ML082540053

Response to Request for Additional Information Concerning Authorization to Extend the Third 10-Year Inservice Inspection Interval for Reactor Vessel Weld Examination

ML082540053
Person / Time
Site:	Palisades
Issue date:	08/14/2008
From:	Schwarz C Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
FOIA/PA-2010-0209
Download: ML082540053 (275)

Text

{{#Wiki_filter:Entergy Nuclear Operations, Inc. Palisades Nuclear Plant 27780 Blue Star Memorial Highway Covert, MI 49043 Tel 269 764 2000 August 14, 2008 10 CFR 50.55a U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Palisades Nuclear Plant

,Docket 50-255 License No. DPR-20 Request for Additional Information Concerning Authorization to Extend the Third 10-Year Inservice Inspection Interval for Reactor Vessel Weld Examination

Dear Sir or Madam:

By letter dated July21, 2008, Entergy Nuclear Operations, Inc. (ENO) requested Nuclear Regulatory Commission (NRC) review and approval for the use of an alternative to the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section Xl, paragraph IWB-2412, Inspection Program B, for the Palisades Nuclear Plant (PNP) (ADAMS Accession number ML082040342.) NRC approval was requested to extend the third inspection interval for examination of the reactor vessel (RV) welds (Category B-A), the nozzle-to-vessel welds and inner radius sections (Category B-D) until December 12, 2015. During a telephone conference on July 31, 2008, and subsequent electronic mail on August 6, 2008, the NRC requested additional information on the proposed alternative. Enclosure 1 provides ENO's responses. Summary of Commitments This letter contains no new commitments and no revisions to existing commitments. -ýGýhT ifher J. Sch Site Vice President Palisades Nuclear Plant Enclosure CC Administrator, Region Ill, USNRC Project Manager, Palisades, USNRC Resident Inspector, Palisades, USNRC Ioq 7

                                     *ENCLOSURE REQUEST FOR ADDITIONAL INFORMATION CONCERNING AUTHORIZATION TO EXTEND THE THIRD 10-YEAR INSERVICE INSPECTION INTERVAL FOR REACTOR VESSEL WELD EXAMINATION PALISADES NUCLEAR PLANT By letter dated July 21, 2008, Entergy Nuclear Operations, Inc. (ENO) requested Nuclear Regulatory Commission (NRC) review and approval for the use of an alternative to the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI, paragraph IWB-2412, Inspection Program B, for the Palisades Nuclear Plant (PNP) (ADAMS Accession number ML082040342.) NRC approval was requested to extend the third inspection interval for examination of the reactor vessel (RV) welds (Category B-A); the nozzle-to-vessel welds and inner radius sections (Category B-D) until December 12, 2015. During a telephone conference on July 31, 2008, and subsequent electronic mail on August 6, 2008, the NRC requested additional information on the proposed alternative. ENO's responses are provided below.

NRC Request

1. The dates for the next two scheduled [inservice inspection] ISI of the reactorvessel beltline welds.

ENO Response

1. ENO plans to conduct a RV'examination during the 2010 refueling outage in anticipation of issuance of the proposed Pressurized Thermal Shock rule. In

    -accordance with WCAP-16168-NP-A, revision 2, the subsequent RV examination after that would occur in 2030 or earlier, with the interval not exceeding twenty years.

NRC Request

2. A copy of WCAP-16168-NP-A, Revision 2.

ENO Response

2. A copy of WCAP-16168-NP-A, revision 2 is attached (Attachment 1).

Page 1 of 1

Attachment 1 WCAP-16168-NP-A, Revision 2, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval" 263 Pages Follow

Westinghouse Non-Proprietary Class 3 WCAP-16168-NP-A June 2( )08 Revision 2 Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval )Westinghouse

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-16168-NP-A Revision 2 Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval Bruce A. Bishop Cheryl L. Boggess Nathan A. Palm June 2008 Approved: Electronicall'Approved* Patricia C. Paesano, Manager Primary Component Asset Management Approved: ElectronicallyApproved* Gordon C. Bischoff PWR Owners Group This work was performed for the PWR Owners Group under PWROG Project MUHP-5097, MUHP-5098, MUHP-5099, PWROG Project Authorization MSC-0119, MSC-0120 and CEOG Task 2008, 2059.

   *Electronically approved records are authenticated in the Electronic Document Management System Westinghouse Electric Company LLC P.O. Box 355 Pittsburgh, PA 15230-0355
                               © 2008 Westinghouse Electric Company LLC All Rights Reserved

iii UNITED STATES NUCLEAR REGULATORY COMMISSIOfl E 'E1V 'D WASHINGTON, D.C. 20555-0001E

              * ***May                             8, 2008MA                                       420 Mr. Gordon Bischoff, Manager          M                                             MAY 1 4 2QO0 Owners Group Program Management Office Westinghouse Electric Company                                                          PWROG P.O. Box 355                                                                        Project Office Pittsburgh, PA 15230-0355

SUBJECT:

FINAL SAFETY EVALUATION FOR PRESSURIZED WATER REACTOR OWNERS GROUP (PWROG) TOPICAL REPORT (TR) WCAP-16168-NP, REVISION 2, "RISK- INFORMED EXTENSION OF THE REACTOR VESSEL IN-SERVICE INSPECTION INTERVAL" (TAC NO. MC9768)

Dear Mr. Bischoff:

By letter dated January 26, 2006, as supplemented by letter dated June 8, 2006, the PWROG submitted TR WCAP-16168-NP, Revision 1, to the U.S. Nuclear Regulatory Commission (NRC) staff. TR WCAP-16168-NP, Revision 2, and responses to the NRC staff's request for additional information (RAI) on TR WCAP-1 6168-NP, Revision 1, were submitted for NRC staff review by PWROG letter dated October 16, 2007, By letter dated March 6, 2008, an NRC draft safety evaluation (SE) regarding our approval of TR WCAP-16168-NP, Revision 2, was provided for your review and comments. By letter dated March 31, 2008, the PWROG commented on the draft SE. The NRC staffs disposition of PWROG's comments on the draft SE are discussed in the attachment to the final SE enclosed with this letter. The NRC staff has found that TR WCAP-16168-NP, Revision 2, is acceptable for referencing in licensing applications for Westinghouse, Combustion Engineering, and Babcock and Wilcox designed pressurized water reactors, for which an operating license was issued under Title 10 of the Code of FederalRegulations (10 CFR) Part 50 prior to the date of this letter, to the extent specified and under the limitations delineated in the TR and in the enclosed final SE. The final SE defines the basis for our acceptance of the TR. The NRC staff has accepted TR WCAP-16168-NP, Revision 2, based on the imposition of a condition related to the augmented evaluation of in-service inspection (IS0) results taken from Section (e) of the proposed 10 CFR 50.61 a, published in the Federal Register on October 3, 2007 (72 FR 56275). The NRC staff is in the process of reviewing public comments on the proposed rule and preparing the final rule. If the final 10 CFR 50.61a differs from the proposed 10 CFR 50.61a with regard to the augmented ISI evaluation requirements, the PWROG will be expected to review the requirements in the final 10 CFR 50.61a and determine whether a revision to the accepted TR WCAP-16168-NP, Revision 2, is required. The PWROG will be expected to notify the NRC staff, in writing, of the results of its determination within six months of the publication date of the final 10 CFR 50.61 a. If, on this basis, a revision to the accepted TR WCAP-16168-NP, Revision 2, is required, the PWROG will be expected to submit the revised TR for NRC staff review Within one year of the publication date of the final 10 CFR 50.6la. Furthermore, licensees that choose to implement 10 CFR 50.61a must perform the ISI required in Section (e) of the rule, and must submit the required information for review and approval to the Director, Office of Nuclear Reactor Regulation, in accordance with Section (c) of the rule, at least three years before the limiting RTP'rs value calculated under 10 CFR 50.61 is projected to WCAP-16168-NP-A June 2008 Revision 2

iv G. Bischoff -2 - exceed the PTS screening criteria in 10 CFR 50.61. Licensees implementing Section (c) of 10 CFR 50.61a must perform the inspections and analyses required by Section (e) of 10 CFR 50.61a prior to implementing the extended interval. Our acceptance applies only to material provided in the subject TR. We do not intend to repeat our review of the acceptable material described in the TR. When the TR appears as a reference in license applications, our review will ensure that the material presented applies to the specific plant involved. License amendment requests that deviate from this TR will be subject to a plant-specific review in accordance with applicable review standards. In accordance with the guidance provided on the NRC website, we request that PWROG publish the accepted version of this TR within three months of receipt of this letter. The accepted version shall incorporate this letter and the enclosed final SE after the tite page. Also, it must contain historical review information, including NRC requests for additional information and your responses. The accepted version shall include an "-A" (designating accepted) following the TR identification symbol. If future changes to the NRC's regulatory requirements affect the acceptability of this TR, the PWROG and/or licensees referencing it will be expected to revise the TR appropriately, or justify its continued applicability for subsequent referencing. Sincerely, Ho K. Nieh, Deputy Director Division of Policy and Rulemaking Office of Nuclear Reactor Regulation Project No. 694

Enclosure:

Final SE cc: Mr. James A. Gresham, Manager Regulatory Compliance and Plant Licensing Westinghouse Electric Company P.O. Box 355 Pittsburgh, PA 15230-0355

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                                                                                                               "t June 2008 WCAP- 16168-NP-A WCAP-16168-NP-A                                                                                             June 2008 Revision 2

V UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2055.000i FINAL SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION TOPICAL REPORT WCAP-16168-NP. REVISION 2, -RISK-INFORMED EXTENSION OF THE REACTOR VESSEL IN-SERVICE INSPECTION INTERVAL' PRESSURIZED WATER REACTOR OWNERS GROUP PROJECT NO. 694

1.0 INTRODUCTION AND BACKGROUND

By letter dated January 26, 2006, as supplemented by letter dated June 8, 2006, the Westinghouse Owners Group (WOG), currently known as the Pressurized Water Reactor Owners Group (PWROG), submitted topical report WCAP-16168-NP, Revision 1, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval" (Reference 1 and Reference 2), for U.S. Nuclear Regulatory Commission (NRC) staff review. By letter dated October 16, 2007, the PWROG submitted responses to the NRC staff's request for additional information (RAI) on WCAP-16168-NP, Revision 1, and provided WCAP-16168-NP, Revision 2 (Reference 3), but did not expand its scope as originally submitted for NRC staff review. In WCAP-16168-NP, Revision 2, (hereafter referred to as the TR) the PWROG provided the technical and regulatory basis for decreasing the frequency of inspections by extending the American Society of MechanicalEngineers Boilerand Pressure Vessel Code (ASME Code) Section XI inservice inspection (ISI) from the current 10 years to 20 years for ASME Code Section Xl, Category B-A and B-D reactor vessel (RV) welds.

• The TR described risk-informed pilot studies based, for the most part, on the results of the NRC's recently-completed pressurized thermal shock (PTS) research program. The NRC's Office of Nuclear Regulatory Research (RES) completed this research program to update the PTS regulations. In an October 3, 2007, Federal Register Notice (72 FR 56275) (Reference 4),

the NRC proposed to amend its regulations to provide updated fracture toughness requirements for protection against PTS events for PWR RVs. NUREG-1806, "'Technical Basis for Revision of the Pressurized Thermal Shock (PTS) Screening Limit in the PTS Rule (10 CFR 50.61)" (the PTS Risk Study) (Reference 5 and Reference 6) and (2) NUREG-1874, "Recommended Screening Limits for Pressurized Thermal Shock (PTS)" (Reference 7), provided the technical basis for the rulemaking. These reports summarized and referenced several additional reports on the same topic.

2.0 REGULATORY EVALUATION

ISI of ASME Code Class 1, 2, and 3 components is performed in accordance with Section XI of the ASME Code and applicable addenda as required by Title 10 of the Code of Federal Regulation (10 CFR) 50.55a(g), except where specific relief has been granted by the NRC ENCLOSURE WCAP-16168-NP-A June 2008 Revision 2

vi pursuant to 10 CFR 50.55a(g)(6)(i). The regulation at 10 CFR 50.55a(a)(3) states that alternatives to the requirements of paragraph (g) may be used, when authorized by the NRC, if: (i) the proposed alternatives would provide an acceptable level of quality and safety or

       .(ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

The regulations require that ISI of components and system pressure tests conducted during the first 10-year interval and subsequent intervals comply with the requirements in the latest edition

        *andaddenda of Section Xl of the ASME Code incorporated by reference in 10 CFR 50.55a(b) 12 months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein.

The current requirements for the inspection of RV pressure retaining welds have been in effect since the 1989 Edition of ASME Code, Section XI. Article IWB-2000 of the ASME Code, Section XI establishes an inspection interval of 10 years. The TR proposed a methodology that can be used by individual licensees to demonstrate that extending the inspection interval on their Category B-A pressure retaining RV welds and Category B-D full penetration RV nozzle welds from 10 to 20 years would provide an acceptable level of quality and safety. The NRC staff based its review of the risk information on NUREG-0800, "Standard Review Plan [(SRP)] for the Review of Safety Analysis Reports for Nuclear Power Plants," Chapter 19.2, "Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis: General Guidance" (Reference 8). SRP Chapter 19.2 directs the NRC staff to

• review each of the four elements suggested in Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Section 2 (Reference 9). These elements are: (1) Define the Proposed Changes, (2) Conduct Engineering Evaluations, (3) Develop Implementation and Monitoring Strategies, and (4) Document the Evaluations and Submit the Request.

The NRC staff also used further guidance in RG 1.174. RG 1.174 describes a risk-informed approach, acceptable to the NRC, for assessing the nature and impact of proposed licensing-basis changes by considering engineering issues and applying risk insights. One acceptable approach to making risk-informed decisions about the proposed change is to show that the proposed changes meet five key principles stated in RG 1.174, Section 2:

1. The proposed change meets the current regulations unless it is explicitly related to a

                -requested exemption or rule change.

2. The proposed change is consistent with the defense-in-depth philosophy.

3. The proposed change maintains sufficient safety margins.

4. When proposed changes result in an increase in core-damage frequency or risk, the increases should be small and consistent with the intent of the Commission's Safety Goal Policy Statement.

5. The impact of the proposed change should be monitored -using performance measurement strategies.

WCAP-16168-NP-A June 2008 Revision 2

vii RG 1.174 provides numerical risk acceptance guidelines that are helpful in determining whether or not the fourth key principle has been satisfied. These guidelines are not to be applied in an overly prescriptive manner; rather, they provide an indication, in numerical terms, of what is considered acceptable. The intent in comparing risk results with the risk acceptance guidelines is to demonstrate with reasonable assurance that the fourth key principle has been satisfied.

3.0 TECHNICAL EVALUATION

The objective of ISI is to identify conditions, such as flaw indications, that are precursors to leaks and ruptures and which violate pressure boundary integrity principles for plant safety. The TR includes a detailed analysis of the potential effects of extending the RV weld ISI interval for three pilot plants: Beaver Valley, Unit I (BV1), Palisades, and Oconee, Unit 1 (OCI). These three that were units include one unit from each of the PWR vendors and are the same plants evaluated in detail in the NRC PTS Risk Study. The TR proposed a method that each licensee could use to apply the results from the three pilot plant applications to its plant. The TR used the estimated through wall cracking frequency (TWCF) as a measure of the risk of RV failure. The correlation for determining plant-specific TWCF was basedon plant-specific data and can be found in NUREG-1874 (Reference 7). This correlation took into consideration the contribution to TWCF from each of the most limiting plate, forging, axial weld, and circumferential welds. These individual TWCF contributions were then weighted based on pilot plant data and summed to determine a total RV TWCF. 3.1 Define the Proposed Change The TR proposed to extend the inspection interval for ASME Code, Section XI, Category B-A and B-D RV welds from 10 years to a maximum of 20 years. The change will be accomplished through plant-specific requests for an alternative pursuant to 10 CFR 50.55a(a)(3Xi) on the basis that the alternative inspection interval provides an acceptable level of quality and safety. The 20 year inspection interval is a maximum interval and the PWROG did not request, and the NRC staff does not endorse, that all RV inspections be discontinued for the 10 years following approval of this methodology (as would occur if every licensee were granted an extension from 10to 20 years). In response to RAI 1 lb from Reference 3, the PWROG explained how a sampling of plants performing reactor inspections over the next 10 years can be achieved. In its request for an alternative, each licensee shall identify the years in which future inspections will be performed. The dates provided must be within plus or minus one refueling cycle of the dates identified in the implementation plan provided to the NRC in PWROG letter OG-06-356, "Plan for Plant Specific Implementation of Extended Inservice Inspection Interval per WCAP 16168-NP, Revision 1, "Risk Informed Extension of the Reactor Vessel In-Service Inspection Interval," MUHP 5097-99, Task 2059," dated October 31, 2006 (Reference 10).

• The inspection method, the acceptance criteria, and reporting requirements for inspection results that will modify from ASME Code requirements are discussed in section 3.3 of this safety evaluation (SE).

WCAP- 16168-NP-A June 2008 Revision 2

viii 3.2 Conduct Engineering Evaluations According to the guidelines in RG 1.174 and SRP Chapter 19.2, the second element associated with a risk-informed application is an analysis of the proposed change using a combination of traditional engineering analysis with supporting insights from a risk assessment. The objective of this study was to verify that a reduction in the frequency of volumetric examination of the RV full-penetration welds could be accomplished with an acceptably small change in risk. The methodology used to justify this reduction involved estimating the potential increase in risk caused by extending the RV inspection interval from 10 to 20 years. The increase in risk was evaluated against RG 1.174 criteria to determine if the values met the specified regulatory guidelines. The other key principles in RG 1.174 were also addressed in the evaluation. The intent was that licensees can then use the results of this bounding assessment to demonstrate that their RV and plant are bounded by the generic analysis, thereby justifying an extension of their plant-specific RV weld inspection interval. The engineering evaluations in the TR were based on the NRC staffs PTS Risk Study that is the technical basis for the proposed alternative fracture toughness requirements for pressurized thermal shock in 10 CFR 50.61a (Reference 4). 3.2.1 Engineering Evaluation The ISI interval extension methodology was primarily based on a risk analysis, including a probabilistic fracture mechanics (PFM) analysis of the effect of different inspection intervals on the frequency of RV failure due to postulated PTS transients. RV failure is defined for the purposes of this study as through-wall cracking of the RV wall. The likelihood of RV failure was postulated to increase with increasing time of operation due to the growth of pre-existing fabrication flaws by fatigue in combination with a decrease in RV fracture resistance due to irradiation. Credible, postulated PTS transients that could potentially lead to RV failure were considered to occur at the worst time in the life of the plant (as defined by flaw size and level of RV embrittlement). The PFM methodology allowed for the consideration of distributions and

       .uncertainties in flaw number and size, material properties, crack growth resulting from fatigue, accident transients, stresses, and the effectiveness of inspections. The PFM. approach led to a conditional RV failure frequency due to a given, loading condition and a prescribed inspection interval. The PFM analyses documented in the TR evaluated the impact of different inspection intervals on the three, previously-identified pilot plants.

