite Element Analysis ResultsWastage Distributed Around Single

Nozzle (CE-Fabricated Westinghouse 4-Loop)................................................................F-13

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LIST OF TABLES

Table 2-1 Summary of Results of Probabilistic Calculations for Basic Inspection PlanOptions.............................................................................................................................2-11

Table 4-1 Bottom Head Design/Fabricator Groups..................................................................4-10

Table 6-1 Input Statistical Distributions Used in Monte Carlo Calculations of Wastage ..........6-26

Table 6-2 Additional Inputs to Monte Carlo Calculations to Investigate Benefit ofSupplemental Visual Examinations and Ultrasonic Volumetric Examinations .................6-27

Table 6-3 Results of the Monte Carlo Simulations: Base Cases for Inspection Programswith Periodic Bare Metal Visual Examinations .................................................................6-28

Table 6-4 Results of the Monte Carlo Simulations: Effect of SV Sensitivity for BMVEvery Third Outage for 18-month Cycle (Base Case 22).................................................6-29

Table 6-5 Results of the Monte Carlo Simulations: Effect of SV Sensitivity for BMVEvery Other Outage for 24-month Cycle (Base Case 16) ................................................6-29

Table 6-6 Results of the Monte Carlo Simulations: Cases for Inspection Programswithout Future Periodic Bare Metal Visual Examinations for 18-month Cycle .................6-29

Table 6-7 Results of the Monte Carlo Simulations: Cases for Inspection Programswithout Future Periodic Bare Metal Visual Examinations for 24-month Cycle .................6-30

Table 6-8 Results of the Monte Carlo Simulations: Sensitivity of Results to VolumetricExamination Coverage Assumption .................................................................................6-30

Table 6-9 Results of the Monte Carlo Simulations: Distribution Parameter ValueSensitivity Cases Based on Case 22 (BMV Every Third Outage with SV duringOther Outages for 18-month Cycle) .................................................................................6-31

Table 6-10 Results of the Monte Carlo Simulations: Distribution Type Sensitivity CasesBased on Case 22 (BMV Every Third Outage with SV during Other Outages for 18-month Cycle) ....................................................................................................................6-32

Table 6-11 Results of the Monte Carlo Simulations: Sensitivity of Case 13 (BMV EveryOther Outage for 24-month Cycle) and Case 14 (BMV Every Other Outage plusVolumetric Examination Every 10 Years for 24-month Cycle) to 38% Reduction inthe Wastage Rate Multiplicative Factor Parameter..........................................................6-32

Table D-1 Leak Rate Calculation Results................................................................................ D-3

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1INTRODUCTION

This introductory section provides a brief background discussion, describes the basic BMNdesign configuration, defines the purpose and scope of this safety evaluation, and outlines theapproach used. This section also outlines how this report is organized.

1.1 Background

As described in Section 3.2.1, in 2003, two leaking bottom-mounted nozzles (BMNs) located on

the reactor vessel bottom head were detected at one PWR unit. The leakage was the result ofprimary water stress corrosion cracks in the Alloy 600 nozzle material and Alloy 182 J-grooveattachment weld. Although no significant wastage resulted from this leakage, boric acidcorrosion of the low-alloy steel shell material of the bottom head due to BMN leakage is one ofthe two principal potential safety concerns associated with aging degradation of BMNs. Theconcern for boric acid wastage of low-alloy steel components due to concentration of leakingprimary coolant was demonstrated by the large wastage cavity produced on the reactor vesselclosure head at the Davis-Besse plant, which was detected in 2002 [1, 2].

The other potential safety concern is ejection of a BMN due to circumferential primary waterstress corrosion cracking (PWSCC) in the nozzle tube material below the elevation of the bottomof the J-groove attachment weld. This other concern has been evaluated by the Westinghouse

and B&W Owners Groups [3, 4, 5, 6].