Limiting Location for RV Failure To determine the limiting location in the RV, the PWROG evaluated the impact of flaws in each RV region. The PWROG used deterministic fracture mechanics analyses, which utilized a 10 percent through-wall flaw, assumed 40 effective full power years (EFPY) of embrittlement for the flaws in the RV beltline and included fatigue crack growth due to normal plant operating transients for all flaws. Each crack length was evaluated at the end of a 10 year interval to determine the maximum applied stress intensity factor (K4,,ppd). The ratio of the maximum allowable stress intensity factor (Kt1o4ae), per the ASME Code, Section Xl, Appendix A criteria, to KL~d was used as a measure of the margins to failure. The lower the ratio of KI.wbfdKI

        *

• the lower the margin to failure and the more'limiting the location. Figures 3-1 and 3-2 in WCAP- 16168-NP-A June 2008 Revision 2

ix the TR indicated that the beltline welds have the lowest ratio of ASME Code allowable stress intensity values (Kj ab,,*bj/KI 01,d)_ These figures do not include the full penetration nozzle-to-vessel welds. The NRC staff requested that the PWROG provide the ratio of ASME Code allowable stress intensity value for full penetration nozzle-to-vessel welds to demonstrate that the beitline welds were the limiting locations. In the response to RAI 5 from Reference 3, the PWROG provided the requested information. The PWROG analyses indicated that the bettline is more limiting than the full penetration nozzle-to-vessel welds. The results from the PWROG deterministic analyses were consistent with assumptions utilized in the NRC PTS Risk Study which concluded that the limiting RV region was the belttine region. Since the RV beltline region has the lowest margin to failure, the NRC staff also concluded that the bettline region is the most limiting location and the beltilne location can be used to determine the impact of different inspection intervals on the frequency of RV failure. Distributions and Uncertainties in Flaw Number and Size Section 3.2 of the TR indicated that surface-breaking and embedded flaws were used in the PFM analysisý Since embedded flaws do not grow significantly due to fatigue, they were not evaluated as part of the fatigue growth analysis. To simulate embedded flaws in welds and plates, the PWROG pilot plant studies for the RV ISI interval extension used the embedded flaw distribution for welds and plates from the NRC PTS Risk Study. Surface-breaking flaws were assumed to grow by fatigue as a result of normal operating conditions. A discussion of the initial size and distribution of the assumed surface-breaking flaws was provided by the PWROG in response to RAI I from Reference 3. The PWROG indicated that the initial size and distribution of the surface flaws were consistent with the size and distribution developed by Pacific Northwest National Laboratory (PNNL) for use in the NRC PTS Risk Study. The initial size and distribution of surface-breaking flaws utilized the computer code VFLA W03, which was developed by PNNL and is described in NUREG/CR-6817, Revision 1, "A Generalized Procedure for Generating Flaw-Related Inputs for the FAVOR Code* (Reference 11). The initial surface-breaking flaw size and distribution were input into a fatigue crack growth and IS[ analysis to determine a surface flaw density file after any ISI. Surface flaw density files were created to simulate two inspection routines. The first case simulated inspections performed on a 10 year interval as currently required by the ASME Code. The second case simulated a single inspection performed after the first 10 years of operation with no subsequent inspection. These surface-breaking flaw density files are then input into the PFM analysis as surface-breaking flaw density files. Since the characterization of embedded flaws in plates and welds and the initial surface-breaking flaw size for the~fatigue analysis used distributions that were used in the NRC PTS Risk Study, they are applicable for use in RV ISI interval extension analyses. In Attachment 1 to the June 8, 2006 letter (Reference 2), the PWROG indicated that underclad cracks in forgings are so shallow that the probability of them growing through-wall during a severe PTS transient would be fairly small. NUREG-1874 indicated that for severe PTS transients, the TWCF for forgings with underclad cracks can be greater than those for axial welds, plates and forgings without underclad cracks. In its response to RAI 2 from Reference 3, the PWROG provided an analysis of the TWCF for axial welds, plates, forgings without underclad cracks, and forgings with underclad cracks. The analysis, which used correlations WCAP- 16168-NP-A June 2008 Revision 2

X from NUREG-1874, indicated forgings with underclad cracks have a higher TWCF than welds, plates and forgings without underclad cracks when the RTMAX.,O 1 is greater than 240 OF. Table 3.4 in NUREG-1874 indicated that the highest RTMAx.Fo for a PWR RV ring forging is 187.3 OF at 32 EFPY and 198.6 OF at 48 EFPY. Therefore, it is unlikely that the RThx.po value for any domestic PWR will ever exceed 240 OF and the TWCF value for all such forgings will remain below that for axial welds with equivalent reference temperatures. The PWROG indicated that the analyses performed in the TR would not be applicable without further evaluation for RVs with RTmAK.FO values exceeding 240 "F. Fatique Crack Growth Analysis Section 3.2 of the TR indicated that the. pilot plant studies included a probabilistic representation of the fatigue crack growth correlation for ferritic materials in water consistent with the previous and current models contained in ASME Code, Section XI, Appendix A. The probabilistic representation was consistent with those used in the pc-PRAISE computer code and NRC-approved structural reliability and risk assessment (SRRA) tool for piping risk-informed ISI. In Appendix A of the NRC staff SE on WCAP-1 4572, Revision 1, "Westinghouse Owners Group Application of Risk-Informed Methods to Piping Inservice Inspection Topical Report" (Reference 13), the NRC staff concluded that the SRRA tool addresses fatigue crack growth in an acceptable manner since it is consistent with the technical approach used by other state-of-the-art PFM computer codes. The NRC staff noted that realistic predictions of failure probabilities require that the user define input parameters which accurately represent all sources of fatigue stress and the probability for preexisting fabrication defects in welds. As discussed in the preceding section of this SE, the size and distribution of preexisting surface-breaking. fabrication flaws was consistent with the size and distribution developed by PNNL for use in the NRC PTS Risk Study. Design basis transients for the pilot plants were reviewed and the PWROG determined that the greatest contributor to fatigue crack growth for surface-breaking flaws initiating from the inside surface of the RV for the pilot plants is the RV heat-up and cool-down transient. Each transient represents a full heat-up and cool-down cycle between atmospheric pressure at room temperature and full-system pressure at 100-percent power operating temperature. This transient envelopes many transients with smaller ranges of conditions. For the pilot plant evaluations, seven heat-up and cool-down cycles per year were used for the Westinghouse-designed plant, BV1, 13 heat-up and cool-down cycles were used for the Combustion Engineering (CE)-designed plant, Palisades, and 12 heat-up and cool-down cycles were used for the Babcock and Wilcox (B&W)-designed plant, OCI, to bound all the design basis transients for the respective PWR plant designs in each fleet. 1 RTAx-Fo means the material property which characterizes the RV's resistance to fracture initiation from flaws in forgings that are not associated with welds in the forgings. RTmA-Fo value is calculated under the provisions of Sections (f) and (g) of 10 CFR 50.61 a, Alternative fracture toucihness requirements for protection agqainst pressurized thermal shock, in Enclosure 1 to the. Proposed Rulemaking in SECY-07-0104 (Reference 12). WCAP- 16168-NP-A June 2008 Revision 2

Xi In response to RAt 1 from Reference 3, the PWROG provided a description of the analyses performed to determine whether the seven heat-up and cool-down cycles per year for Westinghouse plants and the 13 heat-up and cool-down cycles per year for CE plants bound all the design basis transients for the respective PWR Nuclear Steam Supply System (NSSS) designs in each fleet For Westinghouse plants, previous fatigue crack growth analyses of flaws on tie inside surface of the RV had shown that only four transients result in measurable crack growth. Sensitivity studies for the four contributing transients were performed. These analyses indicated that the only design transient that resulted in significant crack growth was the cool-down transient. The design basis for the Westinghouse plant was based on five cool-down cycles per year. An additional two cycles per year were added to the analysis to envelope the contribution of the other three transients which contributed to measurable fatigue crack growth. Previous fatigue growth studies were not available for CE-designed plants. Therefore, all design transients were evaluated in the CE transient fatigue crack growth sensitivity study. This study indicated that the cool-down transient produced the largest amount of fatigue growth for a RV inside surface flaw. The loss of secondary pressure transient also produced measurable growth. Assuming 12 cool-down cycles per year was considered to be conservative in comparison to the actual number of cool-downs a plant might experience in a given year based on plant operating experience. One additional cool-down cycle was added to the analysis to envelope the contribution to fatigue crack growth of the loss of secondary pressure transient. Based on the results of the fatigue crack growth sensitivity studies, the number of cool-down transients assumed for the Westinghouse and CE-designed pilot plants will envelope the fatigue crack growth from all Westinghouse and CE NSSS design transients. All RVs are inspected before operation providing confidence that there are no large flaws throughout the RV that have a high likelihood of failure given a PTS event. Only surface-breaking flaws are assumed to grow

     *from fatigue crack growth.

Fatigue crack growth sensitivity studies were not performed to determine the effect of B&W design transients for fatigue crack growth in B&W designed plants. Therefore, any B&W plant

      -licensee using the results of the TR to extend the RV ISI interval from 10 to 20 years, including the pilot plant, must demonstrate that the assumption of 12 heat-up/cool-down transients per

.year in the TR analysis bounds the fatigue crack growth for all design basis transients for that

,unit.

For the purpose of the pilot plant studies in the TR, an 80-year life for fatigue crack growth was used. This 80-year life envelopes plants seeking to obtain license extensions to 60 years and provides an additional margin of conservatism. This result in a total of 560 heat-up/cool-down transients for the Westinghouse-designed unit, 1040 heat-up/cool-down transients for the CE-designed unit, and 960 heat-up/cool-down transients for the B&W-designed unit. The PWROG indicated that most plants operational histories indicate that they will not reach this number of design transients by end of 80 years of operation. Hence, this calculation was performed as a bounding analysis based on actual plant operating histories. In response to RAI 1 from Reference 3, the PWROG indicated that the fatigue crack growth rates that are used in the fatigue crack growth analysis are taken from Section 4.2.2 of the Theoretical and Users Manual for PC-PRAISE (Reference 14). As noted in this report, these "equations provide a probabilistic representation of the fatigue growth relationship for fermtic materials in water contained in Appendix A of Section XI of the ASME Boiler and Pressure WCAP-16168-NP-A June 2008 Revision 2

xii Vessel Code." Figure A-4300-2, "Reference Fatigue Crack Growth Curves for Carbon and Low Alloy Feritic Steels Exposed to Water Environments," from Appendix A to Section XI in the current edition of the ASME Code, provides a graphical representation of these equations. It should be noted that the fatigue crack growth curves in Appendix A of Section XI of the ASME Code have not changed since they were originally included in the 1978 Edition of Section XI. Since the crack growth rate code used in the PWROG analysis was taken directly from a code that was previously reviewed and approved by the NRC staff in Reference 13 and is based on the ASME Code crack growth rate curves, the crack growth rate code used in the PWROG analysis is acceptable. Effectiveness of ISI To determine the impact.of different inspection intervals on the frequency of RV failure, the effectiveness of the ISI must be considered. The PWROG considered the impact of the probability of detection (POD) of flaws when ultrasonic inspection is performed on the RV welds and adjacent base metal. The basis for the POD used in the pilot plant studies for the RV ISI interval extension was taken from studies performed at the Electric Power Research Institute (EPRI) Nondestructive Examination (NDE) Center on the detection and sizing qualification of ISIs of the RV bettline welds (Reference 15). Figure 3-4 in the TR illustrates the POD as a

• function of flaw size. The POD ranges from 0.5 for very small flaws up to 0.9 and greater for

       *flaws with through-wall depths greater than 0.25 inches.

For the pilot plant evaluations, ultrasonic examinations were assumed to be conducted in accordance with ASME Code, Section Xl, Appendix VIII. Flaws that were detected were assumed to be repaired with the repaired area returned to a flaw-free condition. If the quality of inspection is not as good as assumed or the quality of the repair is less than 100 percent, then the result would be fewer flaws found and fewer flaws removed during repair, resulting in less difference in risk from one inspection interval to another. The POD values used in the analysis were relatively high and, therefore, the pilot plant studies conservatively calculated a larger potential difference in risk by maximizing the benefits of inspection. Material Fracture Toughness and Neutron Embrittlement The RV material properties for each of the pilot plant studies used plant-specific properties that are identified In Appendices B, F, and J in the TR. These material properties are input to the Fracture Analysis of Vessels - Oak Ridge (FAVOR) Code (Reference 16). The FAVOR Code, which was developed by Oak Ridge National Laboratory (ORNL) to perform PFM analyses for the NRC PTS Risk Studies, includes fracture toughness models which are based on extended databases of empirically obtained plane strain fracture toughness (Kk) and crack arrest fracture toughness (Ki1) data points and include the effects of statistical bias for direct measurement of fracture toughness. The input to the FAVOR Code includes plant-specific neutron fluence maps for each of the pilot plants. For the pilot plant evaluations in the TR, the input neutron fluence distributions were taken directly from the NRC PTS Risk Study. A series of neutron transport calculations were performed for the NRC PTS Risk Study to determine the neutron fluence on the inner wall of the pilot plant RVs. The modeling procedures were based on the guidance contained in RG 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence" (Reference 17). The models incorporated pilot plant-specific geometry and operating data. The WCAP- 16168-NP-A June 2008 Revision 2

Xiii neutron fluence for energies greater than one million electron volts (E > 1MeV) was calculated as a function of the azimuthal and axial location on the inner wall of the RV. The neutron fluence was extrapolated from the current state point to various EFPY of operation assuming a linear extrapolation of the most recent operating cycles. The neutron fluence values used in the RV ISI interval extension evaluations were for 60 EFPY for BV1 and Palisades and were for 500 EFPY for OCI. 500 EFPY were used for OCl rather than 60 EFPY to envelope license extension consideration and because it is recognized that OC1 is not the most radiation sensitive RV in the B&W fleet. The use of 500 EFPY for OCI should bound the embrittlement of the most highly embrittled RV in the B&W fleet. Accident Transients PTS events are viewed as providing the greatest challenge to PWR RV structural integrity. If a RV had an existing flaw of critical size and certain PTS transients were to occur, this flaw could rapidly propagate through the RV wall, resulting in a through-wall crack and challenging the integrity of the RV. The PTS Risk Study utilized plant-specific probabilistic risk assessment (PRA) models to determine the possible sequences which could result in a PTS event for each of the pilot plants. Due to the large number of sequences which were identified, it was necessary to group (i.e., bin) sequences with like characteristics into representative transients (PTS transients) that are analyzed using thermal-hydraulic (TH) codes. TH analyses were performed for each PTS transient to develop time histories of temperature, pressure, and heat transfer coefficients. These histories were then input into the FAVOR code where they were used during the calculation of the conditional probability of RV failure for each PTS transient. From this analysis, it was determined that only a portion of the PTS transients contribute to the total risk of RV failure, while the remaining transients have an insignificant or zero contribution. The transients which were identified to be contributors to PTS risk were then used for the PFM analysis in the PTS study and for the pilot plant studies in the TR. Stresses Resulting from PTS Transients, Cladding and Welding For each PTS transient, deterministic calculations were performed to produce a load definition input file that includes time-dependent, through-wall temperature profiles, through-wall circumferential and axial stress profiles, and stress intensity factors for a range of axially and circumferentially-oriented embedded and inner surface-breaking flaw geometries. This load definition file was input into the FAVOR code to produce the conditional probability of failure (CPF) (i.e., the conditional probability of a through-wall crack) for each PTS transient. These probabilities estimated by the FAVOR code (complete with uncertainties) are conditional in the sense that, within the FAVOR code probabilistic fracture mechanics module (FAVPFM), the TH transients are assumed to occur. In addition to the stress resulting from PTS transients, the PWROG analysis included the impact of cladding and residual stresses on the probability of failure. The pilot plant studies for RV ISI interval extension used a residual weld stress distribution through the wall that was taken from the NRC PTS Risk Study and is described in the FAVOR Code Theory Manual (Reference 16). The cladding stress used in the pilot plant studies was taken from the NRC PTS Risk Study. The cladding temperature dependence due to differential thermal expansion was based on a stress free temperature of 488 'F, which is consistent with that used in the NRC PTS Risk Study. WCAP-16168-NP-A June 2008 Revision 2

.Xiv Staff Evaluation of Engineering Considerations in PFM Analysis The material fracture toughness, neutron embrittlement, distribution and uncertainties in embedded and surface-breaking flaws, accident transients, frequency of transients, and stress resulting from PTS transients, cladding, and welding used-in the PWROG ISI interval extension study are acceptable because the values and methodologies were derived from the NRC PTS Risk Studies. The fatigue crack growth analysis used in the PWROG IS! interval extension study is acceptable because it was performed using a code approved by the NRC and has considered all sources of fatigue stress and the probability for preexisting fabrication flaws. The effectiveness of ISI has been adequately determined because it used data from studies performed at the EPRI NDE Center on the detection and sizing qualification of ISIs of RV beltline welds. Based on the above conclusions, the NRC staff considers that the PWROG has adequately considered the engineering variables in determining the risk of RV failure in its ISI interval extension study. The PWROG has identified two items that must be further evaluated. They are:

1) Licensees for B&W plants using the results of TR WCAP-16168-NP, Revision 2 to extend the RV ISl interval from 10 to 20 years (including the pilot plant) must demonstrate that the assumption of 12 heat-up/cool-down transients per year in the TR analysis bounds the fatigue crack growth for all design basis transients for that unit.

2) RVs with RTt*,-Fo values exceeding 240 OF require further evaluation because the analyses performed in TR WCAP-16168-NP, Revision 2 are not applicable.

3.2.2 Probabilistic Risk Assessment PTS events were viewed as providing the greatest challenge to PWR RV structural integrity and, therefore, the. PRA had to estimate the frequency and severity of PTS transients. PTS transients are not normally modeled in PRAs and the analyses of the pilot plants in the TR used the PTS transients and frequencies from the NRC PTS Risk Study. As part of the NRC PTS Risk Study, PRA models were developed by the NRC staff for each of the three pilot plants using plant-specific information (References 18, 19, and 20). These three units included one unit from each of the PWR vendors. These PRA models included an event tree analysis that defined the sequences of events that are likely to produce a PTS challenge to RV structural integrity for each of the pilot plants. As discussed above, individual event tree sequences with like characteristics were binned into representative PTS transients. The results of the PRA in the PTS Risk Study included descriptions of each PTS transient from, which the TH characteristics of each transient can be developed, and estimates of the frequency with which each transient was expected to occur. The final transient frequency estimates were distributions (histograms) which represented the combined frequency, including uncertainties, of all the event tree sequences incorporated into each bin. Appendices D, H, and L in the TR briefly described the failures and the mean estimated frequency for each bin for each of the three pilot plants. The transient frequencies were input into the FAVPOST module, the final module in the FAVOR Code. This module combined the conditional initiation and through-wall cracking probabilities WCAP- 16168-NP-A June 2008 Revision 2

Xv through a matrix multiplication with the frequency histograms for each PTS transient provided by the PRA analyses. 3.2.2.1 Estimating the Risk Associated with Extending the RV Weld Inspection Interval from 10 to 20 Years The likelihood of RV failure was postulated to increase with increasing time of operation due to the growth of pre-existing fabrication flaws by fatigue in combination with a decrease in RV toughness due to irradiation. The PFM approach in the TR simulated the growth of flaws over time and the repair of flaws that are detected during a periodic ISI. The largest cracks were expected to exist at the end of the plant's operating life because, even with periodic inspection, flaws may be missed during an inspection. These flaws would remain in service and grow until eventually detected by ISI, causing RV failure during a PTS event, or the end of plant life is reached. The end of operating life is also the time when the RV will be most embrittled and most subject to failure for any size crack. Therefore, instead of assuming that PTS transients can occur randomly during the operating life, the PWROG's response to RAI 9 from Reference 3 explained that the TR conservatively estimated the CPF for each PTS transient by applying the PTS loadings to the material properties and the distribution of flaws sizes expected to exist on the first day of full power operation following the refueling outage after the last operating year of the extended license of the plant. The NRC staff concurred that this process approximates the greatest CPF expected to exist during the life of the plant, The PTS transients' frequencies were not expected to change over the plant life so the product of these frequencies with the maximum CPF is acceptable because it results in a bounding estimate for the TWCF and associated increase in risk. The current inspection interval is 10 years and the base case scenario for the change in risk

       *analysis is one inspection every 10 years. Rather than evaluate each plants' specific inspection cycle, the TR bounded the impact of extending the interval by estimating the risk increase as the difference between the base case risk (assuming that the RV was inspected every ten years)
       -and the risk assuming that a plant only had one inspection after the first 10 years and then was
       *never inspected again for the remaining life of the plant Plant life was assumed to be 80 years, 2Jfor both the base case (every 10 year inspection) and the bounding case (only one inspection).

The NRC staff concurred that this evaluation is applicable to all plants and the change in risk estimated for this scenario will bound the change expected by extending the 10 year interval to a

       .20 year interval.

The TR assumed that a through-wall crack will lead to core damage and. that core damage will

• lead to a large early release. The RG 1.174 guideline addressing an acceptable increase in large early release frequency (LERF) is the smallest guideline value. Requiring that the TWCF is less than the LERF guideline ensured that both the core damage frequency (CDF) and LERF guidelines are met. The equation in FAVPOST that was used to estimate risk with and without periodic inspection for plant j is; LERFj = CDF1 TWCF1 = Z lEp
• CPFV

       *where, WCAP- 16168-NP-A                                                                                            June 2008 Revision 2

xvi lEp is the initiating event frequency (events per year) for each of the i representative PTS transients for plant j developed during the PTS Risk Study. The PTS Risk Study developed full distributions for the frequency of each PTS transient bin and the TR used the full distribution. 2 IEi does not change when the inspection period changes. CPFp is the conditional probability of RV vessel failure (conservatively assumed to occur if a through-wall crack develops) given the thermal-hydraulic characteristics of each of the i representative PTS transients for plant j. As described above, the RV material properties and the distribution of flaw sizes are those expected to exist at the end of plant j's operating life. The distribution of flaw sizes is the parameter that changes when the inspection period changes and, therefore, CPFi changes when the inspection period changes. The NRC staff concurs that the PRA models of PTS transient frequency, the lEp and CPFj, parameters, and the above equation appropriately capture the significant contributors to risk from RV failure and, therefore, fulfill the RG 1.174 guidance that the analysis is capable of modeling the impact of the proposed change. The NRC staff also concurs that the bounding estimates from only one inspection versus an inspection every ten years appropriately envelops the impact of the proposed change for any facility regardless of its inspections schedule and history, IS[ is directed toward identifying surface-breaking and embedded flaws that have grown large enough to require repair. In the response to RAI 12a from Reference 3, the PWROG noted that the frequency of surface-breaking flaws should be very small because none had ever been discovered during either pre-service or in-service examinations of beitline welds. With few such flaws, few failures were observed from the simulations even when fatigue crack growth was included. With few failures, it was difficult to obtain a converged solution using Monte Carlo simulation in the FAVOR Code because its precision is based upon the number of failures in the total number of simulations. In order to obtain a converged solution, the dominant contribution to TWCF from embedded flaws was included3 in the simulations. The result of including the dominant contribution from embedded flaws in the simulation was that direct comparison of the mean TWCF with only one inspection and the mean TWCF with inspections every ten years did not produce a stable metric. This is illustrated by, for example, the results in Table 4-1 in the TR which reported that the estimated TWCF for BVI with only one inspection (5.04E-9/year) was smaller than the TWCF with one inspection every ten years (5.23E-9/year) although the more frequent inspection program should result in a smaller TWCF, In the response to RAI 12b from Reference 3, the PWROG, reported on a sensitivity study that was performed by running the Monte Carlo simulation without the embedded flaws. The PWROG reported that the number of FAVOR simulations was increased from 70,000 to 500,000 but that no failures were obtained for both the only one inspection and the inspection every ten 2 Appendices D, H, and L include only the mean frequency estimates from the PTS transient bins, but the calculations illustrated in Appendices E, I, and M are performed using the full initiating event frequency distributions. 3 The NRC staff concluded during the PTS Risk Study, that embedded flaws do not grow over time and therefore their contribution to TWCF is driven by the initial flaw distribution and is unaffected by the ISI interval. WCAP- 16168-NP-A June 2008 Revision 2

Xvii years simulations. The PWROG noted that excluding embedded flaws results in a zero TWCF for both inspection intervals and, therefore, a zero increase In TWCF given the proposed interval extension. Because of the uncertainty in how accurately an insignificant (null) effect can be calculated using standard Monte Carlo simulation, the PWROG included embedded flaws and estimated the change in risk by subtracting the lower bound mean estimate for one inspection every ten years from the upper bound mean estimate for only one inspection.. The PWROG argued that this difference represents the maximum statistically calculated value for the potential change in risk at a number of RV simulations for which the Monte Carlo statistical analysis has reached a stable solution. In its response to RAI 12c from Reference 3, the PWROG described the derivation of the standard error on the mean which was used to calculate the upper and lower bound estimates. The standard error is a statistical estimate reflecting how much sampling fluctuation was observed which can be used to estimate confidence intervals about the mean estimate. The PWROG chose to use two times the standard error to develop its confidence bounds. Therefore, if repetitive simulations (each with 70,000 trials) were performed, it is expect that in only 2,5% of the mean estimates would exceed the upper bound value and 2.5% would be less than the lower bound value. The NRC staff concluded that the analyses described in the TR provided a reasonable or. bounding estimate of the increase'in risk associated with extending the inspection interval for RV welds from 10 to 20 years. As discussed above, the NRC staff based this conclusion on: the PRA models of PTS transient frequency, the IEj1 and CPFf parameters, and the equation used to calculate the risk from PTS events appropriately capturing the significantcontributors to risk from RV failure,

• the bounding estimates from only one inspection versus an inspection every ten years appropriately modeling the impact of the proposed change for any facility regardless of its RV inspections schedule and history,
• the TWCF from surface-breaking flaws being so small that the Monte Carlo estimation techniques in the FAVOR code do not converge to a stable solution indicating that the TWCF from surface-breaking flaws is small regardless of the inspection program interval, and the subtraction of the lower bound mean estimate for one inspection every ten years from the upper bound mean estimate for only one inspection being consistent with the guidance in RG 1.174 that the difference in the means (in this case confidence estimates on the means) is the risk metric that should be compared with the acceptance guidelines.

3.2.2.2 Evaluation of PRA Technical Adequacy Technically adequate is defined, at the highest level, as an analysis that is performed correctly, in a manner consistent with accepted practices, commensurate with the scope and level of detail required to support the proposed change. The PWROG used the PTS transient frequencies developed in the NRC PTS Risk Study in its analysis. The TR conservatively assumed that core damage and large eariy release will inevitably follow a PTS transient that results in a WCAP- 16168-NP-A June 2008 Revision 2

Xvill through-wall crack. Therefore, there is no PRA event and sequence modeling needed beyond the determination of the PTS transient frequencies. The NRC staff developed plant-specific PRA analyses to estimate the PTS transient frequencies for each of the three pilot plants using a process described in detail in NUREGICR-6859, "PRA Procedures and Uncertainty for PTS Analysis" (Reference 21). The analyses were described in detail in the plant-specific PRA reports (References 18, 19, and 20) and summarized in Chapter 5 of the PTS Risk Study. The process included a review of the PRA analyses performed during the 1980s in support of the first PTS rule and a search of licensee event

• reports for the years 1980 through 2000 to gain an understanding of the frequency and severity of observed overcooling events. The PRA analyses used realistic input values and models and an explicit treatment of uncertainties. Best estimate equipment failure values were used throughout based on generic nuclear industry data or, in cases where it Was available, on plant-specific data. Parameters related to human performance were based on review of plant-specific procedures and training, observation of plant personnel responding to PTS-related sequences on their simulator, and performance data from actual plant operations. The scope of the study covered all event sequences in the range from zero power hot stand-by up to 100% power.

As discussed in the individual pilot plants' PRA reports, all analyses were conducted through plant visits and by numerous interactions (vocal, written, and e-mail exchanges) with each licensee as the analysis evolved. During a first site visit, the PTS study team collected information. After preliminary results were completed; reviews were performed both by licensee and NRC project staff during a second site visit. The OCI and BV1 models used system level fault trees and system level failure data. The Palisades model used detailed system level fault trees from the licensee's PRA. Formal reviews were carried out for OC1 and BV1. Palisades' models were developed by the licensee and reviewed by the NRC staff. A final peer review was carried out by a panel of six experts to provide an independent review of the technical basis developed for the PTS Rulemaking (Reference, 6), The objective of the peer review was to assess the adequacy and reasonableness of the technical basis to support the proposed revision of the PTS rule. The peer reviewers focused on different parts of the PTS analysis. Comments related to the PRA aspects generally concluded that the work was well founded and reasonable and no serious weaknesses were identified. Based on the PTS Risk Study's detailed review of past studies and operating experience, extensive interactions between the analysis team and the plant personnel at all units, and the opportunity for the same. team to benefit from the multiple plant study insights while performing all the analyses, the NRC has confidence that the PTS transient frequency results from the PRA analyses in the PTS Risk Study are sufficiently well developed to be able to demonstrate that the change in risk estimates as developed in the TR does not exceed the acceptance guidelines in RG 1.174. 3.2.2.3 Generic Applicability and External Events During the development of the PTS Risk Study, the NRC staff investigated the applicability of the results from the three pilot plants to the operating fleet of PWRs. These three units included one unit from each of the three PWR vendors. This investigation examined plant design and operational characteristics of five additional plants as described in Letter Report, "Generalization of Plant-Specific Pressurized Thermal Shock (PTS) Risk Results to Additional Plants," WCAP-16168-NP-A June 2008 Revision 2

XiX (Reference 22). The overall approach was to compare potentially important design and operational features (as related to PTS) of the other PWRs to the same features of the pilot plants to determine the extent these features are similar or different. In 72 FR 56275 (Reference 4), the NRC staff reported its conclusion that the TWCF results from the PTS Risk Study can be applied to the entire fleet of operating PWRs. This conclusion was

     *based on an understanding of characteristics of the dominant transients that drive their risk significance. The generic evaluation revealed no design, operational, training, or procedural factors that could credibly increase the severity of these transients or the frequency of their occurrence in the general PWR population above the severity/frequency characteristics of the three plants that were modeled in detail. As applied to the analyses included in the TR, this conclusion indicated that the PTS transient frequencies and TH characteristics used to estimate the change in risk are dependent only on the reactor vendor and are generally applicable to all PWRs from that vendor.

The detailed plant-specific PRAs in the PTS Risk Study evaluated the contribution of internal initiating events to TWCF. The study group also evaluated the potential contribution of external initiating events to PTS risk as described in Reference 23 and summarized in Section 9.4 of the PTS Risk Study. The external events included in the evaluation were fires, floods, high winds and tornados, and seismic events. This analysis was structured by identifying three broad types of overcooling scenarios and making conservative judgments with regard to the type and frequency of external events that could directly contribute to causing each overcooling scenario. The conservative judgments were directed toward bounding the PTS TWCF contributions attributable to external events for the worst situation that might arise at virtually any plant. The study's results indicated that the bounding total external event TWCF is approximately 2E-8/year, quantitatively comparable to the highest' internal events contribution of 2E-8lyear. The study concluded that there was considerable assurance that the external event contribution to the overall TWCF as a result of external event initiated PTS events is at least no greater than the highest best estimate contribution from internal events. Based on the results of the PTS Generalization Study, the NRC staff has concluded that the PTS transient characteristics (both frequency and TH characteristics) are generically applicable for all similar plants (i.e., plants from the same vendor) in the fleet. Based on the results of the

      ;external events analyses, the NRC staff has also concluded that the contribution of external events to the change in risk has been adequately evaluated and that the contribution to risk from external events is equal or less than the contribution for internal events.

3.2.2.4 Comparison with RG 1.174 Acceptance Guidelines The results of the change in risk analyses were summarized in Table 4-1 in the TR where the bounding increases in risk were reported as 9.37E-1 0/year, 1.81 E-8/year, and 1.26E-8/year for BV1 (Westinghouse-designed plant), Palisades (CE-designed plant), and OCl (B&W-designed plant), respectively. These increases are well below the guideline for a very small increase in LERF of IE-7/year in RG 1.174. The TR only incorporated the Internal events PTS sequence frequency results from the PTS rulemaking into its change in risk analysis. The largest increase in LERF was estimated as 1.8E-8/year for the Palisades plant. The NRC staffs evaluation of external event contributions to PTS risk determined that the total PTS risk would, at most, double compared to the risk from WCAP-16168-NP-A June 2008 Revision 2

XX internal events when the risk from external events are included. Since the total risk for the base case and the only one inspection case would both double, the total change in risk would also double. The NRC staff concluded that the greatest change in risk associated with extending the inspection interval at any PWR using the methods and guidelines described in the TR and endorsed in this SE is less than 5E-B/year. The NRC staff finds that this increase is small and consistent with the intent of the Commission's safety goals. 3.3 Implementation and Monitoring The third element in the RG 1.174 approach is to develop an implementation and monitoring program to ensure that no adverse safety degradation occurs because of the proposed changes. Therefore, an implementation and monitoring plan should be developed to ensure that the engineering evaluation conducted to examine the impact of the proposed changes continues to be valid after the change has been implemented. This will ensure that the conclusions that have been drawn from the evaluation remain valid. RV integrity depends upon licensees ensuring that the critical elements of the PFM analysis described in the TR are valid, Licensees must monitor the number of cycles of transients that could affect the fatigue crack growth analysis, the change in fracture toughness of the limiting RV material due to exposure to radiation, and the flaw distribution in the RV welds and adjacent base metal. The number of transient cycles that were utilized in the fatigue crack growth analysis was discussed in Section 3.2.1 of this SE. The PWROG used 7 heat-up and cooldown cycles per year for Westinghouse-designed plants, 13 heat-up and cooldown cycles per year for CE-designed plants, and 12 heat-up and cooldown cycles per year for B&W-designed plants. The design basis for the Westinghouse plant was 5 cooldown cycles per year. Although it was determined that three other transients did not significantly contribute to fatigue crack growth in RV welds, an additional 2 cycles were conservatively added to envelope the contribution of

       *these three transients. Since the PWROG fatigue crack growth analysis for Westinghouse NSSS designed plants determined that the only design basis transient that resulted in significant crack growth was the cool-down transient, it is the only design basis transientthat needs to be monitored. Since the PWROG fatigue crack growth analysis of CE NSSS designed plants determined that the amount of crack growth from 13 cool-down transients bounds the expected
        *crack growth from both cool-down and loss of secondary pressure transients, CE plants should monitor the number of cool-down transients. Fatigue crack growth sensitivity studies were not performed to determine the effect of B&W design transient for fatigue crack growth in B&W designed plants. Therefore, any B&W plant using the results of the TR to extend the RV ISI interval from 10 to 20 years (including the pilot plants), must determine the design basis transients that contribute to significant crack growth in RV welds. These transients must be monitored by the licensee.

Material fracture toughness was discussed in Section 3.2.1 of this SE and must be monitored by determining whether the 951h percentile TWCFTOTAL4 for the plant requesting to implement the pilot plantstudy is less than the 95Qh percentile TWCFTOTAL from the pilot plant study. The 95'h 4 The 95 percentile TWCFToTAL is the sum of the 95 percentile TWCF for all beltline materials. It is calculated in accordance with NUREG-1874. WCAP- 16168-NP-A June 2008 Revision 2

xxi percentile TWCFTOTAL was calculated based on the material properly indexing parameter RTi4Ax.x5 Appendix A in the TR identifies the 95Vh percentile TWCF7GTAL from the pilot plant studies for BVI, Palisades, and OCI. The 950 percentile TWCFTotAL value calculated for BVI at 60 EFPY was 1.76E-08 events per year. The 950 percentile TWCFToT* value calculated for Palisades at 60 EFPY was 3.16E-07 events per year. The 95"' percentile TWCFTOTAL value calculated for OCi at 500 EFPY was 4.42E-07 events per year. The flaw distributions used in the PWROG PFM analyses are described in Section 3.2.1 of this SE. The PWROG utilized the flaw sizes and distributions in the NRC PTS Risk Study to simulate embedded flaws in welds, forgings, and plates and to simulate the initial size and distribution of surface-breaking flaws. Section (e) of the proposed 10 CFR 50.61a, Alternative fracture toughness requirements for protection against pressurized thermal shock, in Enclosure I to the proposed rulemaking in SECY-07-0104 described the allowable flaw distribution for embedded flaws and surface-breaking flaws that would be permitted, for RVs that are at the PTS screening limits described in the proposed 10 CFR 50.61a. By monitoring flaw sizes in accordance with the criteria described in Section (e) of the final 10 CFR 50,61a (or the proposed 10 CFR 50.61a, given in 72 FR 56275 prior to issuance of the final 10 CFR 50.61a) licensees will ensure that their RVs do not have flaws that invalidate the results of the PWROG PFM analyses. The NRC staff concludes that the implementation and monitoring described above will ensure that the conclusions that have been drawn from the evaluation remain valid. 3.4 Submit Proposed Change The fourth and final element in RG 1.174 approach is the development and submittal of the proposed change to the NRC. Since the 10 year ISI interval is required by Section XI, IWB-2412, as codified in 10 CFR 50.55a. a relief for an alternative, in accordance 10 CFR 50.55a(aX3)(i), must be submitted and approved by the NRC to extend the ISI interval. Licensees that submit a request for an alternative based on the TR need to submit the following plant-specific information:

     -1)       Licensees must demonstrate that the embrittlement of their RV is within the envelope used in the supporting analyses, Licensees must provide the 9 5t" percentile TWCFToTAL, and its supporting material properties at the end of the period in which the relief is requested to extend the inspection interval from 10 to 20 years. The 95t' percentile TWCFToTAL must be calculated using the methodology in NUREG-1 874. The RT~x-yx and the shift in the Charpy transition temperature produced by irradiation defined at the 30 ft-lb energy level, ATso, must be calculated using* the latest approved methodology documented in Regulatory Guide 1.99, "Radiation Embrittlement of Reactor Vessel Materials," or other NRC-approved methodology. The PWROG response to RAI 3 from Reference 3 and Appendix A in the TR identifies the information that is to be submitted.

5 RT.*x-x values are determined for each beltline material. RTx-x is a material property which characterizes the RVs resistance to fracture initiating from flaws in welds, plates, and forgings. The method of determining RTmvx.x is described in Sections (f) and (g) of 10 CFR 50.61 a, Alternative fracture toughness requirements for protection against pressurized thermal shock, in Enclosure 1 to the Proposed Rulemaking in SECY-07-0104. WCAP-16168-NP-A June 2008 Revision 2

Xxii

2) Licensees must report whether the frequency of the limiting design basis transients during prior plant operation are less than the frequency of the design basis transients identified in the PWROG fatigue analysis that are considered to significantly contribute to fatigue crack growth.

3) Licensees must report the results of prior ISI of RV welds and the proposed schedule for the next 20 year IS] interval. The 20 year inspection interval is a maximum interval. In its request for an alternative, each licensee shall identify the years in which future inspections will be performed. The dates provided must be within plus or minus one refueling cycle of the dates identified in the implementation plan provided to the NRC in PWROG letter OG-06-356. "Plan for Plant Specific Implementation of Extended Inservice Inspection Interval per WCAP 16168-NP, Revision 1, "Risk Informed Extension of the Reactor Vessel In-Service Inspection Interval," MUHP 5097-99, Task 2059," dated October 31, 2006 (Reference 10).

4) Licensees with B&W plants must (a) verify that the fatigue crack growth of 12 heat-up/cool-down transients per year that was used in the PWROG fatigue analysis bound the fatigue crack growth for all of its design basis transients and (b) identify the design bases transients that contribute to significant fatigue crack growth.

5) Licensees with RVs having forgings that are susceptible to underclad cracking and with RTWX.Fo values exceeding 240 OF must submit a plant-specific evaluation to extend the inspection interval for ASME Code, Section Xl, Category B-A and B-D RV welds from 10 to a maximum of 20 years because the analyses performed in the TR are not be applicable.

Within one year of completing each of the ASME Code, Section XI, Category B-A and B-D RV weld inspections required in the proposed ISI interval, the licensee must provide the information and analyses requested in Section (e) of the final 10 CFR 50.61a (or Alternative fracture toughness requirements for protection aqainst pressurized thermal shock, in Enclosure 1 to the proposed rulemaking in SECY-07-0104, Reference 12, given in 72 FR 56275 prior to issuance of the final 10 CFR 50,61a). Licensees that do not implement 10 CFR 50.61a must amend their licenses to require that the information and analyses requested in Section (e) of the final 10 CFR 50.61a (or the proposed 10 CFR 50.61a, given in 72 FR 56275 prior to issuance of the final 10 CFR 50,61 a) will be submitted for NRC staff review and approval. The amendment to the license shall be submitted at the same time as the request for alternative. Licensees that implement 10 CFR 50.61a must perform the ISts required in Section (e) of the rule and must submit the required information for review and approval to the Director, Office of Nuclear Reactor Regulation, in accordance with Section (c) of the rule, at least three years before the limiting RTPTs value calculated under 10 CFR 50.61 is projected to exceed the PTS screening criteria in 10 CFR 50.61. Licensees implementing Section (c) of 10 CFR 50.61a must perform the inspections and analyses required by Section (e) of 10 CFR 50.61a prior to implementing the extended interval. WCAP- 16168-NP-A June 2008 Revision 2

Xxiii 3.5 Conformance to RG 1.174 In addition to the four element approach discussed above, RG 1.174 states that risk-informed plant changes are expected to meet a set of key principles. This section summarizes these principles and the NRC staff findings related to the conformance of the TR methodology with these principles. Principle 1 states that the proposed change must meet the current regulations unless it is explicitly related to a requested exemption or rule change. ISI of ASME Code Class 1, 2, and 3 components is performed in accordance with Section Xl of the ASME Code and applicable addenda as required by 10 CFR 50.55a(g), except where specific relief has been granted by the NRC pursuant to 10 CFR 50.55a(g)(6Xi). This risk-informed application requires a request for an alternative under CFR 50.55a(a)(3)(i) which meets the current regulations and, therefore, satisfies Principle 1. Principle 2 states that the proposed change shall be consistent with the defense-in-depth philosophy. In the response to RAI 11 a from Reference 3, the PWROG argued that the proposed change is consistent with the defense-in-depth philosophy because there is no change in RV design and no change in the robustness of the RV or other systems at the plant. The NRC staff believes that ISI is an integral part of defense-in-depth and extending the interval may change the robustness of the RV, albeit very slightly. However, the extension of the inspection interval is accompanied by various evaluations and a monitoring program, and the NRC staff concludes that, in total, the proposed ISI program provides reasonable assurance that RV integrity will be maintained consistent with the philosophy of defense-in-depth. Therefore, Principle 2 is met. Principle 3 states that the proposed change shall maintain sufficient safety margins. Section 12 in PTS Risk Study concluded that the calculations demonstrate that PTS events are associated with an extremely small risk of RV failure, suggesting the existence of considerable safety margin. Section 4.3 in the TR clarified that no safety analysis margins are changed and, aside from extending the inspection interval, no portions of the current inspection requirements are eliminated. The NRC staff concurred that the proposed change maintains sufficient safety margins because the change simply extends the inspection interval and does not change, for

       .example, the acceptance criteria used to determine whether any identified flaws are acceptable or need to be repaired. Therefore, Principle 3 is met.

Principle 4 states that when proposed changes result in an increase in CDF or risk, the increases should be small and consistent with the intent of the Commission's Safety Goals, The NRC staff concluded that the greatest increase in LERF associated with extending the inspection interval at any PWR using the methods and guidelines described in the TR and endorsed in this SE is less than 5E-8/year. The NRC staff found that this increase is small and consistent with the intent of the Commission's Safety Goals. Therefore, Principle 4 is met. Principle 5 states that the impact of the proposed change should be monitored using performance measurement strategies. As described in Section 3.3 of this SE, licensees must monitor the number of cycles of transients that could effect the fatigue crack growth analysis, the fracture toughness of the limiting RV material, and the flaw distribution in the RV welds and adjacent base metal. The NRC staff found that the planned monitoring program provides confidence that no adverse safety degradation will occur because of the proposed changes and WCAP- 16168-NP-A June 2008 Revision 2

Xxiv that the engineering evaluation conducted to examine the impact of the proposed changes will continue to be valid after the change has been implemented. Therefore, Principle 5 is met. 3.6 NRC Staff Findings The NRC recently proposed a new rulemaking (72 FR 56275) which would change the regulations regarding the requirements for protection against PTS events. In support of this rulemaking, the NRC staff concluded that the risk of through-wall cracking caused by PTS events is much lower than previously estimated. The proposed rule provided new PTS screening criteria that are selected based on an evaluation that indicated that, after applying these new, relaxed criteria, the risk of through-wall cracking due to a PTS event at any PWR would be less than 1E-6tyear. Most PWRs are not expected to need the new screening criteria and, therefore, would have a TWCF less than, or substantially less than, 1 E-6/year. The analysis developed to support this TR uses mostly the same inputs and models used in the PTS Risk Study. The PTS Rfsk Study concluded that embedded flaws do not grow and, therefore, after the first inspection, periodic ISIs do not affect the risk from embedded cracks. Surface cracks that penetrate through the cladding and into the ferritic alloy steel were not part of the PTS Risk Study because these types of flaws have not been observed in the beltline of operating PWR reactors. PFM analyses indicate, however, that surface cracks can grow over time when subject to fatigue. The TR has analyzed the growth of postulated surface cracks because extending the RV inspection interval could increase the risk of RV failure from such cracks. The NRC staff has concluded that the TR has appropriately postulated and modeled the potential change in risk that could be caused by fatigue crack growth over the life of operating facilities. Based on the results of the PTS Generalization Study, the NRC staff has concluded that the PTS transient characteristics (both frequency and TH characteristics) are generically applicable' for plants from the same reactor vendor. RV embrittlement is, however, RV material, operating history, and age specific. Therefore, the NRC staff found that, while the PTS transient work need not be repeated by each plant seeking to extend its interval, the analyses and monitoring to demonstrate that the RV embrittlement is within the envelope used in the supporting analyses and must be performed by each plant as described. The NRC staff found that licensees implementing the ISI interval extension program documented in the TR and endorsed in the SE will have a program that meets the five key principles stated in RG 1.174 and, therefore, the proposed alternatives would provide an acceptable level of quality and safety, in accordance with 10 CFR 50.55a(aX3Xi). Based on the above conclusions, the ASME Code Section XI ISI interval for examination categories B-A and B-D welds in PWR RVs can be extended from 10 years to a maximum of 20 years. Since the 10 year ISI interval is required by Section XI, IWB-2412, as codified in 10 CFR 50.55a, a request for an alternative, in accordance 10 CFR 50.55a(gX6Xi), must be submitted and approved by the NRC to extend any facility's ISI interval. In addition, licensees that do not implement 10 CFR 50.61a must amend their licenses to require that the information and analyses requested in Section (e) of the final 10 CFR 50.61a (or the proposed 10 CFR 50.61a, given in 72 FR 56275 prior to issuance of the final 10 CFR 50.61a) will be submitted for NRC staff review and approval. The amendment to the license shall be submitted WCAP- 16168-NP-A June 2008 Revision 2