1.2 Basic BMN Configuration

The basic BMN design configuration, which is very similar to that of other small-bore Alloy 600J-groove penetrations such as RCS piping and pressurizer instrumentation penetrations andpressurizer heater sleeves, is shown in Figure 1-1. The Alloy 600 nozzle typically has a nominaloutside diameter of 1.5 inches. The function of the BMNs is to provide access to the reactor corefor incore instrumentation. Fifty-eight of the set of 69 currently operating domestic PWR plantshave BMNs located on the bottom head of the reactor vessel. In the other 11 units, the incore

instrumentation (ICI) penetrations are in the closure head of the reactor rather than the bottomhead. In Westinghouse plants the BMNs are typically referred to as BMI nozzles, and at B&Wplants the term IMI nozzle is used.

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Introduction

The seven B&W design units currently operating in the U.S. have a modified configuration asshown in Figure 1-2. The modified configuration was implemented before these plants werecommercially operated because the original design was structurally insufficient given the flowforces in the vessel lower plenum. Note that the original weld and nozzle were installed at thetime of the vessel stress relief, meaning that the peak welding residual stresses in the original

nozzle were significantly reduced prior to the new weld being applied. This repairedconfiguration was included in the set of finite element stress calculations described inAppendix A. Finally, note that the B&W BMN configuration in the region below the bottom ofthe original J-groove weld, where through-wall nozzle cracks would result in leakage, isbasically identical to the standard BMN design.

1.3 Purpose

The purpose of the wastage evaluations documented in this report is to determine the effect onnuclear safety of the potential for boric acid corrosion of the low-alloy steel material of thereactor vessel bottom head due to potential BMN leakage. This report is one of the principalinputs to the Materials Reliability Program (MRP) bottom-mounted nozzle inspection plan(MRP-206; EPRI 1015003), which is expected to define inspection requirements for the long-term management of the potential for PWSCC of the Alloy 600 nozzles and Alloy 82/182J-groove welds of the penetrations mounted on the reactor vessel bottom head of PWRs. Thisdocument does not make any inspection requirements or commitments (nor any other types ofrequirements or commitments), but rather is an input to the development of appropriate long-term BMN inspection practices.

1.4 Scope

The evaluations of this report address the potential for bottom head wastage due to potentialBMN leakage at any of the 58 operating domestic PWR units that have BMNs. The other 11operating domestic units do not have any penetrations in the hemispherical lower head. Asdescribed in Section 4.2.2, stress calculations were performed for six designs representative ofthe full group of 58 units. The results of these representative cases were used to develop inputsto the deterministic and probabilistic wastage models (including crack-tip stress intensity factorand leak rate inputs) representing the entire set of 58 units. Note that the butt welds between theAlloy 600 instrumentation tube and the stainless steel guide tube or the instrumentation tubesafe-end are located well below the outer head surface and as such are outside the scope of thisboric acid wastage evaluation.

1.5 Approach

The safety evaluations in this study are based on plant experience with leaking J-groovepenetrations in combination with deterministic and probabilistic modeling of the wastageprocess. The approach taken is similar to the approach documented in Section 7 and AppendicesD and E of report MRP-110 [7] to evaluate the potential for wastage of reactor vessel closureheads given potential leakage of CRDM or other closure head penetrations.

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Introduction

The deterministic modeling approach, which investigates the benefit of periodic bare metalvisual examinations, is based on the time for a leak to increase from the point it is detectable viathe presence of boric acid deposits until the point that the leak rate is sufficiently high to supportextensive local cooling and potentially rapid wastage rates. This time is the available period foridentifying the presence of a leak prior to the potential for a structurally significant wastagecavity to be produced. The probabilistic modeling approach, which investigates the benefit ofinspection programs including supplemental visual inspections and volumetric examinations ofthe BMN tube in addition to bare metal visual examinations, results in a calculated frequency ofhead rupture that can be compared to the standard risk criterion of 1E-06 per year frequency ofcore damage if it is conservatively assumed that the head rupture results in a loss of coolant rategreater than that which can be replaced by emergency core cooling systems.