Xxv at the same time as the request for an alternative. The request for an alternative will be for the remainder of the licensed period for the plant. The methodology in the TR is applicable to all operating PWR plants by confirming the applicability of the parameters in Appendix A of the TR on a plant-specific basis. Licensees must submit a request for an alternative that contains all the information in Section 3.4 of this SE. However, since the analysis documented in the TR used plant-specific data for BV1, Palisades, and OC1, these plants need not confirm the applicability of the parameters in Appendix A of the TR for the current license term. The NRC staff will not repeat its review of the matters described in WCAP-16168-NP, Revision 2, as modified by this SE, when the report appears as a reference in a request for an alternative, except to ensure that the material presented applies to the specific plant involved and the licensee has submitted all the information requested in Section 3.4 of this SE. 4.0 CONDITIONS AND LIMITATIONS The 20 year inspection interval is a maximum interval. In its request for an alternative, each licensee shall identify the years in which future inspections will be performed. The dates provided must be within plus or minus one refueling cycle of the dates identified in the implementation plan provided to the NRC in PWROG letter OG-06-356, "Plan for Plant Specific Implementation of Extended Inservice Inspection Interval per WCAP 16168-NP, Revision 1, "Risk Informed Extension of the Reactor Vessel InwService Inspection Interval," MUHP 5097-99, Task 2059," dated October 31, 2006 (Reference 10). Within one year of completing each of the ASME Code, Section Xl, Category B-A and B-D RV weld inspections required in the proposed ISI interval, the licensee must provide the information and analyses requested in Section (e) of the final 10 CFR 50.61a (or Alternative fracture toughness requirements for protection against pressurized thermal shock, in Enclosure 1 to the proposed rulemaking in SECY-07-0104, Reference 12, given in 72 FR 56275 prior to issuance of the final 10 CFR 50.61a). Licensees that do not implement 10 CFR 50.61a must amend their licenses to require that the information and analyses requested in Section (e) of the final 10 CFR 50.61a (or the proposed 10 CFR 50.61a, given in 72 FR 56275 prior to issuance of the final 10 CFR 50.61a) will be submitted for NRC staff review and approval. The amendment to the license shall be submitted at the same time as the request for alternative. Licensees that implement 10 CFR 50.61a must perform the ISIs required in Section (e) of the rule and must submit the required information for review and approval to the Director, Office of Nuclear Reactor Regulation, in accordance with Section (c) of the rule, at least three years before the limiting RTpTs value calculated under 10 CFR 50.61 is projected to exceed the PTS screening criteria in 10 CFR 50.61. Licensees implementing Section (c) of 10 CFR 50.61a must perform the inspections and analyses required by Section (e) of 10 CFR 50.61a prior to implementing the extended interval. The methodology in the TR is applicable to all operating PWR plants by confirming the applicability of the parameters in Appendix A of the TR on a plant-specific basis. Licensees must submit a request for an alternative that contains all the information in Section 3.4 of this WCAP-16168-NP-A June 2008 Revision 2

xxvi SE. However, since the analysis documented in the TR used plant-specific data for BV1, Palisades, and OC, these plants need not confirm the applicability of the parameters in Appendix A of the TR for the current license term. The NRC staff has accepted TR WCAP-16168-NP, Revision 2, based on the imposition of a condition related to the augmented evaluation of in-service inspection (1Sl) results taken from Section (e) of the proposed Title 10 of the Code of FederalRegulations 50.61a, published in the Federal Register on October 3, 2007 (72 FR 56275). The NRC staff is in the process of reviewing public comments on the proposed rule and preparing the final rule. Ifthe final 10 CFR 50.61a differs from the proposed 10 CFR 50.61a with regard to the augmented ISI evaluation requirements, the PWROG will be expected to review the requirements in the final 10 CFR 50.61a and determine whether a revision to the accepted TR WCAP-16168-NP, Revision 2, is required. The PWROG will be expected to notify the NRC staff, in writing, of the results of its determination within six months of the publication date of the final 10 CFR 50.61 a. If,on this basis, a revision to the accepted TR WCAP-16168-NP, Revision 2, is required, the PWROG will be expected to submit the revised TR for NRC staff review within one year of the publication date of the final 10 CFR 50,61a.

5.0 CONCLUSION

The NRC staff has found that the methodology presented in WCAP-16168-NP, Revision 2, in concert with the guidance provided by RG 1.174, is acceptable for referencing in license amendment requests for PWR plants in accordance with the limitations and conditions in Section 4.0 of this SE. The NRC staff will consider extending the RV weld inspection interval beyond 10 years based on plant-specific requests for an alternative that reference WCAP-16168-NP, Revision 2.

6.0 REFERENCES

1. Letter from F. P. Schiffley, Westinghouse Owners' Group, "Transmittal of WCAP-161 68-NP, Revision 1, 'Risk-Informed Extension of Reactor Vessel In-Service Inspection Interval', MUHP-509715098l5099, Tasks 2008/2059," January 26, 2006 (ADAMS Accession No. ML060330504)

2. Letter from F. P. Schiffley, PWR Owners Group, "Evaluation of NRC Questions on the Technical Bases for Revision of the PTS Rule Relative to Their Effects on the Risk Results in WCAP-16168-NP, Revision 1, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval," June 8, 2006 (ADAMS Accession No. ML061600431 1)

3. Letter from F. P. Schiffley, PWR Owners Group, "Responses to the NRC Request for Additional Information (RAI) on PWR Owners' Group (PWROG) WCAP-16168-NP, Revision 1, 'Risk-Informed Extension of Reactor vessel In-Service Inspection Interval',

MUHP-5097/5098/5099, Tasks 2008/2059," October 16, 2007, and Enclosure 1, RAI responses (ADAMS Accession No. ML0729204120). Enclosure 2, WCAP-16168-NP, Revision 2, 'Risk-Informed Extension of Reactor vessel In-Service Inspection Interval', October 2007 (ADAMS Accession No. ML072920413). WCAP-16168-NP-A June 2008 Revision 2

Xxvii

4. Federal Register Notice, (72 FR 56275) "Alternative Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events," October 3, 2007 (ADAMS Accession No. ML072780354)

5. NUREG-1 806, "Technical Basis for Revision of the Pressurized Thermal Shock (PTS)

Screening Limit in the PTS Rule (10 CFR 50.61): Summary Report," August 2007 (ADAMS Accession Nos. ML072830076 and ML072830081)

6. NUREG-1806, "Technical Basis for Revision of the Pressurized Thermal Shock (PTS)

Screening Limit in the PTS Rule (10 CFR 50.61): Appendices," August 2007 (ADAMS Accession No. ML07282069)

7. NUREG-1 874, "Recommended Screening Limits for Pressurized Thermal Shock (PTS),

2007 (ADAMS Accession No. ML070860156)

8. U.S. NRC, NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," Section 19.2, "Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis: General Guidance,"

June, 2007 (ADAMS Accession No. ML071700658)

9. U.S. NRC, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed

             *Decisions on Plant-Specific Changes to the Licensing Basis," Regulatory Guide 1.174, Revision 1, November 2002 (Adams Accession No. ML023240437)

10. PWR Owners Group letter OG-06-356, "Plan for Plant Specific Implementation of Extended Inservice Inspection Interval per WCAP 16168-NP, Revision 1, "Risk Informed

• Extension of the Reactor Vessel In-Service Inspection Interval," MUHP 5097-99, Task 2059," dated October 31, 2006

11. NUREGICR-6817, Revision 1, "A Generalized Procedure for Generating Flaw-Related Inputs for the FAVOR Code," October 31, 2003 (ADAMS Accession No. ML051790410)

12. SEC-07-0104, "Proposed Rulemaking-Altemate Fracture Toughness Requirements For Protection Against Pressurized Thermal Shock Events," June 25, 2007 (ADAMS Accession No. ML070570525)

13. WCAP-14572, Revision 1-NP-A, Westinghouse Owners Group Application of Risk-Informed Methods to Piping lnservice Inspection Topical Report, February 1999 (ADAMS Accession Nos. ML042610469 and ML042610375)

14. Theoretical and Users Manual for PC-PRAISE, NUREGICR-5864, July 1992

15. Electric Power Research Institute (EPRI) Nondestructive Examination (NDE) Center on the detection and sizing qualification of IS!s on the RV beltline welds

16. Letter Report, Oak Ridge National Laboratories/TM-2007/0030, "Fracture Analysis of Vessels"(FAVOR Code, Version 06.1)

WCAP- 16168-NP-A June 2008 Revision 2

xxviii

17. Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," (ADAMS Accession No. ML010890301)

18. Letter Report, "Beaver Valley Pressurized Thermal Shook (PTS) Probabilistic Risk Assessment (PRA)," March 3, 2005 (ADAMS Accession No. ML042880454)

19. Letter Report, "Palisades Pressurized Thermal Shock (PTS) Probabilistic Risk Assessment (PRA)", March 3, 2005 (ADAMS Accession No. ML042880473)

20. Letter Report, "Oconee Pressurized Thermal Shock (PTS) Probabilistic Risk Assessment (PRA)," March 3, 2005 (ADAMS Accession No. ML042880452)

21. NUREG/CR-6859, "PRA Procedures and Uncertainty for PTS Analysis," October 6, 2004 (ADAMS Accession No. ML061580379)

22. Letter Report, "Generalization of Plant-Specific Pressurized Thermal Shock (PTS) Risk Results to Additional Plants," December 14, 2004 (ADAMS Accession No. ML042880482)

23. Letter Report, "Estimate of External Events Contribution to Pressurized Thermal Shock (PTS) Risk," October 1, 2004 (ADAMS Accession No. ML042880476)

Attachment Resolution of PWROG Comments on Draft SE Principle Contributors: Barry Elliott Stephen Dinsmore Date: May 8, 2008 WCAP- 16168-NP-A June 2008 Revision 2

Xxix RESOLUTION OF PRESSURIZED WATER REACTOR OWNERS GROUP (PWROG) COMMENTS ON DRAFT SAFETY EVALUATION (SE) FOR TOPICAL REPORT (TR) WCAP-16168-NP, REVISION 2, 'RISK-INFORMED EXTENSION OF THE REACTOR VESSEL IN-SERVICE INSPECTION INTERVAL" (TAC NO. MC9768) By letter dated March 31, 2008, the PWROG provided thirteen comments on the draft SE for TR WCAP-16168-NP, Revision 2. The following are the NRC staff's resolution of these comments. To ensure consistency when discussing the final and proposed rule within the SE, the NRC staff has made one additional change, noted at Number 14.

1. Page 3, Lines 19-21 PWROG Comment:

It is stated in the draft SE that: "This correlation took into considerati6n the contribution to TWCF [through wall cracking frequency] from each of the most limiting plate, axial weld, and circumferential welds." This correlation also took into consideration forgings. Therefore, the following change is suggested: "This correlation took into consideration the contribution to TWCF from each of the most limiting plate, forging, axial weld, and circumferential welds." NRC Response: The NRC staff agrees with this change.

2. Page 15, Line 41 PWROG Comment:

The change in risk (9.43E-10!year) for Beaver Valley Unit I (BVI) should be revised to 9.37E-t0/year to be consistent with the value documented in WCAP-16168-NP, Revision 2, and the response to Request for Additional Information question number 8. NRC Response: The NRC staff agrees with this change.

3. Page 17, Lines 11-18 PWROG Comment:

The draft SE requires that the qualified vessel inspection results be evaluated per the existing requirements in Section (e) of 10 CFR 50.61a in Enclosure I of SECY-07-0104, Reference 12. It is requested that the SE be revised to state that the requirements of Section (e) in Enclosure 1 of SECY 0104 should only be used until the applicable requirements in the final ATTACHMENT WCAP- 16168-NP-A June 2008 Revision 2

xxx version of 10 CFR 50.61a are published in the Federal Register. The following revision is recommended, "By monitoring flaw sizes in accordance with the criteria described in Section (e) of the proposed rulemaking in SECY-07-0104, or the final published version of 10 CFR 50.61a, licensees will ensure...." NRC Response: While the NRC staff agrees with the intent of the requested change, the NRC staff does not agree the revised wording accomplishes the intent. The NRC staff has made the following change: "By monitoring flaw sizes in accordance with the criteria described in Section (e) of the final 10 CFR 50.61a (or the proposed 10 CFR 50.61a, given in 72 FR 56275 prior to issuance of the final 10 CFR 50.61a) licensees will ensure that their RVs do not. have~flaws that invalidate the results of the PWROG PFM analyses.7

4. Page 18, Lines 27-34 PWROG Comment:

The draft SE requires that the qualified vessel inspection results be evaluated per the existing requirements in Section (e) of 10 CFR,50.61a in Enclosure 1 of SECY-07-01 04, Reference 12. It is requested that the SE be revised to state that the requirements of Section (e) in Enclosure 1 of SECY-07-0104 should only be used until the applicable requirements in the final version of 10 CFR 50.61a are published in the Federal Register. The following revisions are recommended, "...in Enclosure 1 to the proposed rulemaking in SECY-07-01 04, Reference 12, or the final published version of 10 CFR 50.61 a." and "...and analyses requested in Section (e) of the proposed rulemaking in SECY-07-0104, or the final published version of 10 CFR 50.61a, will be submitted....2 NRC Response: While the NRC staff agrees with the intent of the requested change, the NRC staff does not agree the'revised wording accomplishes the intent. The NRC staff has made the following change: "Within one year of completing each of the ASME Code, Section XI, Category B-A and B-D RV weld inspections required in the proposed ISI interval, the licensee must provide the information and analyses requested in Section (e) of the final 10 CFR 50.61a (or Alternative fracture tou-ghness requirements for protection against pressurized thermal shock, in Enclosure 1 to the proposed rulemaking in SECY-07-0104, Reference 12, given in 72 FR 56275 prior to issuance of the final 10 CFR 50.61a). Licensees that do not implement 10 CFR 50.61a must amend their licenses to require that the information and analyses requested in Section (e) of the final 10 CFR 50.61a (or the proposed 10 CFR 50.61a, given in 72 FR 56275 prior to issuance of the final 10 CFR 50.61 a) will be submitted for NRC staff review and approval. The amendment to the license shall be submitted at the same time as the request for alternative."

5. Page 18, Lines 41-44 PWROG Comment:

It is stated in the draft SE that: "Licensees also implementing Section (c) of the proposed WCAP- 16168-NP-A June 2008 Revision 2

                                                                                                              . Xxxi 10 CFR 50.61a must perform the inspections and analyses required by Section (e) of the proposed 10 CFR 50.61a and may not defer the 1SI inspection of the RV beltline welds." The following revision is recommended: "Licensees also implementing Section (c) of the proposed 10 CFR 50.61 a must perform the inspections and analyses required by Section (e) of the proposed 10 CFR 50.61a prior to implementing the extended interval."

NRC Response: The NRC staff has made the following change: "Licensees that implement 10 CFR 50,61a must perform the ISIs required in Section (e) of the rule and must submit the required information for review and approval to the Director, Office of Nuclear Reactor Regulation, in accordance with Section (c) of the rule, at least three years before the limiting RTp-rs value calculated under 10 CFR 50.61 is projected to exceed the PTS screening criteria in 10 CFR 50.61. Licensees implementing Section (c) of 10 CFR 50.61a must perform the inspections and analyses required by Section (e) of 10 CFR 50.61 a prior to implementing the extended interval."

6. Page 20, Lines 15-17 PWROG Comment:

It is stated in the draft SE that: "Surface cracks that penetrate through the cladding ....were not part of the PTS Risk Study." However, Oconee Unit 1 included these surface cracks in the PTS risk analyses of NUREG-1806 and NUREG-1874; even though they did not contribute to the TWCF. It is suggested that the SE be revised to state, "Surface cracks that penetrate through the cladding and into the ferritic steel have not been observed in the beltline of operating PWR Reactors. PFM analyses indicate.... NRC Response: The NRC staff does not agree with the change. Surface defects through the clad were included in the PTS study. However, surface defects though the clad that penetrate into the ferritic steel were not included in the PTS study, Therefore, the SE will not be revised with the suggested wording.

7. Page 21, Lines 21-28 PWROG Comment:

The draft SE requires that the qualified vessel inspection results be evaluated per the existing requirements in Section (e) of 10 CFR 50.61a in Enclosure 1 of SECY-07-0104, Reference 12., It is requested that the SE be revised to state that the requirements of Section (e) in Enclosure 1 of SECY-07-0104 should only be used until the applicable requirements in the final version of 10 CFR 50.61a are published in the Federal Register. The following revisions are recommended, "...in'Enclosure 1 to the proposed rulemaking in SECY-07-0104, Reference 12, or the final publishedversion of 10 CFR 50.61a." and "...and analyses requested in Section (e) WCAP- 16168-NP-A June 2008 Revision 2

XXXII of the proposed rulemaking in SECY-07-0104, or the final published version of 10 CFR 50.61 a, wil be submitted...." NRC Response: While the NRC staff agrees with the intent of the requested change, the NRC staff does not agree the revised wording accomplishes the intent. The NRC staff has made the following change: 'Within one year of completing each of the ASME Code, Section X1, Category B-A and B-D RV weld inspections required in the proposed ISI interval, the licensee must provide the information and analyses requested in Section (e) of the final 10 CFR 50.61a (or Altemative fracture toughness requirements for protection against pressurized thermal shock, in Enclosure I to the proposed rulemaking in SECY-07-0104, Reference 12, given in 72 FR 56275 prior to issuance of the final 10 CFR 50.61a). Licensees that do not implement 10 CFR 50.61a must amend their licenses to require that the information and analyses requested in Section (e) of the final 10 CFR 50.61a (or the proposed 10 CFR 50.61a, given in 72 FR 56275 prior to issuance of the final 10 CFR 50.61a) will be submitted for NRC staff review and approval. The amendment to the license shall be submitted at the same time as the request for alternative."

8. Page 21, Lines 35-38 PWROG Comment:

It is stated in the draft SE that:. "Licensees also implementing Section (c) of the proposed 10 CFR 50,61 a must perform the inspections and analyses required by Section (e) of the proposed 10 CFR 50.61a and may not defer the ISI inspection of the RV beltline welds." The following revision is recommended, "Licensees also implementing Section (c) of the proposed 10 CFR 50.61a must perform the inspections and analyses required by Section (e) of the proposed 10 CFR 50.61a prior to implementing the extended interval." NRC Response: The NRC staff has made the following change: "Licensees that implement 10 CFR S0.61a must perform the ISIs required in Section (e) of the rule and must submit the required information for review and approval to the Director, Office of Nuclear Reactor Regulation, in accordance with Section (c) of the rule, at least three years before the limiting RTpTs value calculated under 10 CFR 50.61 is projected to exceed the PTS screening criteria in 10 CFR 50.61. Licensees implementing Section (c) of 10 CFR 50.61a must perform the inspections and analyses required by Section (e) of 10 CFR 50.61a prior to implementing the extended interval."

9. Page 23, Line 27 PWROG Comment:

The date and Agencywide Documents Access and Management System (ADAMS) Accession number for Revision 1 of Reference 11 are October 31, 2003, and ML051790410, respectively. WCAP-16168-NP-A June 2008 Revision 2

xxxiii NRC Response: The NRC staff agrees with this change.

10. Page 23, Line 35 PWROG Comment:

ADAMS Accession number ML012630333 for Reference 13 could not be found on ADAMS. ADAMS Accession numbers ML042610469 and ML042610375 can be used for WCAP-14572 and Supplement I on the probabilistic structural reliability and risk assessment tool, respectively. It is recommended that the SE be revised to include these accession numbers for Reference 13. NRC Response: The NRC staff agrees with this change.

11. Page 23, Line 42 PWROG Comment:

For version 06.1 of FAVOR, Reference 16, the WCAP Technical Report used letter ORNL/TM-2007/0030, which is the same as "Williams 07" in NUREG-1874. It is recommended that this reference for FAVOR be used in the SE.

      *NRC Response:

The NRC staff agrees with this change.

12. Page 24, Line 11 PWROG Comment:

For Reference 22, the ADAMS Accession Number is ML042880482. It is recommended that this accession number be added to the SE. NRC Response: The NRC staff agrees with this change.

13. " Page 24, Line 13 PWROG Comment:

Reference 23 is cited in Section 3.2.2.3 (Page 15, Line 18) but not included inthe list of references in Section 5.0. The following text is suggested for addition to the SE: "23. Letter Report, "Estimate of External Events Contribution to Pressurized Thermal Shock (PTS) Risk," October 1, 2004 (ADAMS Accession No. ML042880476)" WCAP- 16168-NP-A June 2008 Revision 2

xxxiv NRC Response: The NRC staff agrees with this change.

14. Page 20, Lines 4144 NRC Comment:

To ensure consistency when discussing the final and proposed rule within the SE, the NRC staff has made one additional change to the SE. The NRC staff has modified the following sentence:

        "In addition, licensees that do not implement the proposed 10 CFR 50.61a must amend their licenses to require that the information and analyses requested in Section (e) of the proposed 10 CFR 50.61 a will be submitted for NRC staff review and approval."