The model of the wastage process includes the main physical processes that are judged to controlthe potential for a loss-of-coolant accident (LOCA): crack initiation, growth of a part-depthcrack, leak initiation, growth of a through-wall crack, increase in leak rate and extent of localcooling, increase in resulting wastage rate, increase in area of unsupported vessel inside claddingmaterial, and ultimate rupture of the unsupported cladding. Additional calculations are used to

verify the tolerance of the low-alloy steel shell to significant material volume loss. Theprobabilistic wastage model assumes that the wastage progression continues to rupture unless thecracking is detected via a volumetric examination or the leakage is detected via a bare metal orsupplemental visual examination, or via an increase in the magnitude of the unidentified leak rateto 1 gpm or more. The supplemental visual examination is distinct from the more sensitive baremetal visual examination, which is a 360 examination of the nozzle annulus and adjacent area.The supplemental visual examination is a general examination from multiple vantage points inwhich the extent of visual access that must be attained is defined by the specified volume ofboron deposits that must be detectable (see Section 2.2).

1.6 MRP Efforts to Refine Understanding of the Wastage Process

The wastage models presented in this report were designed as tools for investigating the benefitof various programs of periodic inspections on the likelihood of significant low-alloy steelwastage occurring on the basis of an assumed linear wastage rate versus leak rate relationship.This relationship was developed on the basis of the extensive local cooling that is required for ahighly concentrated aqueous solution or hydrated slurry to exist over a relatively large (i.e., > 2in

2) surface of the low-alloy steel shell material. Such conditions are believed to be necessary for

the potential to exist for rapid corrosion over a structurally significant region. The wastagemodels are not intended to reflect the physical details of the several complex corrosionmechanisms that potentially may be active. Insufficient experimental data are currently availableto support development of such explicit models. Rather, the wastage models are intended to

conservatively bound the net effects of all active corrosion mechanisms on the structuralintegrity of the head by assuming that the input wastage rate versus leak rate curve describes theaverage linear material loss rate over a relatively large (i.e., > 2 in

2) area of the low-alloy steel

material within the nozzle crevice.

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Introduction

The statistical inputs to the probabilistic evaluation presented here were designed to capture theprocess uncertainties to the extent possible at the current time. However, in order to refineunderstanding of the wastage process, an extensive MRP laboratory boric acid corrosion testprogram is currently ongoing (see Section 5.5). This program is now beginning to yield resultsthat will allow more refined models of the individual wastage mechanisms to be developed and

the main assumptions of the models presented in this report to be checked. The MRP testprogram is expected to culminate in full-scale mockup tests that will simulate actual conditionsfor a leaking CRDM nozzle geometry, with possible consideration of an inverted nozzle configuration.As additional results are generated under the MRP test program and the results are evaluatedwith regard to the BMN location, it is possible that the conclusions of this report could beaffected. The MRP will continue to consider the latest available data in its development ofinspection guidance.

1.7 Report Structure

The organization of this safety assessment report is described below. The appendices provide

additional details on the supporting calculations that are summarized in Section 4.

1. Summary and Conclusions (Section 2)

Section 2 summarizes the results of this study investigating the benefit of candidate programsof BMN inspection including bare metal visual, supplemental visual, and ultrasonicexamination for reducing the risk associated with the potential for bottom head wastage dueto BMN leakage. Basic inspection programs are identified that result in an acceptably loweffect on nuclear safety.

2. Relevant Plant Experience and Laboratory Testing (Section 3)

Section 3 discusses the limited plant experience with leaking BMNs and the more extensiveexperience with leakage from other J-groove Alloy 600 penetrations. The experience withleaking Alloy 600 pressurizer heater sleeves is of particular relevance to the inverted BMNgeometry. Section 3 also cites the body of laboratory test data for boric acid corrosion. Adiscussion of the ongoing MRP boric acid corrosion test program in the context of thewastage evaluations of this report is deferred until Section 5.