NRC Response: The NRC staff has made the following change: "in addition, licensees that do not implement 10 CFR 50.61 a must amend their licenses to require that the information and analyses requested in Section (e) of the final 10 CFR 50.61a (or the proposed 10 CFR 50.61a, given in 72 FR 56275 prior to issuance of the final 10 CFR 50.61a) will be submitted for NRC staff review and approval." June 2008 WCAP- 16168-NP-A WCAP-16168-NP-A June 2008 Revision 2

xxxv LEGAL NOTICE This report was prepared as an account of work performed by Westinghouse Electric Company LLC. Neither Westinghouse Electric Company LLC, nor any person acting on its behalf: A. Makes any warranty or representation, express or implied including the warranties of fitness for a particular purpose or merchantability, with respect to the accuracy, completeness, or usefulness of the information contained in this report, or that the use of any information, apparatus, method, or process disclosed in this report may not infringe privately owned rights; or B. Assumes any liabilities with respect to the use of, or for damages resulting from the use of, any information, apparatus, method, or process disclosed in this report. COPYRIGHT NOTICE This report has been prepared by Westinghouse Electric Company LLC and bears a Westinghouse Electric Company copyright notice. As a member of the Westinghouse Owners Group, you are permitted to copy and redistribute all or portions of the report within your organization; however all copies made by you must include the copyright notice in all instances. DISTRIBUTION NOTICE This report was prepared for the PWR Owners Group. This Distribution Notice is intended to establish guidance for access to this information. This report (including proprietary and non-proprietary versions) is not to be provided to any individual or organization outside of the PWR Owners Group program participants without prior written approval of the PWR Owners Group Program Management Office. However, prior written approval is not required for program participants to provide copies of Class 3 Non Proprietary reports to third parties that are supporting implementation at their plant, and for submittals to the NRC. .WCAP-16168-NP-A June 2008 Revision 2

xxxvii ACKNOWLEDGEMENTS The authors acknowledge with appreciation those utility representatives and Westinghouse personnel who provided support in developing this risk-informed application and Topical Report. In particular the authors acknowledge the following individuals: Dennis Weakland FirstEnergy Corporation Michael Acker Nuclear Management Company - Palisades Mel Arey Duke Energy Kevin Hall Entergy South Maurice Dingler Wolf Creek Nuclear Operating Company The authors also acknowledge the following past and present employees of Westinghouse Electric Company LLC who contributed to this Topical Report: Chris Hoffmann Richard Haessler Jim Andrachek Barry Sloane Owen Hedden Finally, the authors would like to thank members of the PWR Owners Group (PWROG) Materials Subcommittee and the former Combustion Engineering Owners Group (CEOG) Section XI Subcommittee, Jim Molkenthin and Gordon Bischoff of the PWROG Project Office, and Ted Schiffley, Chairman of the PWROG, for their continued support in the development of this risk-informed application and Topical Report. WCAP-16168-NP-A June 2008 Revision 2

xxxix PWR Owners Group Member Participation* for PWROG Project MUHP-5097, MUHP-5098, MUHP-5099, and CEOG Task 2059. Plant Site(s) Participant Utility Member Yes No AmerenUE Callaway (W) X American Electric Power D.C. Cook l&2 (W) X Arizona Public Service Palo Verde Unit 1, 2, & 3 (CE) X Constellation Energy Group Calvert Cliffs 1 & 2 (CE) X Constellation Energy Group Ginna (W) - X Dominion Connecticut Millstone 2 (CE) X Dominion Connecticut Millstone 3 (W) X Dominion Kewaunee Kewaunee (W) X Dominion VA North Anna 1 & 2, Surry 1 & 2 (W) X Duke Energy Catawba 1 & 2, McGuire 1 & 2 (W), X Oconee 1, 2, 3 (B&W) Entergy Palisades (CE) X Entergy Nuclear Northeast Indian Point 2 & 3 (W) X Arkansas 2, Waterford 3 (CE), X Entergy Operations South Akna 1 (B&W) Arkansas BW __ Exelon Generation Co. LLC Braidwood 1 & 2, Byron 1 & 2 (W), X TMI I (B&W) FirstEnergy Nuclear Operating Co Beaver Valley 1 & 2 (W), Davis-Besse X (B&W) Florida Power & Light Group St. Lucie 1 & 2 (CE) X Florida Power & Light Group Turkey Point 3 & 4, Seabrook (W) X Florida Power & Light Group Pt. Beach l&2 (W) X Luminant Power Comanche Peak 1 & 2 (W) X Nuclear Management Company Prairie Island l&2 X Omaha Public Power District Fort Calhoun (CE) X Pacific Gas & Electric Diablo Canyon I & 2 (W) X Progress Energy Robinson 2, Shearon Harris (W), X Crystal River 3 (B&W) PSEG - Nuclear Salem 1 & 2 (W) X WCAP- 16168-NP-A June 2008 Revision 2

xl Plant Site(s) Participant Utility Member Yes No Southern California Edison SONGS 2 & 3 (CE) X South Carolina Electric & Gas V.C. Summer (W) X So. Texas Project Nuclear Operating Co. South Texas Project 1 & 2 (W) X Southern Nuclear Operating Co. Farley 1 & 2, Vogtle I & 2 (W) X Tennessee Valley Authority Sequoyah I & 2, Watts Bar (W) X Wolf Creek Nuclear Operating Co. Wolf Creek (W) X Project participants as of the date the final deliverable was completed. On occasion, additional members will join a project. Please contact the PWR Owners Group Program Management Office to verify participation before sending this document to participants not listed above. June 2008 WCAP- 16168-NP-A WCAP-16168-NP-A June 2008 Revision 2

xli PWR Owners Group International Member Participation* for PWROG Project MUHP-5097, MUHP-5098, MUHP-5099, and CEOG Task 2059. Plant Site(s) Participant Utility Member Yes No British Energy Sizewell B X Electrabel (Belgian Utilities) Doel 1, 2 & 4, Tihange 1 & 3 X Hokkaido Tomari 1 & 2 (MHJ) X Japan Atomic Power Company Tsuruga 2 (MHI) X Kansai Electric Co., LTD Mihama 1, 2 &3, Ohi 1, 2, 3 & 4, X Takahama 1, 2, 3 &4 (W & MHI) Kori1, 2, 3& 4 X Korea Hydro & Nuclear Power Corp. Yonggwang,1 & 2 ( Yonggwang I & 2 (W) Korea Hydro & Nuclear Power Corp. Yonggwang 3, 4, 5 & 6 X Ulchin 3, 4,5 & 6(CE) Kyushu Genkai 1, 2, 3 & 4, Sendai 1 & 2 (MI1) X Nuklearna Electrarna KRSKO Krsko (W) X Nordostschweizerische Kraftwerke AG Beznau 1 & 2.(W) X (NOK) Ringhals AB Ringhals 2, 3 & 4 (W) X Shikoku Ikata 1, 2 & 3 (MHI) X Spanish Utilities Asco 1 & 2, Vandellos 2, X Almaraz 1 & 2 (W) Taiwan Power Co. Maanshan 1 & 2 (W) X Electricite de France 54 Units X

• This is a list of participants in this project as of the date the final deliverable was completed. On occasion, additional members will join a project. Please contact the PWR Owners Group Program Management Office to verify participation before sending documents to participants not listed above.

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xliii TABLE OF CONTENTS L IST OF TA BL E S ..................................................................................................................................... x lv LIST O F FIG U RE S .......................................................................................................... xlvii LIST OF ACRONYMS AND ABBREVIATIONS .............................................................................. xlix E X E CU TIV E SUM M A RY ........................................................................................................................... li 1 IN TR O D U C T ION ........................................................................................................................ 1-1 2 BA C KG RO UN D ........................................................................................................................... 2-1 2.1 REACTOR VESSEL IN-SERVICE INSPECTION ..................................................................... 2-2 2.2 LOCATION-SPECIFIC ISI DATA FROM PARTICIPATING PLANTS ..................................... 2-4 2.3 EXPOSURE AND COST REDUCTION ..................................................................................... 2-7 2.4 GENERIC REACTOR VESSEL WELD EXPERIENCE AT VARIOUS PLANTS ..................... 2-7 2.5 DEVELOPMENT OF ISI INTERVAL EXTENSION METHODOLOGY .................................. 2-8 3 PILO T PLA NT SU MM ARY ........................................................................................................ 3-1 3.1 BO U N D IN G LO CATIO N ....................................................................................................... 3-1 3.2 BASIS FOR RISK DETERMINATION ....................................................................................... 3-5 3.3 RESULTS FOR THE WESTINGHOUSE PILOT PLANT: BVI ............................................. 3-17 3.4 RESULTS FOR THE COMBUSTION ENGINEERING PILOT PLANT: PALISADES ......... 3-20 3.5 RESULTS FOR THE BABCOCK AND WILCOX PILOT PLANT: OC1 ............................... 3-23 4 RISK A SSE SSM EN T ................................................................................................................... 4-1 4.1 RISK-INFORMED REGULATORY GUIDE 1.174 METHODOLOGY ..................................... 4-1 4.2 FA ILU RE M O DES AN D EFFECTS ............................................................................................ 4-6 4.3 CORE DAMAGE RISK EVALUATION ..................................................................................... 4-7 5 C O N CL U SIO N S .......................................................................................................................... 5-1 6 RE FE REN CE S ............................................................................................................................. 6-1 WCAP- 16168-NP-A June 2008 Revision 2

xliv TABLE OF CONTENTS (cont.) APPENDIX A PLANT SPECIFIC APPLICATION ......................................................................... A-1 APPENDIX B INPUTS FOR THE BEAVER VALLEY UNIT 1 PILOT PLANT EVALUATION ....... B-1 APPENDIX C BEAVER VALLEY UNIT 1 PROBSBFD OUTPUT ........................ C-1 APPENDIX D BEAVER VALLEY UNIT 1 DOMINANT PTS TRANSIENTS ................ D-1 APPENDIX E BEAVER VALLEY UNIT 1 FAVPOST OUTPUT ......................................................... E-1 APPENDIX F INPUTS FOR THE PALISADES PILOT PLANT EVALUATION .......................... F-i APPENDIX G PALISADES PROBSBFD OUTPUT ....................................................................... G-1 APPENDIX H PALISADES DOMINANT PTS TRANSIENTS ....................................................... H-1 APPENDIX I PALISADES FAVPOST OUTPUT ............................................................................... I-1 APPENDIX J INPUTS FOR THE OCONEE UNIT 1 PILOT PLANT EVALUATION ................... J-1 APPENDIX K OCONEE UNIT 1,PROBSBFD OUTPUT ............................... K-1 APPENDIX L OCONEE UNIT 1 DOMINANT PTS TRANSIENTS .............................................. L-I APPENDIX M OCONEE UNIT I FAVPOST OUTPUT ....................................................................... M -1 APPENDIX N RESPONSES TO THE NRC REQUEST FOR ADDITIONAL INFORMATION (RAI) REGARDING THE REVIEW OF WCAP-I6168-NP, REVISION 1 ............. N-1 WCAP- 16168-NP-A June 2008 Revision 2

xlv LIST OF TABLES Table 2-1 Summary of Survey Results on RV ISI Findings [21 ]..................................................... 2-5 Table 2-2 Savings on the Proposed Extension of RV ISI Interval from 10-Years to 20-Years (Per P la n t) [2 1)) ......................................................................................................................... 2-7 Table 3-1 ASME Section XI [1] ISI Requirements for RPVs (ASME Section XI, Table IWB-2500-

1) ...................................................................................................................................... 3 -2 Table 3-2 BV I Reactor Vessel Failure Frequency Results ............................................................ 3-17 Table 3-3 Palisades Reactor Vessel Failure Frequency Results ..................................................... 3-20 Table 3-4 OCI Reactor Vessel Failure Frequency Results ............................................................ 3-23 Table 4-1 Large Early Release Frequencies .................................................................................... 4-8 Table 4-2 Evaluation with Respect to Regulatory Guide 1.174 [4] Key Principles ......................... 4-9 Table A- I Critical Parameters for the Application of the Bounding Analysis ............................ A-1 Table A-2 Additional Information Pertaining to the Reactor Vessel Inspection ......................... A-5 Table A -3 D etails of TW CF Calculation .................................................................................... ...A -7 Table B-I Cladding Material Properties ......................................................................................... B -4 Table B-2 Base M etal M aterial Properties ................................................................................. B-5 Table B-3 BV 1-Specific Material Values Drawn from the RVID (see Ref. 44, Table 4.1) ............ B-6 Table B-4 Summary of Reactor Vessel-Specific Inputs for Flaw Distribution ................................ B-7 Table D-1 PTS Transient Descriptions for BVI ......................................................................... D-1 Table F-1 Cladding M aterial Properties ......................................................................................... F-3 Table F-2 Base M etal M aterial Properties ...................................................................................... F-4 Table F-3 Palisades-Specific Material Values Drawn from the RVID (see Ref. 44 Table 4.1) ....... F-5 Table F-4 Summary of Reactor Vessel-Specific Inputs for Flaw Distribution ................................ F-6 Table H-1 PTS Transient Descriptions for Palisades ................................................................... H-1 Table J-1 Cladding M aterial Properties .......................................................................................... J-4 Table J-2 Base M etal M aterial Properties ................................................................................... J-5 Table J-3 OC I-Specific Material Values Drawn from the RVID (see Ref. 44 Table 4.1) ............... J-6 Table J-4 Summary of Reactor Vessel-Specific Inputs for Flaw Distribution ................................. J-7 Table L-1 PTS Transient Descriptions for OC1 ......................................................................... L-l WCAP- 16168-NP-A June 2008 Revision 2

xlvii LIST OF FIGURES Figure 3-1 Comparison to Acceptance Criteria - Minimum Margins Code Allowable ............. 3-4 Figure 3-2 Comparison to Acceptance Criteria - Minimum Margins Code Allowable ............................ 3-5 Figure 3-3 Weld Stress P rofi le ................................................................................................................ 3-11 Figure 3-4 ISI D etection Probability ....................................................................................................... 3-12 Figure 3-5 Software and Data Flow for Pilot Plant Analyses ................................................................. 3-15 Figure 3-6 Growth of Flaws with an Aspect Ratio of 10 for BV 1 .......................................................... 3-18 Figure 3-7 Growth of Flaws with an Infinite Aspect Ratio for BV1 ........................ 3-19 Figure 3-8 Growth of Flaws with an Aspect Ratio of 10 for Palisades ................................................... 3-21 Figure 3-9 Growth of Flaws with an Infinite Aspect Ratio for Palisades .............................................. 3-22 Figure 3-10 Growth of Flaws with an Aspect Ratio of 10 for OC I ........................................................ 3-24 Figure 3-11 Growth of Flaws with an Infinite Aspect Ratio for OCI ..................................................... 3-25 Figure 4-1 Basic Steps in (Principal Elements of) Risk-Informed, Plant-Specific Decision Making (from N R C RG 1.174) ...................................................................................................................... 4 -2 Figure 4-2 Principles of Risk-Informed Regulation (from NRC RG 1.174) ............................................ 4-3 Figure B-1 Rollout Diagram of Beltline Materials and Representative Fluence Maps for BV1 ...... B-3 Figure F-I Rollout Diagram of Beltline Materials and Representative Fluence Maps for Palisades ....... F-2 Figure J-1 Rollout Diagram of Beltline Materials and Representative Fluence Maps for OC I ............... J-3 WCAP-16168-NP-A June 2008 Revision 2

xlix LIST OF ACRONYMS AND ABBREVIATIONS ADV Atmospheric dump valve AFW Auxiliary feedwater ART Adjusted reference temperature ASME American Society of Mechanical Engineers B&PV Boiler and Pressure Vessel B&W Babcock & Wilcox BVl Beaver Valley Unit 1 CCDP Conditional core damage probability CDF Core damage frequency CE Combustion Engineering ECT Eddy current examination EFPY Effective full-power year EOL End of life EPRI Electric Power Research Institute FENOC FirstEnergy Nuclear Operating Company FCG Fatigue crack growth FP Failure probability FSAR Final Safety Analysis Report GQA Graded quality assurance HPI High-pressure injection HUCD Heat-up and cool-down transient HZP Hot-zero power IEF Initiating event frequency IGSCC Intergranular stress corrosion cracking ID Inner diameter ISI In-service inspection-IST In-service testing LBLOCA Large-break loss-of-coolant accident LERF Large early release frequency LOCA Loss-of-coolant accident MBLOCA Medium-break loss-of-coolant accident MSIV Main steam isolation valve MSLB Main steam line break MT Magnetic particle examination NDE Non-destructive examination NMC Nuclear Management Company NRC Nuclear Regulatory Commission NSSS Nuclear Steam Supply System OC1 Oconee Unit 1 OD Outer diameter ORNL Oak Ridge National Laboratory PFM Probabilistic fracture mechanics PNNL Pacific Northwest National Laboratory POD Probability of detection WCAP- 16168-NP-A June 2008 Revision 2

1 LIST OF ACRONYMS AND ABBREVIATIONS (cont.) PRA Probabilistic risk assessment PT Liquid penetrant examination PTS Pressurized thermal shock PVRUF Pressurized Vessel Research User Facility PWR Pressurized water reactor PWROG PWR Owners Group QA Quality Assurance RAI NRC Request for Additional Information RCP Reactor coolant pump RCS Reactor Coolant System RG NRC Regulatory Guide RI-ISI Risk-informed ISI RPV Reactor pressure vessel RTNDT Reference nil-ductility transition temperature RV Reactor vessel RV ISI Reactor Vessel In-service Inspection RVID Reactor vessel integrity database SBLOCA Small-break loss-of-coolant accident SER NRC Safety Evaluation Report SG Steam generator SRP Standard Review Plan SRRA Structural Reliability and Risk Assessment SRV Safety and relief valve SSC Structures, systems, and components TH Thermal hydraulics TWCF Through Wall Cracking Frequency UT Ultrasonic examination VT Visual examination WCAP- 16168-NP-A June 2008 Revision 2

li EXECUTIVE

SUMMARY

The current requirements for the inspection of reactor vessel pressure-containing welds have been in effect since the 1989 Edition of American Society of Mechanical Engineers (ASME,) Boiler and Pressure Vessel Code, Section XI, as supplemented by Nuclear Regulatory Commission (NRC) Regulatory Guide 1.150. The manner in which these examinations are conducted has recently been augmented by Appendix VIII of Section XI, 1996 Addenda, as implemented by the NRC in an amendment to 10CFR50.55a effective November 22, 1999. The industry has expended significant cost and man-rem exposure that have shown no service-induced flaws in the reactor vessel (RV) for ASME Section XI Category B-A or B-D RV welds. The objective of the methodology discussed in this report is to provide the technical basis for decreasing the frequency of inspection by extending the Section XI Inspection interval from the current 10 years to 20 years for ASME Section XI Category B-A and B-D RV nozzle welds. Specific pilot studies have been performed on the Westinghouse, Combustion Engineering, and Babcock and Wilcox reactor vessel and NSSS designs. The results show that the change in risk associated with eliminating all inspections after the initial 10-year in-service inspection satisfies the guidelines specified in Regulatory Guide 1.174 for an acceptable change in risk for large early release frequency (LERF). This conclusion is applicable to all Westinghouse, Combustion Engineering, and Babcock and Wilcox reactor vessel designs given that the applicable individual plant parameters are bounded by the critical parameters identified in Appendix A. WCAP- 16168-NP-A June 2008 Revision 2

1-1 1 INTRODUCTION The current requirements for the inspection of reactor vessel (RV) pressure containing welds have been in effect since the 1989 Edition of American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section XI [1], as supplemented by the U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.150 [2]. The manner in which these examinations are conducted has been augmented by Appendix VIII of Section XI, 1996 Addenda, as implemented by NRC in an amendment to 10CFR50.55a effective November 22, 1999 [3]. The industry has expended significant cost and man-rem exposure by performing the required examinations that have shown no service-induced flaws in the RV for ASME Section XI Category B-A or B-D RV nozzle welds. The current code criteria for the selection of examination areas and the frequency of examinations is not be an effective way to expend inspection resources. The objective of this study was to verify that a reduction in frequency of volumetric examination of the RV full-penetration welds could be accomplished with an acceptably small change in risk. The methodology used to justify this reduction involved an evaluation of the change in risk associated with extending the 10-year in-service inspection (ISI) interval for three pilot plant bounding cases based on the calculated difference in the frequency of RV failure: RV failure was defined for this study to be the extension of a crack all the way through the RV wall. The difference in frequency of RV failure was evaluated using RG 1.174 [4] to determine if the values met the specified regulatory guidelines. The intent was that licensees can then use the results of this bounding assessment to demonstrate that their RV and plant are bounded by the generic analysis, thereby justifying a plant-specific extension in the RV weld inspection interval. This study followed the approach specified in ASME Code Case N-691 [5], which provides guidelines for using risk-informed insights to increase the inspection interval for pressurized water reactor (PWR) vessel welds. WCAP-16168-NP-A June 2008 Revision 2

2-1 2 BACKGROUND The original objective of the ASME B&PV Code, Section XI [1] ISI program was to assess the condition of pressure-containing components in nuclear power plants to ensure continued safe operation. If non-destructive examination (NDE) found indications that exceeded the allowable standards, examinations were extended to additional welds in components in the same examination category. If NDE found indications that exceeded the acceptance standards in those welds, then the examinations were extended further to similar welds in similar components, etc. With respect to the method defined in this report, 100 percent of the present examination areas will be retained. The methodology is limited to justification of a reduction in the frequency of examination, i.e., increasing the time interval between inspections. The original examination interval of 10 years was based on "wear-out" rate experience in the pre-nuclear utility and petrochemical process industries. As with some other Section XI ISI requirements, with no indications being found in the vessel welds under evaluation in this report, these inspections are decreasing in value with increasing industry experience to rely upon. The U.S. NRC has granted a number of exemptions to inspections for other areas and components (e.g., piping [6], reactor coolant pump motor flywheels [7], etc.) based on experience and man-rem reductions. This has been attributed to the combined design, fabrication, examination, and Quality Assurance (QA) rigor of the nuclear codes, and more careful control of plant operating parameters by the utilities. Acritical component of the justification of the interval extension is a fracture mechanics evaluation of the reactor vessel, which shows that flaws, if they do exist, would not grow to a critical size if the inspection interval is increased to more than 10 years. This can be demonstrated by selecting critical areas of the reactor vessel for the evaluation such as, the beltline, flange, and outlet nozzle regions. These locations are known-to be areas of primary concern and are currently considered in ASME Section III, Appendix G [1] evaluations for protection against nonductile failure of the reactor vessel. As part of this study, a deterministic fracture mechanics evaluation of limiting locations in a typical geometry for a RV identified that the beltline region was the critical location with respect to the potential for growth of fatigue cracks. Fatigue crack growth is recognized as the primary degradation mechanism in the carbon and low alloy steel components in PWR Nuclear Steam Supply System (NSSS), that could contribute to any potential growth of existing flaws in the component base materials and weld metals. Fatigue can be defined as repeated exposure to cyclic loading resulting from a variety of operating conditions or events (e.g., heatups, cooldowns, reactor trips). Design basis documents provided descriptions of the conditions that would contribute to cyclic fatigue. This information was used to identify and define the frequency of occurrence for each of the events that was considered when determining the potential for fatigue crack growth. Atechnical consideration critical to success was the application of risk-informed assessment techniques to substantiate the deterministic fracture mechanics flaw growth evaluation. Risk assessment techniques provided a means to quantify and calculate cumulative results from contributing mechanisms and uncertainties associated with the critical parameters. A probabilistic fracture mechanics (PFM) methodology was used to consider the distributions and uncertainties in flaw numbers, flaw sizes, fluence, material properties, crack growth rate, stresses, and the effectiveness of inspections. The PFM WCAP- 16168-NP-A June 2008 Revision 2

2-2 methodology was also used to calculate the change in the frequency of RV failure due to a change in inspection interval. This change in RV failure frequency was used to evaluate the viability of such an inspection interval change. Recognized guidelines for evaluating the change in failure frequencies are provided in RG 1.174 [4] and the NRC risk assessment developed in conjunction with the current pressurized thermal shock (PTS) evaluations [8, 9]. Significant work is on-going in the nuclear industry to investigate the impacts from PTS or "off-normal" plant transients that may be outside the current design basis. These transients are commonly understood to present the most severe challenge to RV structural integrity. The NRC effort to address PTS has identified FirstEnergy Nuclear Operating Company's (FENOC's) Beaver Valley Unit I (BV 1), Nuclear Management Company's (NMC's) Palisades, and Duke Energy's Oconee Unit 1 (OC 1) as the representative plants based on geometry and embrittlement for the Westinghouse, Combustion Engineering (CE), and Babcock and Wilcox (B&W) PWR designs. These are the primary PWR manufacturers in the U.S. and were evaluated by the NRC and Oak Ridge National Laboratory (ORNL) as part of the NRC PTS Risk Study [8, 9]. This report summarizes the results from an evaluation of the extension of the inspection of ASME Section XI [1] Examination Category B-A and B-D welds in the reactor pressure vessel (RPV) from the current requirement of every 10 years to an extension of 20 years. It demonstrates that for the pilot plant reactor vessel geometry and fabrication history, any potential change in risk when the inspection interval is extended meets the change in risk evaluation guidelines defined in RG 1.174 [4]. The evaluation documented in this report considers FENOC's BVI as the Westinghouse pilot plant. NMC's Palisades Plant and Duke Energy's OC1 are the respective Combustion Engineering (CE) and Babcock and Wilcox (B&W) pilot plants for this evaluation. To apply the results of this report to non-pilot plants, it must be shown, using the tables contained in Appendix A that the pilot plant evaluations for the respective design bound the non-pilot plant. The following paragraphs address the current Section XI ISI requirements for PWR RV welds under consideration for the proposed extension. The following topics are included:

1. Reactor Vessel In-Service Inspection (RV ISI)

2. Location-specific ISI data from participating plants

3. The man-rem exposure and other costs of RV weld inspection

4. Generic RV weld experience at various plants

5. Development of the ISI interval extension methodology

6. Pilot plants

7. Safety impact 2.1 REACTOR VESSEL IN-SERVICE INSPECTION Since its beginning, ASME B&PV Code, Section XI [1] has required inspections of weld areas of reactor vessels and other pressure-containing nuclear system components. The selection of inspection locations was based on areas known to have high-service factors and additional areas to provide a representative sampling for the condition of pressure-containing nuclear system components. While weld and adjoining areas were specified, it was recognized that the volumetric examination of the weld and adjoining base material would result in a significant degree of examination of the base metal.