3. Supporting Calculations (Section 4)

Development of the bottom head wastage models required that a number of supportinganalyses be performed. This section (with related appendices to this report) presents themethodologies and key results of stress analyses, stress intensity factor fracture mechanicscalculations, crack opening displacement calculations, leak rate calculations, thermalmodeling of leakage, calculations of allowable wastage volume without ASME Codeallowable stress levels being exceeded in the head shell, and a brief review of the claddingburst-strength study performed by Oak Ridge National Laboratory (ORNL) under NRCfunding.

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Introduction

4. Deterministic Wastage Evaluation (Section 5)

Section 5 describes the deterministic evaluation of the wastage process that specificallyexamines the benefit of bare metal visual examinations. Similar to the deterministicapproach taken in MRP-110 [7] for leaking reactor vessel closure head nozzles, thedeterministic approach in Section 5 is based on the time that is required for the leak rate toincrease from the rate that is readily detectable via bare metal visual examination to the leakrate that is sufficiently high to produce the extensive local cooling that is necessary for rapidwastage to proceed over a relatively large surface area (i.e., > 2 in

2) of the low-alloy steel

head material (within the nozzle crevice and/or on the outer head surface). Section 5.5discusses the ongoing MRP boric acid corrosion test program in the context of the wastageevaluations of this report.

5. Probabilistic Wastage Modeling (Section 6)

Similar to the probabilistic approach taken in Appendix E of report MRP-110 [7] to evaluatethe potential for wastage of reactor vessel closure heads given potential leakage of CRDM or

other closure head penetrations, Section 6 presents the methodology and results of theprobabilistic model of the bottom head wastage process. The probabilistic model for bottomhead wastage addresses the main physical processes that are judged to control the potentialfor a loss-of-coolant accident (LOCA), and it also evaluates the benefit of periodic bare metalor supplemental visual examinations in detecting the leakage before a LOCA can occur dueto wastage. The model also was used to evaluate the benefit of ultrasonic examinations ofthe BMN tube in detecting cracking before a through-wall crack and leakage are produced, orafter leakage is produced but not yet detected.

6. References (Section 7)

This section lists the references cited in the body and appendices of this report.

7. Appendix A: Bottom Head Penetration FEA Stress Model

Appendix A provides background regarding the finite element analysis (FEA) penetrationstress model which was used to support the development of several wastage model inputs.The model, which was initially developed by Dominion Engineering, Inc. (DEI) in the early2-2

8. Appendix A summarizes the basic modeling approach, geometry, and key assumptions,including the finite element grids used for each of the six representative BMN cases.

9. Appendix B: Crack-Tip Stress intensity factor Computations

The methodology for computing the crack-tip stress intensity factor of a through-wall axial

nozzle crack is presented in Appendix B. This crack geometry is assumed in the wastagemodels to produce the leakage necessary for boric acid corrosion of the low-alloy steel headto occur. The stress intensity factor calculations are based on an FEA linear elastic fracturemechanics model, which takes as an input an initial stress state from the FEA weldingresidual stress model described in Appendix A.

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10.Appendix C: Crack Opening Displacement Calculations

Appendix C presents the crack opening displacement (and crack opening area) calculationsthat are a key input to the leak rate calculations. Finite element modeling was used toevaluate crack opening displacements for postulated through-wall axial nozzle cracks ofvarying lengths.

11.Appendix D: Leak Rate Calculations

The bottom head wastage models take as one of their key inputs the estimated leak rate thatresults from a postulated through-wall axial nozzle crack. Appendix D summarizes the leakrate calculations that were performed using two different existing methods: the EPRI PICEPcomputer code (designed for leaks in LWR piping and steam generator tubes) and amethodology developed by Laborelec for steam generator tube leaks. Along with theempirical plant experience for leaking J-groove penetrations, these leak rate calculationswere applied in the selection of the leak rate curve input to the bottom head wastage models.

12.Appendix E: Thermal Modeling of Leaks

The FEA stress model was modified to perform thermal modeling to assess the coolingeffects of a BMN leak on the adjacent region of the bottom head (annulus and outer surface).Key aspects of the methodology that was used to perform this assessment are presented inAppendix E. The extent of cooling is a key parameter controlling the extent of expectedwastage.

13.Appendix F: Allowable Wastage Volume Calculations

Appendix F presents calculations used to verify the tolerance of the low-alloy steel shell tosignificant material volume loss. In the probabilistic wastage model, rupture of anunsupported area of the stainless steel head cladding is assumed to be the ultimate failure

mechanism rather than rupture of the low-alloy steel shell material. This assumption is basedon the application of available cladding burst-strength data (Section 4.8) in comparison withthe results of Appendix F.

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Introduction

Figure 1-1Location of Bottom-Mounted Nozzles (BMNs) on Bottom Head of PWR Reactor Vessel and

Typical Basic BMN Configuration

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Introduction

Figure 1-2BMN Configuration for B&W Design Plants

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2SUMMARY AND CONCLUSIONS

2.1 Summary

This report documents the methods and results of evaluations of the potential for wastage of thelow-alloy steel shell of the bottom head given potential BMN PWSCC and leakage. Thepotential for significant wastage was evaluated through a detailed review of plant experiencewith leaking J-groove nozzles and other sources of leakage in combination with deterministicand probabilistic models of the wastage process. These models are based on the results of the

supporting calculations documented in Section 4 and the appendices along with the available setof laboratory boric acid corrosion tests.

2.1.1 Review of Relevant Plant Experience

Section 7 of MRP-110 [7] documented and discussed in detail the plant experience with primarycoolant leakage that strongly supports the conclusion that visual examinations performed atappropriate intervals are effective at detecting J-groove nozzle leakage before such leakage couldproduce structurally significant amounts of carbon or low-alloy steel wastage. The review inMRP-110 included the following experience:

Davis-Besse Operating Experience (2002) [1, 2]. The Davis-Besse experience indicates thatbare metal visual examinations performed every refueling outageand likely even ifperformed less frequentlywith proper follow-up action would have caught this headdegradation early in the material loss process.

U.S. Experience with Leaking CRDM Nozzles. Although about 55 CRDM penetrations havebeen detected as leaking in the U.S., significant wastage in the surrounding head materialwas only reported for Davis-Besse nozzle nos. 2 and 3. The wastage at these nozzles wasaccompanied by evidence of leakage that was readily detectable several years (estimated tobe at least 6 years) prior to the large cavity adjacent to nozzle no. 3 being discovered.

In general, no measurable wastage was reported in the surrounding head material for the

other leaking CRDM nozzles although ultrasonic leak path technology inspections appearto show loss of the small interference fit (nominally within the range 0.000 to 0.002 inches (0

to 50 m) radial interference for most plants) along a portion of the nozzle OD. The leakpath technology results appear to be consistent with the 65 m of material loss cited forBugey 3 in France (see below). It is noted that for two other cases of leaking CRDMnozzles, visible but small wastage volumes (estimated to be less than 1 in

3in each case) have

been observed.

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Summary and Conclusions

U.S. Experience with Other Alloy 600 Penetrations and Alloy 182/82 Welds. Despite over100 of such components (including pressurizer instrumentation nozzles and heater sleeves,RCS hot-leg piping instrumentation nozzles, and small-bore reactor vessel closure headthermocouple nozzles) reported to have leaked due to PWSCC, significant wastage of thesurrounding carbon or low-alloy steel material has rarely been reported. The repair activities

that were performed subsequent to such leakage being detected would generally have beenexpected to reveal significant wastage if present given the guidance developed by plantsregarding boric acid corrosion following NRC Generic Letter 88-05 [8].

U.S. Experience with Leakage from Mechanical Joints and Seal Welds. This additionalleakage experience also tends to support the conclusion that relatively large leak rates (e.g.,on the order of 0.05 gpm or more) are required to produce significant volumes of wastage oflow-alloy steel vessel material. However, it should be noted that in many cases, reliablemeasures of the actual relevant crack leak rate are not available.