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2-3 Examination Volumes Initially, for longitudinal and circumferential welds in a reactor vessel shell, Section XI required examination of 10 percent of the length of longitudinal welds, and 5 percent of the length of circumferential welds. Welds receiving exposure in excess of specified neutron fluence would require an inspection of 50 percent of the length. The 1977 Edition of Section XI increased the examination of RV welds from 5 or 10 percent of the length to 100 percent, with all welds examined in the first 10-year interval. Subsequent intervals required 100 percent examination of specified circumferential and longitudinal welds. The 1989 Edition of Section XI [1] extended the examination to include all welds. There has been no report of structural failure or leakage from any full-penetration weld being addressed in this report in a PWR RV shell, globally. In volumetric examinations of these welds in ISIs performed in accordance with the requirements of Section XI (and RG 1.150 [2]), flaws identified in the original construction have been detected and were acceptable under Section XI requirements. These flaws have been monitored and to date, no growth has been identified. There has been no evidence of in-service flaw initiation in these welds. Examination Approaches The preceding discussion of RV welds addresses the Category B-A, RV seam welds of Table IWB-2500-1 of Section XI. Category B-D, RV nozzle welds and nozzle inner radius are also included in this evaluation. The ultrasonic examinations (UTs) of these RV welds, as of the 1996 Addenda of Section XI, were conducted in accordance with Appendix I, 1-2110. This Addenda requires Appendix VIII inspections for:

• Shell and head welds excluding flange welds
• Nozzle-to-vessel welds
• Nozzle inside radius region Precedent for Change There have been a number of revisions (often by ASME Code Case) to the Section XI ISI program that have eliminated or reduced the extent of examinations and tests based on successful operating experience and analytical evaluation. Examples of ASME Code Cases applicable to the RV and its piping connections include:

N-481 [10] Associated with cast austenitic pump casings. This was the first example of substituting an analysis plus a visual examination (VT) for a volumetric examination, for a Class 1 component. N-560 [11] Permits a reduction in the examination of Class 1 Category B-J piping welds from 25 to 10 percent, provided a specified risk-importance ranking selection process is followed. This was a substantive reduction of an established Class 1 examination. WCAP-16168-NP-A June 2008 Revision 2

2-4 N-577 [12] Provide requirements for risk-informed ISI of Class 1, 2, and 3 piping. The cases provide N-578 [13] different methods to achieve the same objective. This was the first use of the plant probabilistic risk assessment (PRA). Both methods have received extensive implementation in the U.S. and in several other countries in Europe and Asia. N-613 [14] Reduces the examination volume of Category B-D nozzle welds in adjacent material from 1/2 shell thickness to 1/2 inch. This permits a significant reduction in qualification and scanning time. N-552 [15] Permits computational modeling for the qualification of nozzle inner radius examination techniques, in lieu of qualification on a multitude of configurations. N-610 [16] Permits a KIR curve in Appendix G, in lieu of a KIA curve. Indirectly, this is beneficial to the pressure-temperature limit curve during plant startup. Not all of the changes in Section XI, due to operating considerations, have led to a relaxation in inspection or evaluation requirements. Over the past 10 years, there have also been a number of changes (often by code case) to the Section XI ISI program that have increased the extent of examinations and tests based on operating experience and analytical evaluation. The following examples of ASME Code Cases are limited to those applicable to the RV and its piping connections. N-409 [17] Introduced procedure and personnel qualification requirements for UT of intergranular stress corrosion cracking (IGSCC) in austenitic piping welds, a precursor to Appendix VIII, UT performance demonstration requirements. N-512 [18] Provided requirements for the assessment of RVs with low upper shelf Charpy impact energy levels. N-557 [19] Introduced requirements for in-place dry annealing of a PWR RV. 2.2 LOCATION-SPECIFIC ISI DATA FROM PARTICIPATING PLANTS While it is known that the number of flaws found in RPV welds is very small, it is important to relate their number to the number of welds that have been examined over the past 30 years with no evidence of the development of service-induced flaws. To develop location-specific ISI data from nuclear plants, ISI data on the RV weld categories noted above were gathered in a survey [20]. This information focused on service-induced flaws. It did not address the detection of original fabrication flaws, unless the flaws had grown due to service conditions. The response to this survey is summarized in Table 2-1. WCAP- 16168-NP-A June 2008 Revisionr2

2-5 Table 2-1 Summary of Survey Results on RV IS! Findings 1201 Total Years of No. of No. of Service Prior to ASME Weld Welds in Welds with No Welds with Means of Plants Survey Category / Item Category Flaws Flaws Detectioni Cause of Flaw/Failure 14 301 B-A Shell, B1.10 112 112 0 Head. B1.20 105 105 0 Shell-to-flange. B1.30 16 16 0 One plant reported 3 indications that may be just scratches. Head-to-flange, B 1.40 16 16 0 One plant reported 3 indications that may be just scratches. B-D Nozzle-to-shell, B3.90 102 102 0 Nozzle inside radius 102 102 0 B3.100 B-F Dissimilar metal, 84 84 0 B.5.10 B5.30 32 32 0 B-K Welded attach, B 10. 10 4 4 *:2. 00 B-N WCAP-16168-NP-A June 2008 Revision 2

2-6 Table 2-1 Summary of Survey Results on ISI Findings 1201 (cont.) Total Years of No. of No. of Service Prior to ASME Weld Welds in Welds with No Welds with Means of Plants Survey Category Category Flaws Flaws Detection' Cause of Flaw/Failure Vessel interior, 34 34 0 B13.10 Interior attach.- 6 6 0 beltline, B 13.50 Other interior attach., 53 53 0 VT-3, UT, One plant reported crack B13.60 ECT arrest holes drilled in core barrel. Core support struct., 41 5 0 B 13.70 Note 1: VT Visual Inspection, UT = Ultrasonic Inspection, ECT = Eddy Current Inspection WCAP- 16168-NP-A June 2008 Revision 2

2-7 2.3 EXPOSURE AND COST REDUCTION Data was gathered on CE and Westinghouse plants related to the cost of a typical RV ISI outage, as well as the cost of the exposure affecting the involved personnel [20]. The objective of this effort was to investigate the exposure and financial aspects of the RV IS. The results of the survey were tabulated based on the probability of a life extension program (60 years), and the potential savings were calculated with regards to a proposed extension of the RV ISI interval to 20 years. The radiation exposure cost is contingent on the utility and is typically $15,000 to $20,000 per man-rem. A summary of the results is presented in Table 2-2. Table 2-2 Savings on the Proposed Extension of RV ISI Interval from 10-Years to 20-Years (Per Plant) 1201 Probability of 20-Year Life Extension (%) 0% 50% 100% Cost of Typical RV min 506,410 759,615 1,012,820 ISI Outage, $ max 7,680,000 9,600,000 11,520,000 average 3,878,521 5,391,656 7,115,317 Dose of Exposure, min 0.2 0.4 0.6 Man-reins max 6.5 9.75 13.0 average 1.66 2.32 2.98 Cost of Dose of min 2,492 4,984 7,476 Exposure, $ max 65,000 97,500 130,000 average 20,611 28,856 37,101 As shown in Table 2-2, the savings associated with even the most conservative assumption, i.e., no life extension program (40 years) for any of the surveyed plants, are significant. The extension of the RV ISI interval to 20 years will save every unit an average of $3,878,521 for the cost of the outage, and 1.66 man-reins of exposure. The saving values associated with the less conservative assumption of the guaranteed life extension program (60 years) for any of the surveyed plants are considerably higher. The extension of the RV ISI interval to 20 years will save every unit an average of $7,115,317 for the cost of outage, and 2.98 man-reins of exposure. The critical path outage time for RV inspections is approximately 3 1/22 days. While this data was gathered for Westinghouse and CE designed plants, the savings for B&W designed plants are. expected to be similar. 2.4 GENERIC REACTOR VESSEL WELD EXPERIENCE AT VARIOUS PLANTS Section XI ISI requirements developed in the early 1970s were based on the detection of fatigue cracking in primary welds. This has not been substantiated by subsequent operating experience. Fatigue cracking in primary welds has not been a problem. Random sampling for the assessment of condition of pressure-containing components has not been effective; when leakage and other deterioration have been identified, it has been by examinations other than the Section XI ISI NDE. June 2008 WCAP-16 168-NP-A WCAP-16168-NP-A June 2008 Revision 2

2-8 Primary system failures/leakage have almost always been associated with dissimilar metal welds or control rod drive, bottom mounted instrumentation, or vent connections of the RV and its head. The latter connections are all partial penetration welds. They were not included in the survey, since the current effort does not propose to recommend changes to their present ISI interval requirements. Their examinations are not contingent on the removal of the reactor internals and the use of the RV inspection tool. Category B-F dissimilar metal welds, Category B-K welded attachments, and Category B-N interior attachment and support welds were not included in the inspection interval extension. In many plants, the most highly stressed reactor vessel weld is the weld between the closure head flange and the dome. There have been no reports of degradation of this joint. This joint ranks quite low in its contribution to cumulative risk determined through typical PFM methods. Calculations [21] have shown that flaw growth due to fatigue would be extremely small, so that even pre-existing flaws that clearly exceed the acceptance standards would not be subject to measurable growth. 2.5 DEVELOPMENT OF ISI INTERVAL EXTENSION METHODOLOGY The ISI interval extension methodology is primarily based on a risk analysis, including a PFM analysis of the effect of different inspection intervals on the frequency of reactor vessel failure due to postulated PTS transients. Reactor vessel failure is defined for the purposes of this study as the point which a crack has extended all the way through the RV wall. The likelihood of reactor vessel failure is postulated to increase with increasing time of operation due to the growth of pre-existing fabrication flaws by fatigue in combination with a decrease in reactor vessel toughness due to irradiation. Credible, postulated PTS transients that could potentially lead to reactor vessel failure must be considered to occur at the worst time in the life of the plant. The PFM methodology allows the consideration of distributions and uncertainties in flaw number and size, fluence, material properties, crack growth rate, stresses, and the effectiveness of inspections. The PFM approach leads to a conditional reactor vessel failure frequency due to a given loading condition and a prescribed inspection interval. All locations of interest in the reactor vessel can be addressed in a similar way or, as in the case of this study, a bounding approach can be used to minimize the areas receiving a detailed evaluation. A feasibility study was performed [20] that showed that this fracture mechanics and risk methodology can be used to calculate the change in the frequency of reactor vessel failure due to a change in inspection interval and to evaluate the acceptability of the associated change in risk. The impact on plant safety from the change in risk presented in this study was based on the standards for risk-informed assessment as defined by RG 1.174 [4]. WCAP- 16168-NP-A June 2008 Revision 2

3-1 3 PILOT PLANT

SUMMARY

The risk evaluations summarized in this report utilized the same pilot plants as used in the NRC PTS Risk Re-evaluation effort [8, 9]. The NRC effort to address PTS risk identified FirstEnergy Nuclear Operating Company's (FENOC's) Beaver Valley Unit 1 (BV 1), Nuclear Management Company's (NMC's) Palisades, and Duke Energy's Oconee Unit 1 (OCI) as the pilot plants. These pilot plant applications also used fleet-specific design transient data for the Combustion Engineering (CE) and Westinghouse designs. A typical generic heatup/cooldown transient was used for the Babcock & Wilcox (B&W) study. A study was also performed to determine the bounding location from among the applicable weld locations on a typical PWR reactor vessel. The results of all of these investigations are included in the following sections. 3.1 BOUNDING LOCATION The focus of the evaluations for reactor vessel inspection interval extension was on the beltline of the RV. To confirm that the beltline location represented the bounding location for the reactor vessel, all locations currently required for examination in the reactor pressure vessel (RPV) needed to be identified and considered. The beltline weld locations were found to be the bounding locations primarily due. to irradiation induced change in the fracture toughness. This was consistent with the location assumptions used to support the NRC PTS Risk Study [8, 9]. Table 3-1 summarizes the current ISI requirements for RPV inspection as identified in Table IWB-2500-1 of the ASME B&PV Code, Section XI [1]. While this table identifies all welds with Section XI inspection requirements, this report only addresses the ISI interval extension of the Category B-A and B-D welds. June 2008 WCAP- 16168-NP-A WCAP-16168-NP-A June 2008 Revision 2

3-2 Table 3-1 ASME Section XI [11 ISI Requirements for RPVs (ASME Section XI, Table IWB-2500-1) Examination Item No. RPV Location Requirement Pressure Retaining Welds in Reactor Vessel B-A B 1.10 Shell Welds Volumetric B-A B 1.11 Circumferential Volumetric B-A B 1.12 Longitudinal Volumetric B-A B 1.20 Head Welds Volumetric B-A B 1.21 Circumferential Volumetric B-A B 1.22 Meridional Volumetric B-A B 1.30 Shell-to-Flange Weld Volumetric B-A B 1.40 Head-to-Flange Weld Surface and Volumetric B-A B 1.50 Repair Welds Volumetric B-A B1.51 Beltline Region Volumetric Full Penetration Welded Nozzles in Vessels B-D B3.90 RPV Nozzle-to-Vessel Welds Volumetric B-D B3.100 RPV Nozzle Inside Radius Section Volumetric Pressure Retaining Dissimilar Metal Welds in Vessel Nozzles B-F B5.10 RPV Nozzle-to-Safe End Butt Welds, Surface and Volumnetric NPS 4 or Larger B-F B5.20 RPV Nozzle-to-Safe End Butt Welds, Surface Less Than NPS 4 B-F B5.30 RPV Nozzle-to-Safe End Socket Welds Surface Pressure Retaining Welds in Piping B-J B9.10 NPS 4 or Larger Surface and Volumetric B-J B9.11 Circumferential Welds Surface and Volumetric Welded Attachments for Vessels, Piping, Pumps and Valves B-K BO1.10 Welded Attachments Surface Interior of Reactor Vessel B-N- 1 B13.10 Vessel Interior Visual, VT-3 Welded Core Support Structures and Interior Attachments to Reactor Vessels B-N-2 B 13.50 Interior Attachments within Beltline Region Visual, VT-I B-N-2 B 13.60 Interior Attachments Beyond Beltline Region Visual, VT-3 Removable Core Support Structures B-N-3 B 13.70 Core Support Structure Visual, VT-3 WCAP-16168-NP-A June 2008 Revision 2

3-3 To confirm that the beltline was the limiting location, an assessment was performed using deterministic fracture mechanics that considered the following: Existence of 10-percent through-wall initial flaw In-service fatigue crack growth of the flaw due to normal plant operating transients

• 40 EFPY embrittlement throughout plant life
• Peak reactor vessel ID fluence assumed regardless of flaw depth, i.e., maximum embrittlement Design basis heat-up and cool-down transients

        -       500 cycles/40 years for CE NSSS
        -       200 cycles/40 years for Westinghouse NSSS 7 Weld Locations
        -       Closure Head to Flange
        -       Upper Shell to Flange.
        -       Lower Shell Transition
        -  . Bottom Head to Shell
        -       Beltline
        -       Inlet Nozzle to Safe End
        -       Outlet Nozzle to Safe End The study evaluated the effect of various ISI intervals by comparing the change in margins on ASME Code allowable flaw sizes for the respective locations. This approach was preceded by considering 3 iterative steps:

1. Select the first inspection interval, I1, based on the growth of the assumed initial flaw to a fraction of the tolerable flaw size.

2. Perform the inspection. If no defects larger than the assumed flaw size are found, the second inspection interval, 12, is the same as the first.

3. Continue subsequent inspections until actual flaws are detected that require repair or augmented inspections.

The results of the study are summarized in Figures 3-1 and 3-2. Inspection intervals were based on 10-, 20-, 30-, or 40-year inspection intervals over a 40-year plant life. Each reactor vessel location was evaluated by calculating the amount of crack extension that would occur due to fatigue crack growth over a 10-year period of operation. Each crack length was then evaluated for the maximum applied K, from a transient. The ratio of the maximum allowable KI, per the ASME Section XI [1] Appendix A criteria, to the maximum K1 applied, was used as a measure of the margin a flaw in a given location has to the acceptance criteria. Note that in Figure 3-1 the margins on the acceptance standard are greater than 1, except for the beltline region axial and circumferential flaws. This indicates that all of the flaw sizes in other locations are acceptable with varying degrees of margin. The margin less than one for the beltline WCAP- 16168-NP-A June 2008 Revision 2

3-4 locations is an indication that the assumed initial flaw size of 10-percent throughwall was greater than the acceptable flaw size. The other feature to note in Figures 3-1 and 3-2 is that, for each subsequent 10-year period that was evaluated, there was an insignificant change in the degree of margin for all of the locations. This observation was simply a reflection of the fact that the increments of fatigue crack growth of the flaws were so small that the applied K1 values were not changing. Therefore, the ratios of the applied to allowable K, did not change. Though not shown in figures 3-1 and 3-2, the reactor vessel nozzle to shell weld was also evaluated and found to be have a margin greater than that of the reactor vessel beltline axial and circumferential welds. These results confirmed that the beltline was the limiting location and that the change in fatigue crack growth increment for RPV flaws was insignificant relative to the inspection interval. While a specific number of design basis heat-up and cool-down transients was not analyzed for B&W designs in this bounding location assessment, it is reasonable to expect that the conclusions of this assessment would also be applicable to B&W plants due to similarities in the RV and NSSS designs. 10 C9 c8

         ~7 C

0.-* 6 5 4 3 _ 2

   .E-CD 0

0" cJ , Figure 3-1 Comparison to Acceptance Criteria - Minimum Margins Code Allowable June 2008 WCAP- 16168-NP-A WCAP46168-NP-A June 2008 Revision 2

3-5 Figure 3-2 Comparison to Acceptance Criteria - Minimum Margins Code Allowable 3.2 BASIS FOR RISK DETERMINATION As indicated in ASME Code Case N-691 [5], the application of risk-informed insights from PFM and risk analyses can be used to justify an increase from 10 to 20 years in the requirements of Section XI, IWB-2412 for the inspection interval, for the examination of Category B-A and B-D welds in PWR reactor vessels. The guidelines in Regulatory Guide 1.174 provide the basis for an acceptable change in risk resulting from an extension in inspection interval. As the basis for determining the change in risk, the inputs to the RV PFM and risk analyses included the following: Accident Transients and Frequency ASME Code Case N-691 [5] states that it is necessary to define a complete set of accident transients that can be postulated to realistically result in RV failure and their frequencies of occurrence. As previously mentioned, PTS events are viewed as providing the greatest challenge to PWR RPV structural integrity. For this reason, the pilot plant applications in this report used the PTS transients and frequencies from the NRC PTS Risk Study [8, 9]. As part of the NRC study, probabilistic risk assessment (PRA) models were developed for each of the pilot plants using plant specific information [22, 23, 24]. These PRA models included an event-tree analysis that defined both the sequences of events that are likely to produce a PTS challenge to RPV structural integrity and the frequency with which such events can be expected to occur. The typical sequence of concern was cool-down and depressurization due to the initiating event, followed by repressurization due to high-pressure safety injection or charging. Historically, a small-break loss-of-coolant accident (SBLOCA) with low decay heat has been the sequence identified as a major contributor WCAP- 16168-NP-A June 2008 Revision 2

3-6 to PTS risk. However, other events considered included a large break in the main steam line upstream of the main steam isolation valves, a double-ended main steam line break (MSLB) upstream of the main steam isolation valves (MSIVs), small steam line break downstream of the MSIVs, and excessive feedwater flow, all with the reactor coolant pump (RCP) shutdown and multiple failures of the operator to take remedial action. The PTS Risk Study utilized the plant specific PRA models to determine the possible sequences which could result in a PTS event for each of the pilot plants. Due to the large number of sequences which were identified, it was necessary to group (i.e., bin) sequences with like characteristics into representative transients that could later be analyzed using thermal-hydraulic codes. This resulted in 178 binned sequences for OC 1, 118 for BV 1, and 65 for Palisades. Thermal-hydraulic analyses were performed for each of these bins (i.e., representative transients) to develop time histories of temperature, pressure, and heat transfer coefficients [25]. These histories were then input into the PFM analysis to determine conditional probability of reactor vessel failure for each transient. From this analysis, it was determined that only a portion of the transients contribute to the total risk of RPV failure, while the remainder have an insignificant or zero contribution. The transients which were identified to be contributors to PTS risk were then used for the PFM analysis in the PTS study and for the pilot plant studies in this report. Consistent with the PTS Risk Study, 61 transients were analyzed for BV1, 30 for Palisades, and 55 for OC 1 in this study on the impact of extending the RV ISI interval. Details of the transients are provided in Appendix D for BV1, Appendix H for Palisades, and Appendix L for OC1. As part of the NRC PTS Risk Reevaluation Program, a study was performed to determine the applicability of the pilot plant detailed analyses to the remainder of the domestic PWR fleet. This' "Generalization" Study [26] examined the results from the three detailed pilot plant studies (BV1, Palisades, and OC I) and identified a set of plant design and operational features considered to be important in determining whether or not certain types of overcooling scenarios are significant contributors to PTS. These features were then analyzed for five additional plants and compared to the features of the pilot plants. These five plants included the following:

     "   Salem Unit 1 (Westinghouse 4-loop plant comparable to Beaver Valley Unit 1)

• TM1 Unit 1 (B&W plant comparable to Oconee Unit 1)
• Fort Calhoun (CE plant comparable to Palisades)

     " Diablo Canyon (Westinghouse 4-loop plant comparable to Beaver Valley Unit 1)

• Sequoyah Unit 1 (Westinghouse 4-loop plant comparable to Beaver Valley Unit 1)

WCAP- 16168-NP-A June 2008 Revision 2

3-7 They were chosen for the generalization study on the basis of:

     " having a high reference temperature metric (RTPTS), which reflects their potential sensitivity to PTS,

• further demonstrating the applicability of the pilot plant analyses to the remainder of the fleet for the nuclear steam supply system (NSSS) vendors, and
• including plants having different limiting materials (i.e., welds, plates, and forgings).

It was determined in the generalization study that there were no differences in plant features that from a PRA, thermal hydraulic, and PFM standpoint would be expected to cause significant differences in the through wall cracking frequencies due to the postulated PTS scenarios. It was further concluded through the generalization study that the pilot plant results at a comparable embrittlement level could be applied to the remainder of the domestic PWR fleet. Operational Transients and Cycles ASME Code Case N-691 [5] states that the operational transients that contribute to fatigue crack growth and the number of cycles occurring each year must be identified. Typically, the start-up (heat-up) and shut-down (cool-down) events are the dominant loading conditions as seen in ASME Code Section XI, Non-Mandatory Appendix A [1] calculations for fatigue crack growth of an existing flaw. For the purpose of the pilot plant studies in this report, an 80-year life for fatigue crack growth was used. This 80-year life envelopes plants seeking to obtain license extensions to 60 years and provides an additional margin of conservatism. The design basis transients for the pilot plants were reviewed and it was~determined that the greatest contributor to fatigue crack growth for the pilot plants is heat-up and cool-down. Each transient represents a full heat-up and cool-down cycle between atmospheric pressure at room temperature and full-system pressure at 100-percent power operating temperature, and thus envelopes many transients with a smaller range of conditions. For the pilot plant evaluations, 7 heat-up and cool-down cycles per year were used for Westinghouse plants (BVl) and 13 cycles were used for CE plants (Palisades) to bound all the design basis transients for the respective PWR plant designs in each fleet. Based upon available information, 12 cycles were used for Babcock and Wilcox plants. For any B&W plant using the results of this WCAP to extend the reactor vessel ISI interval from 10 to 20 years, including the pilot plant (OC 1), the fatigue crack growth for 12 heatup/cooldown transients per year will have to be verified to bound the fatigue crack growth for all design basis transients. It is important to note that most plants' operational histories indicate that they will not reach this number of design transients by end of life (EOL) (80 years). However, this calculation was performed as a bounding analysis and the number of design transients was used rather than the number of operational transients so that plants with operational histories different than those of the pilot plants would be enveloped. WCAP-16168-NP-A June 2008 Revision 2

3-8 Initial Flaw Distribution ASME Code Case N-691 [5] requires credible flaw distributions for a PWR reactor vessel. Significant work by Pacific Northwest National Laboratory (PNNL) and the NRC was performed to more completely specify the initial flaw size distributions and their densities for input into the NRC PTS Risk Study [8, 9]. This work focused on making detailed destructive and non-destructive measurements of fabrication flaws in nuclear grade RPV welds and plates. Whenever possible, this experimental evidence was used exclusively or given the greatest "weight" in establishing the flaw distributions. In cases where experimental evidence was not sufficient, physical models and expert opinion were used to supplement the experimental evidence in establishing the flaw distributions. For the NRC PTS Risk Study, flaw distributions were developed for embedded flaws in welds, plates (includes forgings), and inner surface breaking flaws. The weld flaw distribution was based on the highest densities of the Shoreham reactor vessel and the largest sizes of the PVRUF vessel. The embedded flaws are distributed evenly through the thickness of the weld. Flaws are postulated only in the same orientation as the weld. The flaw distribution represents a blended combination of weld types with 2% of the welds assumed to be repair welds, which have the largest flaw sizes. Empirical evidence to support a plate flaw distribution is much more limited than that for welds. For this reason, the density for flaws of depths less than 6mm is 10% of that for weld flaws, while the density for flaws of depth above 6mm is 2.5% of that for weld flaws. Half of the simulated flaws are assumed to be axially oriented while the other half are assumed to be circumferentially oriented. For weld and plate flaws, the pilot plant studies for the RV ISI interval extension study used the flaw distributions from the NRC PTS Risk Study directly. These densities are input into the FAVOR Code PFM analyses as flaw density files, P.dat (plate-embedded flaws) and W.dat (weld-embedded flaws). This is discussed further in the "PFM Computer Tool and Methodology" section. The inner-diameter of the RPV is clad with a thin layer of stainless steel. Lack of inter-run fusion can occur between adjacent weld beads, resulting in circumferentially oriented cracks (the cladding in the RV is deposited circumferentially). However, none of the cracks discovered in the PNNL studies had broken through the cladding layer on the inside surface of the RV. Therefore, for the NRC PTS Risk Study [8, 9], the BV1 and Palisades evaluations used multi-pass cladding with no surface breaking flaws. Multi-layer cladding is assumed to have no surface breaking flaws due to the small likelihood of two flaws aligning in two different weld layers. The OCI pilot evaluation used an assumed surface flaw completely through the cladding with a density of 1 / 1 0 0 0 th of the embedded flaws through the vessel wall. For this investigation on the impact of extending the RV ISI interval it is important to consider the effects of fatigue crack growth. Due to the fact that embedded flaws do not grow significantly due to fatigue, for the pilot plant studies, the presence of surface breaking flaws with an initial flaw depth equal to the cladding thickness was postulated. Therefore, for the pilot plant evaluations to bound all the plants of the same design, single-pass cladding was conservatively assumed. The initial flaw size and distribution was input into a fatigue crack growth and ISI analysis to determine a surface flaw density file after any inspections (ISI). Surface flaw density files were created two simulate two cases. The first case simulated inspections performed on a 10 year interval as currently required by the ASME Code. The WCAP- 16168-NP-A June 2008 Revision 2

3-9 second case simulated a single inspection performed after the first 10 years of operation with no subsequent inspection. These surface breaking flaw density files are then input into the PFM analysis as surface breaking flaw density file S.dat. The methodology for determining the flaw depth and density included in this file is described in the section on PFM and Computer Tool Methodology. Cladding details for the pilot plants are identified in Appendices B, F, and J. Fluence Distribution ASME Code Case N-691 [5] requires that the fluence distribution versus operating time, both axial and azimuthal, be based on plant-specific or bounding data for the current operating time and extrapolated as applicable to the end of the current 40 year license or for license renewal to 60 years. For the pilot plant evaluations in this report, the input fluence distributions were taken directly from the NRC PTS Risk Study [8, 9]. For the NRC PTS Risk Study a series of neutron transport calculations were performed to determine the neutron fluence on the inner-wall of the pilot plant RPVs. The modeling procedures were based on the guidance contained in NRC Reg. Guide 1.190[27]. The models incorporated pilot plant specific geometry and operating data. The fluence for E>IMeV was calculated as a function of the azimuthal and axial location in the inner reactor vessel wall. The fluence was extrapolated from the current state point to various effective full-power years (EFPYs) assuming a linear extrapolation of the most recent operating cycles. The fluences used in the RV ISI interval extension evaluations were for 60 EFPY for BVl and Palisades and for fluences at 500 EFPY for OC1 to envelope license extension. 500 EFPY were used for OCI rather than 60 EFPY because it is recognized that it is not the most embrittled RV in the B&W fleet. The use of 500 EFPY for OC I should bound the embrittlement of the most highly embrittled RV in the B&W fleet-.when evaluated against the parameters identified in Appendix A. Representative fluence maps for BVIl;ý,Palisades, and OC1 at 32 EFPY, can be found in Appendices B, F, and J, respectively. While the magnitude of the fluence on these maps correspond to 32 EFPY rather than the 60 EFPY and 500 EFPY used in the pilot plant evaluations, the contour of the fluence relative to the reactor vessel weld layout still applies, Material Fracture Toughness ASME Code Case N-691 [5] states that the material fracture toughness of the limiting beltline plates and weld materials need to be based on the following plant-specific data:

• Physical and mechanical properties of the base metal, clad, and welds (e.g., copper and nickel content) and their uncertainties.
• Initial reference nil-ductility transition temperature (RTNDT), including uncertainty 0 ARTNDT due to radiation embrittlement ,versus time and depth, including uncertainty 0 Fracture toughness versus time and depth, including uncertainty WCAP- 16168-NP-A June 2008 Revision 2

3-10 These reactor vessel material properties for the BV 1, Palisades, and OC 1 pilot plants evaluated in this report are identified in Appendices B, F, and J, respectively. Embrittlement due to irradiation in RPV steels occurs due to matrix hardening and age hardening [8, 9]. Based on the physical insights into these hardening mechanisms a relationship between material composition, irradiation-condition variables, and measurable quantities such as yield strength increase, Charpy-transition-temperature shift, and toughness-transition-temperature shift was established for the NRC PTS Risk Study [8, 9]. Furthermore, a quantitative relationship was developed from the database of Charpy shift values generated in domestic reactor surveillance programs. The Eason and Wright irradiation shift model was developed by fitting this data. This model is used in the FAVOR Code [28] for the NRC PTS Risk Study and the RV ISI interval extension pilot plant studies to calculate the shift and irradiated reference temperature as a function of time. The results of the significant work at ORNL, the NRC, and within industry to more completely specify the distribution on fracture toughness and its uncertainty for the NRC PTS Risk Study [8, 9] are included in the FAVOR Code which is used for the pilot plant studies for RV ISI interval extension. The FAVOR Code includes fracture toughness models which are based on extended databases of empirically obtained K1, and K1a data points and include the effects of the statistical bias for direct measurement of fracture toughness (Master Curve Method). Furthermore, the FAVOR Code [28] uses the latest correlation on irradiated upper shelf fracture toughness. It should be noted that along with the inspection of a weld, there is a specified amount of base metal inspected. In the FAVOR Code evaluation, if a flaw is placed within a weld that is adjacent to a more highly embrittled plate, the flaw is assigned the embrittlement characteristics of the plate rather than the weld and is assumed to fracture and propagate in the direction of the plate. The NRC has proposed that through wall cracking frequency (TWCF) can be correlated to the embrittlement index (reference temperature) of the reactor vessel components. The correlation for determining plant specific TWCF based on the plant specific data mentioned can be found in Reference 9. This correlation takes into consideration the contribution to TWCF for each of the most limiting plate, axial weld, and circumferential welds. These individual TWCF contributions are then weighted based on experimental pilot plant data and summed to determine a total reactor vessel TWCF. For application to other plant reactor vessels, the plant specific TWCF must be equal to or less than the values used for the applicable pilot plants evaluated in this report (see Appendix A) at 60 EFPY. Crack Growth Rate Correlation ASME Code Case N-691 [5] requires that the basic physical models for fatigue crack growth due to operational transients (e.g., heat-ups, cool-downs, normal plant operating changes, and reactor trips) including the effects of uncertainties, be used for the PFM analysis. Also used are the basic physical models for crack growth during these transient events (i.e., the change in applied stress intensity and the corresponding change in flaw size) for the surface breaking flaws and their uncertainties. The pilot-plant studies in this report included a probabilistic representation of the fatigue crack growth correlation for ferritic materials in water that was consistent with the previous and current models contained in Appendix A of the ASME Code, Section XI [1]. These correlations represented the behavior WCAP-16168-NP-A June 2008 Revision 2

3-11 of the ferritic reactor vessel materials for all domestic PWRs. This probabilistic representation was consistent with that used by the NRC-supported pc-PRAISE code [29] and the NRC-approved SRRA tool for piping-risk informed ISI [30]. Cladding and Residual Stresses ASME Code Case N-691 [5] requires that the residual stress distribution in welds and the cladding stress and its temperature dependence due to differential thennal expansion be considered. For the pilot plant studies for RV ISI interval extension, the residual stress distribution through the wall was taken from the NRC PTS Risk Study [8, 9] and is described in the FAVOR Code Theory Manual [28]. This distribution is shown in Figure 3-3. The stress profile was deter-mined for the NRC PTS Risk Study thorugh experiments in which a radial slot was cut in a longitudinal weld in a shell segment from an actual RPV and the deformation of the slot was measured after cutting. Finite element analysis was used to determine the residual stress profile from the measured deformations. The cladding stress used in the pilot plant studies was taken from the NRC PTS Risk Study. The cladding temperature dependence due to differential thermal expansion was based on a stress free temperature of 488°F, which is consistent with that used in the NRC PTS Risk Study [8, 9]. 4 v-U, 2 A.4 o

             -2
             -4
             -6 0               2                 4               6               8               "10, Distance from RPV Inner Surface (in.)

Figure 3-3 Weld Stress Profile Effectiveness of ISI The essential requirement for an effective volumetric examination in ASME Code Case N-691 [5] is that it be conducted in accordance with Section XI Appendix VIII [1] or RG 1.150 [2]. WCAP- 16168-NP-A June 2008 Revision 2

3-12 The following effects also need to be considered along with the change in ISI interval:

• Extent of inspection (percent coverage) 0 Probability of detection (POD) with flaw size a Repair criterion for removing flaws from service The POD should correlate to the respective examination method for the RV weld of interest.

The basis for the probability of flaw detection used in the pilot plant studies for the RV ISI interval extension was taken from studies performed at the EPRI NDE Center on the detection and sizing qualification of ISIs on the RV beltline welds [31]. Figure 3-4 shows the probability of detection with respect to flaw size used in the pilot studies in this report. 1.1 1 0.9 0.8 0.7 0.6

    -o                                                                                              -.-- Prob 0.5 0.4 0.3 0.2 0.1 0         0.1       0.2     0.3      0.4             0.5 0.6     0.7       0.8      0.9 Size (inch)

Figure 3-4 ISI Detection Probability For the pilot plant evaluations, examinations were assumed to be conducted in accordance with Section XI Appendix VIII [1], so that Figure 3-4 could be used. Flaws that were detected were assumed to be repaired with the repaired area returned to a flaw-free condition. If the quality of inspection is not as good as assumed (e.g. ISI per Regulatory Guide 1.150) or the quality of the repair is less than 100 percent, then the result would be fewer flaws found and fewer flaws removed during repair, resulting in less difference in risk from one inspection interval to another. Therefore, the pilot plant studies conservatively calculated a larger potential difference in risk by maximizing the benefits of inspection. WCAP- 16168-NP-A June 2008 Revision 2

3-13 Impact of Other ASME Code Cases on RPV Inspection While no ASME Code Cases have been found that directly overlap the actions included in ASME Code Case N-691 [5], there are related ASME Code Cases and "problem areas" that may affect implementation of the Code Case. ASME Code Cases that concern reactor vessel inspections but do not affect the applicability of the Code Case are identified in the following:. ASME Code Case N-697 [32] addresses Examination Requirements for PWR Control Rod Drive and In-Core Instrumentation Housing Welds. It adds requirements for examination of in-core instrumentation housing welds greater than 2" Nominal Pipe Size to Examination Category B-O. If these UT or surface examinations of the housing weld inner surface were conducted from inside the RPV, they could result in examination intervals incompatible with effective implementation of N-691 [5]. However, these welds are not inspected from inside the RPV and, therefore, there is no impact. A top priority in Section XI is to work with the Material Reliability Program Alloy 600 Issue Task Group to identify and incorporate changes needed in the examination of affected partial penetration and dissimilar metal welds. This could result in incompatible examination intervals for Examination Category B-F welds to reactor vessel nozzles, and dissimilar metal welds in Examination Category B-J not covered by Category B-F. A possible approach for some plants, where access permits, would be to examine these welds from the pipe outer diameter (OD) at alternate 10-year intervals, and from the inner diameter (ID) during the Case N-691 [5] examinations. ASME Code Case N-700 [33] addresses Examination Category B-K, surface examination of welded attachments. It permits examination of a single welded reactor vessel attachment each inspection interval. ASME Code Case N-648-1 [34] permits a VT-1 visual examination of a reactor vessel nozzle inner radius in lieu *f a volumetric examination. Applicability of this Code Case would not be affected by the increased examination interval. ASME Code Case N-624 [35] provides for modification of the sequence of successive examinations. The increased examination interval would be applicable. ASME Code Case N-623 [36] permits deferral to the end of the interval of shell-to-flange and head-to-flange welds of a reactor vessel. The methodology of Case N-691 [5] would not be affected by application of this Code Case. ASME Code Case N-615 [37] permits ultrasonic examination as a surface examination method for Category B-F and B-J piping welds of 4" Nominal Pipe Size and larger. It would be compatible with the increased examination interval. ASME Code Case N-613-1 [38] reduces the nozzle weld examination volume of Examination Category B-D. It would be compatible with the increased examination interval. ASME Code Case N-598 [39] provides alternatives to the required percentages of examinations each inspection period. ASME Code Case N-691 [5] would increase the length of the inspection period but would not affect the percentage requirements. WCAP-16168-NP-A June 2008 Revision 2

3-14 Probabilistic Fracture Mechanics Computer Tool and Methodology For the pilot-plant applications of the PFM methodology, the failure frequency distributions for all postulated flaws in the RV were calculated using the latest version (06.1) of the FAVOR code [28]. The Fracture Analysis of Vessels - Oak Ridge (FAVOR) computer program was developed as part of the NRC PTS Risk Study [8, 9]. It is a program that performs a probabilisticanalysis of a nuclear reactor pressure vessel when subjected to events in which the reactor pressure vessel wall is exposed to time-varying thermal-hydraulic boundary conditions. To run the FAVOR code, 3 modules (FAVLOAD, FAVPFM and FAVPOST) and various input files were required as shown in Figure 3-5. In the NRC PTS Risk Study [8, 9], the effects of fatigue crack growth and ISI were not considered. However, to perform the risk evaluation for changing the inspection interval from 10 to 20 years, these effects were quantified. Program PROBSBFD (Probabilistic Surface Breaking Flaw Density) was developed to include these effects by modifying the surface-breaking flaw input file to FAVOR (S.dat) as shown in Figure 3-5. The first module in FAVOR is the load module, FAVLOAD, where the thermal-hydraulic time histories are input for the dominant PTS transients. For each PTS transient, deterministic calculations are performed to produce a load-definition input file for FAVPFM (FAVPFS is also used in this analysis). These load-definition files include time-dependent, through-wall temperature profiles, through-wall circumferential and axial stress profiles, and stress-intensity factors for a range of axially and circumferentially oriented embedded and inner surface-breaking flaw geometries (both infinite and finite-length). The FAVPFS module in Figure 3-5 is a modification of the FAVPFM module, which is the second module contained in the FAVOR code that was used in the NRC PTS risk study. The modification allows FAVPFS to have a 4 times finer depth distribution for surface breaking flaws in S.dat. The FAVPFS FAVOR module uses the input flaw distributions (e.g., S.dat, W.dat, and P.dat), the loads for the PTS events from the FAVLOAD module and fluence/chemistry input data at 60 EFPY (effective full-power years) to calculate the initiation and failure probabilities for each PTS transient. The FAVPOST post-processor is the third module in FAVOR. It combines the distributions of initiating frequencies for the dominant PTS transients with the results of the PFM analysis (performed with the FAVPFS module) to generate probability distributions for the frequencies of reactor vessel crack initiation and reactor vessel failure. This module also generates statistical information on these distributions and the distributions for the conditional probabilities of reactor vessel crack initiation and failure for each PTS transient included in the risk analysis. WCAP- 16168-NP-A June 2008 Revision 2

3-15 Time History Input for Dominant PTS Transients Loads FCG & ISI Time History for HUCD Subroutines Input for HUCD Loads for PTS Events W.dat & P.dat Fluence / Chemistry Input at 60 EFPY Frequencies of Dominant Transients (PTS Representative Plants) PTS Failure