International Experience. Despite hundreds of PWSCC indications having been observedabroad, the only CRDM nozzle indication reported to have resulted in a through-wall leakwas nozzle no. 54 at Bugey 3 (EDF). Corrosion of the head material adjacent to this nozzle

opened a narrow gap between the nozzle and the borehole approximately 65 m (2 to 3 mils)wide. The corrosion was centered on a trail leading from the point in the borehole adjacentto the crack upward to the vessel head surface. It was estimated that the crack had beenleaking for approximately 20,000 hours (2.3 years).

Another key event occurred at the Beznau plant in Switzerland in 1970. Leakage above thereactor vessel head led to the accumulation of 12 m

3(60,000120,000 in

3) of boric acid

deposits and a maximum wastage depth into the top head surface of 1.6 inches. The volumeof boric acid deposits reported for the Beznau experience indicates a leak rate during thecorrosion progression likely greater than 0.1 gpm.

Because this review of plant experience concentrated on the implications for wastage of reactorvessel closure heads, this study examined in further detail the available experience most relevantto the inverted geometry of BMNs:

South Texas Project Unit 1 BMNs (2003). The only reported cases of leaking BMNs werenozzle nos. 1 and 46 at STP Unit 1 in 2003. Visual examinations did not reveal evidence ofwastage around these two penetrations, and a lack of significant wastage was confirmed byfailed attempts to insert a 0.0015-inch feeler gauge between the nozzle and hole in the vesselshell.

Arkansas Nuclear One Unit 2 Pressurizer Heater Sleeve (1987). The experience for theleaking heater sleeve on the pressurizer bottom head at ANO Unit 2 is particularly relevant to

the inverted geometry of the reactor vessel bottom head. The approximate 1.0-inch outsidediameter of the leaking ANO Unit 2 heater sleeve is also similar to the typical 1.5-inchoutside diameter of BMNs.

A region of low-alloy steel wastage approximately 1.5 inches in diameter and 0.75 inch deep(about 18% through the thickness) was detected on the outer head surface adjacent to thecracked heater sleeve. The total wastage volume is estimated to be less than 1 in

3,

significantly smaller than the size that would be structurally significant. Repair activitiesconfirmed that there was no detectable wastage inside the annulus area of the penetration.

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The length of time that the leak existed prior to detection is not known. Therefore, no linearor volumetric wastage rate can be reliably estimated from the ANO-2 experience.

This leak occurred prior to NRC Generic Letter 88-05 requirements and current industrypractices of rigorous walkdowns. However, since the implementation of more rigorousinspections, many leaks at pressurizer heater sleeves have been identified and repaired, andnone of these leaks produced any measurable wastage. From this operating experience, it canbe concluded that the length of time required to produce leakage and wastage of themagnitude reported at ANO-2 in 1987 is much longer than the time between the currentpressurizer bottom head inspections, which have typically been performed each refuelingoutage since the discovery in 1987.

Calvert Cliffs Unit 2 Pressurizer Heater Sleeve H3 Destructive Examination (1989). In1989, after about 9.6 EFPYs of operation, boric acid deposits were found around 20 heatersleeve penetrations at Calvert Cliffs 2 during a routine in-service inspection [9]. Inspectionsshowed that 28 of 120 sleeves had axial indications in the upper 34 inches of the sleeve. Noboric acid deposits had been found during a 1987 inspection. A detailed root cause

evaluation including destructive examinations of two sleeves concluded that PWSCC of theAlloy 600 sleeves caused the leakage observed, accelerated by the presence of the coldworked layer associated with reaming of the sleeve ID surface performed prior to welding[10].

The sample for one of the two sleeves destructively examined (Sleeve H3) was a core samplethat included a ring of low-alloy steel pressurizer shell material and the Alloy 182 J-grooveweld, in addition to the section of Alloy 600 sleeve located within the pressurizer bottomhead. (The other sample examined was cut exclusively from another sleeve.) Photographs ofthe outer head surface of the core sample and of the surface of the shell borehole (made afterthe low-alloy steel ring was split) show evidence only of minor surface corrosion. Asdocumented by photographs of the as-found deposits on the pressurizer bottom head,

evidence of leakage was apparent even with the insulation in place given the tendency of thedeposits to flow down through the gaps between the insulation and the heater sleeves. SleeveH3 was reported to be one of the heater sleeves with the largest amount of boron depositssurrounding it. In summary, this heater sleeve experience demonstrates that only minoramounts of surface corrosion are expected for the inverted nozzle geometry, even in the casethat visual examinations for evidence of leakage are performed with the insulation in place(provided that there is a gap between the nozzle and insulation sufficient for deposits to rundown below the insulation).

RCS Piping Instrumentation Nozzles and Pressurizer Heater Sleeves. An informal survey ofplant experience was completed with regard to the appearance of boric acid deposits andextent of any carbon or low-alloy steel wastage specific to the geometry of small-bore Alloy600 RCS piping instrumentation nozzles and pressurizer heater sleeves. No cases ofsignificant wastage of carbon steel RCS piping have been reported as a result of RCS pipinginstrumentation nozzle leaks. Even in the cases reported for which the accumulated boricacid was significantly discolored due to the presence of corrosion products, generally nomaterial loss of the pipe was visible and any corrosion was limited to light surface corrosion(i.e., less than 1/16 inch deep). Furthermore, experience with over 40 leaking heater sleevessince 1992 without any reported cases of visible loss of shell material shows that periodicvisual examinations, even if performed with the insulation in place, have identified leakage

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prior to the onset of significant wastage. The general experience with leaking pressurizerheater sleeves, including that for Calvert Cliffs 2 in 1989, is that even relatively smallvolumes of boric acid deposits tend to become visible for the inverted nozzle geometry, evenwith the insulation in place given the tendency of the deposits to flow down through the gapsbetween the insulation and nozzles. Finally, this survey did not reveal any trends in thedetectability of leakage through visual examination for boric acid deposits or in the extent ofcarbon steel wastage (if any) as a function of the circumferential position of the nozzlearound the RCS pipe circumference.

Section 3 discusses this experience with the inverted nozzle geometry in more detail, along withsome discussion of the inverted geometry, leaking annulus boric acid corrosion tests performedby ABB-CE in 1991. This further review of plant experience also supports the adequacy ofperiodic visual examinations to prevent significant wastage of the bottom head. The extendedreview of the leakage experience does not indicate any special concerns with respect to wastagerates or the detectability of boric acid deposits for the inverted configuration.

2.1.2 Deterministic Wastage Model

The deterministic evaluation of the wastage process specifically examines the benefit of baremetal visual examinations. The probabilistic evaluation summarized in Section 2.1.3 wasapplied to investigate the benefit of supplemental visual and ultrasonic examinations in additionto bare metal visual examinations.

Model Description. Similar to the deterministic approach taken in MRP-110 [7] for leakingreactor vessel closure head nozzles, the deterministic model for bottom head wastage is based onthe time that is required for the leak rate to increase from the rate that is readily detectable viabare metal visual examination to the leak rate that is sufficiently high to produce the extensivelocal cooling that is necessary for rapid wastage to proceed over a relatively large surface area ofthe low-alloy steel head material (within the nozzle crevice and/or on the outer head surface).The term relatively large area in this report refers to an area similar in magnitude to the area ofthe bore surface of low-alloy steel material on one side of the nozzle (i.e., > 2 in

2) as opposed to

the relatively small zone (e.g., < 0.15 in2) of jet impingement in the case that a jet exiting the

crack impinges the head material a fraction of an inch away. Consistent with plant experienceand laboratory boric acid corrosion testing, large local cooling is a prerequisite for rapid wastageto occur over a relatively large region since large local cooling is required for an aqueousconcentrated boric acid solution (or hydr

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