• FAVPFS is FAVPFM modified to have 4 times Frequency Flaw Distribution Files finer depth distribution for surface breaking Distribution P.dat Embedded Plate Flaws flaws on S.dat. S.dat Surface Breaking Flaws W.dat Embedded Weld Flaws Figure 3-5 Software and Data Flow for Pilot Plant Analyses WCAP-16168-NP-A June 2008 Revision 2

3-16 The PROBSBFD code was specifically developed for the RV ISI interval extension project and verified in accordance with the Westinghouse Quality Assurance requirements. This program utilizes the Westinghouse Structural Reliability and Risk Assessment (SRRA) library program, which provides standard input and output, including probabilistic analysis capabilities (e.g., random number generation and importance sampling). PROBSBFD was used to develop 1000 random surface breaking flaw distributions that fed into the FAVPFS module via an input file (S.dat is the default name). The loads were determined using the FAVLOAD module, for the input with time histories of temperature, pressure, and heat transfer characteristics for the operational transients (e.g., heat-up and cool-down) that could grow the initial flaws by means of fatigue. The applied stress intensity factor (K) at various times and various depths through the reactor vessel wall were taken directly from the FAVLOAD output file and input into PROBSBFD (FAVLOADS.dat for PROBSBFD). The beneficial effects of ISI were modeled in the same way as in the NRC's probabilistic analysis code pc-PRAISE [29] and the SRRA Code [30] used with the PWROG/ASME piping risk-informed in-service inspection (RI-ISI) program. Specifically, only the flaws not detected during an ISI exam, at 10 years for example, remained. For example, if the probability of detection for the first inspection was 90 percent, then the flaw density was effectively multiplied by 10 percent for input to the next iteration. The effects of subsequent inspections, where the probability of detection was increased because the flaw was bigger (see Figure 3-4), could be either cumulative or independent. For each of the 1000 simulations performed by PROBSBFD, the initial flaw depth and density were defined. Four aspect ratios, 2, 6, 10, and infinite, were considered. For each time-step and flaw-aspect ratio, the effects of ISI, the stress intensity factors, and the random crack growth were calculated. After all the time steps were completed, the distribution of flaw densities by depth and aspect ratio were written to a surface-breaking, flaw-distribution input file for FAVPFS, which was in the same format as the default S.dat file (see Figure 3-5). WCAP- 16168-NP-A June 2008 Revision 2

3-17 3.3 RESULTS FOR THE WESTINGHOUSE PILOT PLANT: BV1 Reactor vessel failure frequencies were calculated for BVI for two cases corresponding to the two surface flaw density files discussed in the section on "Initial Flaw Distribution". These cases were referred to as "IS1 Every 10 Years" and "10-year ISI Only". As the names imply, the "ISI Every 10 Years" case simulates the current ASME Code required inspections while the "10-year ISI Only" case simulates a discontinuation of inspections after the first 10-year ISI. Statistically, the difference between the mean failure frequencies for the "ISI Every 10 Years" case and the "10-year ISI Only" case is insignificant. This is due to the fact that the difference between the mean values is less than the standard error for each of the cases. However, to calculate a change in risk for comparison to regulatory guidelines, a change in failure frequency was conservatively calculated based on the difference between an "Upper Bound" and a "Lower Bound." The Lower Bound was determined by subtracting 2 times the standard error as reported by FAVPOST from the mean value of the "ISI Every 10 Years" case. The Upper Bound was determined by adding 2 times the standard error as reported by FAVPOST to the mean value of the "10-Year ISI Only" case. Elimination of ISI after the first 10-year ISI for the BVI RPV results in a difference in failure (through-wall flaw) frequency of less than 1E-09. A summary of the results of the evaluation are included in Table 3-2. The results reflect the maximum statistically calculated value for the potential change in risk at a number of reactor vessel simulations at which the Monte Carlo statistical analysis has reached a stable solution. The difference between the Upper Bound and Lower Bound represents the bounding difference between the 10-year inspection interval currently applicable under ASME criteria and elimination of all future inspections following an inspection within the first 10 years of operation. This change in failure frequency is acceptable per the regulatory guidance discussed in Section 4.1. Transient input was based on design basis transients and the transients used in the NRC PTS Risk Study [8, 9]. The input data included consideration of postulated life extension to 60 EFPY. The FAVPOST outputs for the cases presented in Table 3-2 are presented in Appendix E. Table 3-2 BV1 Reactor Vessel Failure Frequency Results 10-Year ISI Only (Mean Value / Standard Error) 5.04E-09 / 2.54E-10 Upper Bound Value 5.55E-09 ISI Every 10 Years (Mean Value / Standard Error) 5.23E-09 / 3.12E-10 Lower Bound Value 4.6 1E-09 Bounding Difference in Risk 9.4E-10 The mean effects of fatigue crack growth and ISI on the surface breaking flaw density for 1000 simulations are shown in Figures 3-6 and 3-7. These figures plot the flaw density as a function of the flaw depth for the cases of one initial 10-year ISI, a 10-year ISI interval, and a 20-year ISI interval. These plots display the results for the 10-to-I and infinite aspect ratio sizes. The PROBSBFD outputs used to generate these plots are included in Appendix C. The crack growth and density reduction due to ISI WCAP-16168-NP-A June 2008 Revision 2

3-18 would both be reduced for the flaw length-to-depth aspect ratios of 2-to-I and 6-to-I also considered in the pilot plant study. U C V (U

                                                                                      --    10 Year ISI Only 0*

0- ISI Every 10 Years

                                                                                       - -   SI Every 20 Years 0.

CU 2.00% 2.20% 2.40% 2.60% 2.80% 3.00% 3.20% 3.40% Flaw Depth (Percent of Wall Thickness) Figure 3-6 Growth of Flaws with an Aspect Ratio of 10 for BV1 WCAP- 16168-NP-A June 2008 Revision 2

3-19 U 1.000E-07

                                                                                       -    10 Year SI Only u) 1.OOOE-08                                                                         -   - ISI] Every 10 Years
                                                                                   -     I- SI Every 20 Years 1.000E-09 1.000E-10 1.OO0E-11 1.000E-12 2.00%   2.10%   2.20%    2.30%     2.40%     2.50%    2.60% 2.70% 2.80%

Flaw Depth (Percent of Wall Thickness) Figure 3-7 Growth of Flaws with an Infinite Aspect Ratio for BVl WCAP- 16168-NP-A June 2008 Revision 2

3-20 3.4 RESULTS FOR THE COMBUSTION ENGINEERING PILOT PLANT: PALISADES Reactor vessel failure frequencies were calculated for Palisades for two cases corresponding to the two surface flaw density files discussed in the section on "Initial Flaw Distribution". These cases were referred to as "ISI Every 10 Years" and "10-year ISI Only". As the names imply, the "IS1 Every 10 Years" case simulates the current ASME Code required inspections while the "10-year ISI Only" case simulates a discontinuation of inspections after the first 10-year ISI. While the failure frequency for the "IS1 Every 10 Years" case is higher than the "10-Year ISI Only" case, statistically, the difference between the mean failure frequencies for the "ISI Every 10 Years" case and the "10-year ISI Only" case is insignificant. This is due to the fact that the difference between the mean values is less than the standard error for each of the cases. However, to calculate a change in risk for comparison to regulatory guidelines, a bounding change in failure frequency was calculated based on the difference between an "Upper Bound" and a "Lower Bound." The Lower Bound was determined by subtracting 2 times the standard error as reported by FAVPOST from the mean value of the "ISI Every 10 Years" case. The Upper Bound was determined by adding 2 times the standard error as reported by FAVPOST to the mean. value of the "10-Year ISI Only" case. Elimination of ISI after the first 10-year ISI for the Palisades RPV results in a bounding difference in failure (through-wall flaw) frequency of less than 1.81E-08. A summary of the results of the evaluation are included in Table 3-3. The results reflect the maximum statistically calculated value for the potential change in risk at a number of reactor vessel simulations at which the Monte Carlo statistical analysis has reached a stable solution. The difference between the Upper Bound and Lower Bound represents the bounding difference between the 10-year inspection interval currently applicable under ASME criteria and elimination of all future inspections following an inspection within the first 10 years of operation. This change in failure frequency is acceptable per the regulatory guidance discussed in Section 4.1. Transient input was based on design basis transients and the transients used in the NRC PTS Risk Study [8, 9]. The input data included consideration of postulated life extension to 60 EFPY The FAVPOST outputs for the cases presented in Table 3-3 are presented in Appendix I. Table 3-3 Palisades Reactor Vessel Failure Frequency Results 10-Year ISI Only (Mean Value / Standard Error) 7.62E-08/4.08E-09 Upper Bound Value 8.44E-08 ISI Every 10 Years (Mean Value / Standard Error) 7.39E-08/3.80E-09 Lower Bound Value 6.63E-08 Bounding Difference in Risk 1.81E-08 The mean effects of fatigue crack growth and ISI on the surface breaking flaw density for 1000 simulations are shown in Figures 3-8 and 3-9. These figures plot the flaw density as a function of the flaw depth for the cases of 1 initial 10-year ISI, a 10-year ISI interval, and a 20-year ISI interval. These plots display the results for,the of 10-to-I and infinite aspect ratio sizes. The PROBSBFD outputs used to WCAP- 16168-NP-A June 2008 Revision 2

3-21 generate these plots are included in Appendix G. The crack growth and density reduction due to ISI would both be reduced for the flaw length-to-depth aspect ratios of 2-to-I and 6-to-I also considered in the pilot plant study. 1.00E-04 1.OOE-05 1.00E-06 1.OOE-07 C

  -1.00E-08
                      /

10- \ ---- 10 Year ISI Only S1.OOE-09 l---S Every 10 Years Q- - SI Every 20 Years IL A, 1.OOE LL.

              ,'7                           \

1.OOE-1 1

                        --U.

1.OOE-12 - "" N 1.00E-13 I 1.OOE-14 " 3.00% 3.50% 4.00% 4.50% 5.00% 5.50% 6.00% 6.50% 7.00% 7.50% Flaw Depth (Percent of Wall Thickness) Figure 3-8 Growth of Flaws with an Aspect Ratio of 10 for Palisades June 2008 WCAP- 16168-NP-A WCAP-16168-NP-A June 2008 Revision 2

3-22 1.00E-04 1.OOE-05 1.00E-06 1.00E-07

  ,C) 1.00E-08
                                                                                                 --- 10 Year ISI Only Le 1.00E-09              ,                                                                       -   - ISI Every 10 Years
                             \A,-                                                                    -    SI Every 20 Years CL,*1,00E-10                     \-

1.O0E-11 1.OOE-12 N A,. 1.00E-13 i 1.OOE-14 . - 3.00% 3.50% 4.00% 4.50% 5.00% 5.50% 6.00% 6.50% 7.00% 7.50% 8.00% Flaw Depth (Percent of Wall Thickness) Figure 3-9 Growth of Flaws with an Infinite Aspect Ratio for Palisades WCAP- 16168-NP-A June 2008 Revision 2

3-23 3.5 RESULTS FOR THE BABCOCK AND WILCOX PILOT PLANT: OCI Reactor vessel failure frequencies were calculated for OC 1 for two cases corresponding to the two surface flaw density files discussed in the section on "Initial Flaw Distribution". These cases were referred to as "ISI Every 10 Years" and "10-year ISI Only". As the names imply, the "ISI Every 10 Years" case simulates the current ASME Code required inspections while the "10-year ISI Only" case simulates a discontinuation of inspections after the first 10-year ISI. While the failure frequency for the "ISI Every 10 Years" case is higher than the "10-Year ISI Only" case, statistically, the difference between the mean failure frequencies for the "ISI Every 10 Years" case and the "10-year ISI Only" case is insignificant. This is due to the fact that the difference between the mean values is less than the standard error for each of the cases. However, to calculate a change in risk for comparison to regulatory guidelines, a bounding change in failure frequency was calculated based on the difference between an "Upper Bound" and a "Lower Bound." The Lower Bound was determined by subtracting 2 times the standard error as reported by FAVPOST from the mean value of the "ISI Every 10 Years" case. The Upper Bound was determined by adding 2 times the standard error as reported by FAVPOST to the mean value of the "10-Year ISI Only" case. Elimination of ISI after the first 10-year ISI for the OCI RPV results in a difference in failure (through-wall flaw) frequency of 1.26E-08. A summary of the results of the evaluation are included in Table 3-4. The results reflect the maximum statistically calculated value for the potential change in risk at a number of reactor vessel simulations at which the Monte Carlo statistical analysis has reached a stable solution. The difference between the Upper Bound and Lower Bound represents the bounding difference between the 10-year inspection interval currently applicable under ASME criteria and elimination of all future inspections following an inspection within the first 10 years of operation. This change in failure frequency is acceptable per the regulatory guidance discussed in Section 4.1. Transient input was based on design basis transients and the transients used in the NRC PTS Risk Study [8, 9]. The input data included consideration of postulated life extension to 60 EFPY The FAVPOST outputs for the cases presented in Table 3-4 are presented in Appendix M. Table 3-4 OC1 Reactor Vessel Failure Frequency Results 10-Year ISI Only (Mean Value / Standard Error) 3.1 IE-08/2.55E-09 Upper Bound Value 3.62E-08 ISI Every 10 Years (Mean Value / Standard Error) 2.62E-08/1.28E-09 Lower Bound Value 2.36E-08 Bounding Difference in Risk 1.26E-08 The mean effects of fatigue crack growth and ISI on the surface breaking flaw density for 1000 simulations are shown in Figures 3-10 and 3-11. These figures plot the flaw density as a function of the flaw depth for the cases of 1 initial 10-year ISI, a 10-year ISI interval, and a 20-year ISI interval. These plots display the results for the 10-to-i and infinite aspect ratio sizes. The PROBSBFD outputs used to generate these plots are included in Appendix K. The crack growth and density reduction due to ISI WCAP-16168-NP-A June 2008 Revision 2

3-24 would both be reduced for the flaw length-to-depth aspect ratios of 2-to-I and 6-to-1 also considered in the pilot plant study. Figure 3-10 Growth of Flaws with an Aspect Ratio of 10 for OC1 WCAP-16168-NP-A June 2008 Revision 2

3-25 1.OOE-04 C a) m 1.00E-08

                                                                                            -      10 Year ISI Only
     ".                                                                                   -   - - -SI Every 10 Years
                                                                                         -A   -    SI Every 20 Years C\ i.00E-09 1.00E-10 1.00E-11                             A.

1.OOE-12 1.OOE-13 3-00% 3.50% 4.00% 4.50% 5.00% 5.50% 6.00% 6.50% 7.00% Flaw Depth (Percent of Wall Thickness) Figure 3-11 Growth of Flaws with an Infinite Aspect Ratio for OCt June 2008 WCAP- 161 68-NP-A WCAP-16168-NP-A June 2008 Revision 2

4-1 4 RISK ASSESSMENT The quantitative risk assessment discussed below shows that extending the inspection interval from 10 to a maximum of 20 years has an acceptably small impact on risk (core damage frequency [CDF] and large early release frequency [LERF]), i.e., that it is within the bounds of RG 1.174 [4]. A discussion on the requirements of RG 1.174 is included. 4.1 RISK-INFORMED REGULATORY GUIDE 1.174 METHODOLOGY The NRC has developed a risk-informed regulatory framework. The NRC definition of risk-informed regulation is: "insights derived from probabilistic risk assessments are used in combination with deterministic system and engineering analysis to focus licensee and regulatory attention on issues colmmensurate with their importance to safety." The NRC issued RG 1.174, An Approach for Using ProbabilisticRisk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the CurrentLicensing Basis [4]. In addition, the NRC issued application-specific RGs and Standard Review Plans (SRPs):

• RG-1.175 [40] and SRP Chapter 3.9.7, related to in-service testing (IST) programs
• RG-1.176 [41], related to Graded Quality Assurance (GQA) programs
• RG- 1.177 [42] and SRP Chapter 16.1, related to Technical Specifications
• RG-l.178 [44] and SRP-3.9.8, related to ISI of piping programs These RG and SRP chapters provide guidance in their respective application-specific subject areas to reactor licensees and the NRC staff regarding the submittal and review of risk-informed proposals that would change the licensing basis for a power reactor facility.

Regulatory Guide 1.174 Basic Steps The approach described in RG 1.174 was used in each of the application-specific RGs/SRPs, and has 4 basic steps as shown in Figure 4-1. The four basic steps are discussed below. Step 1: Define the Proposed Change This element includes identifying:

1. Those aspects of the plant's licensing bases that may be affected by the change.

2. All systems, structures, and components (SSCs), procedures, and activities that are covered by the change and consider the original reasons for inclusion of each program requirement.

3. Any engineering studies, methods, codes, applicable plant-specific and industry data and operational experience, PRA findings, and research and analysis results relevant to the proposed change.

WCAP- 16168-NP-A June 2008 Revision 2

4-2 Traditional Analysis PRA] PRA I\I / I /

                                      /                 /
                                    /                 /

I I

                                                /

I / - I / - -

                              'I          I--

Figure 4-1 Basic Steps in (Principal Elements of) Risk-Informed, Plant-Specific Decision Making (from NRC RG 1.174) Step 2: Perform Engineering Analysis This element includes performing the evaluation to show that the fundamental safety principles on which the plant design was based are not compromised (defense-in-depth attributes are maintained) and that sufficient safety margins are maintained. The engineering analysis includes both traditional deterministic analysis and probabilistic risk assessment (PRA). The evaluation of risk impact should also assess the expected change in CDF and LERF, including a treatment of uncertainties. The results from the traditional analysis and the PRA must be considered in an integrated manner when making a decision. Step 3: Define Implementation and Monitoring Program This element's goal is to assess SSC performance under the proposed change by establishing performance monitoring strategies to confirm assumptions and analyses that were conducted to justify the change. This is to ensure that no unexpected adverse safety degradation occurs because of the changes. Decisions concerning implementation of changes should be made in light of the uncertainty associated with the results of the evaluation. A monitoring program should have measurable parameters, objective criteria, and parameters that provide an early indication of problems before becoming a safety concern. In addition, the monitoring program should include a cause determination and corrective action plan. June 2008 WCAP- 16168-NP-A WCAP-16168-NP-A June 2008 Revision 2

4-3 Step 4: Submit Proposed Change This element includes:

1. Carefully reviewing the proposed change in order to determine the appropriate form of the change request.

2. Assuring that information required by the relevant regulation(s) in support of the request is developed.

3. Preparing and submitting the request in accordance with relevant procedural requirements.

Regulatory Guide 1.174 Fundamental Safety Principles Five fundamental safety principles are described that each application for a change must meet. These are shown in Figure 4-2, and are discussed below. Change isconsistent with defense-in-depth Change meets current philosophy. Maintain sufficient regulations unless itis explicitly related to a safety margins. requested exemption or rule change. DIntegrated Decisionmaking Proposed increases in

                   *Use performance-CDF or risk are small measurement                                        and are consistent with strategies to monitoor                             the Commission's Safety the change.                                        Goal Policy Statement.

Figure 4-2 Principles of Risk-Informed Regulation (from NRC RG 1.174) Principle 1: Change meets current regulationsunless it is explicitly relatedto a requested exemption or rule change. The proposed change is evaluated against the current regulations (including the general design criteria) to either identify where changes are proposed to the current regulations (e.g., Technical Specification, license conditions, and FSAR), or where additional information may be required to meet the current regulations. WCAP-16168-NP-A June 2008 Revision 2

4-4 Principle2: Change is consistent with defense-in-depth philosophy. Defense-in-depth has traditionally been applied in reactor design and operation to provide a multiple means to accomplish safety functions and prevent the release of radioactive material. As defined in RG 1.174 [4], defense-in-depth is maintained by assuring that:

• A reasonable balance among prevention of core damage, prevention of containment failure, and consequence mitigation is preserved.
• Over-reliance on programmatic activities to compensate for weaknesses in plant design is avoided.
• System redundancy, independence, and diversity are preserved cormnensurate with the expected frequency and consequences to the system (e.g., no risk outliers).
• Defenses against potential common cause failures are preserved and the potential for introduction of new common cause failure mechanisms is assessed.
• Independence of barriers is not degraded (the barriers are identified as the fuel cladding, reactor coolant pressure boundary, and containment structure).

0 Defenses against human errors are preserved. Defense-in-depth philosophy is not expected to change unless:

• A significant increase in the existing challenges to the integrity of the barriers occurs.

0 The probability of failure of each barrier changes significantly. 0 New or additional failure dependencies are introduced that increase the likelihood of failure compared to the existing conditions.

• The overall redundancy and diversity in the barriers changes.

Principle3: Maintain sufficient safety margins. Safety margins must also be maintained. As described in RG 1.174, sufficient safety margins are maintained by assuring that:

• Codes and standards, or alternatives proposed for use by the NRC, are met.
• Safety analysis acceptance criteria in the licensing basis (e.g., FSARs, supporting analyses) are met, or proposed revisions provide sufficient margin to account for analysis and data uncertainty!

WCAP-16168-NP-A June 2008 Revision 2

4-5 Principle4: Proposedincreases in CDF or risk are small and are consistent with the Commission's Sqfety Goal Policy Statement. To evaluate the proposed change with regard to a possible increase in risk, the risk assessment should be of sufficient quality to evaluate the change. The expected change in CDF and LERF are evaluated to address this principle. An assessment of the uncertainties associated with the evaluation is conducted. Additional qualitative assessments are also performed. There are two acceptance guidelines, one for CDF and one for LERF, both of which should be used. The guidelines for CDF are: If the application can be clearly shown to result in a decrease in CDF, the change will be considered to have satisfied the relevant principle of risk-informed regulation with respect to CDF. When the calculated increase in CDF is very small, which is taken as being less than 10-6 per reactor year, the change will be considered regardless of whether there is a calculation of the total CDE When the calculated increase in CDF is in the range of 10-6 per reactor year to 10-5 per reactor year, applications will be considered only if it can be reasonably shown that the total CDF is less than 10-4 per reactor year. Applications that result in increases to CDF above 10-5 per reactor year would not normally be considered. The guidelines for LERF are: If the application can be clearly shown to result in a decrease in LERF, the change will be considered to have satisfied the relevant principle of risk-informed regulation with respect to LERF. When the calculated increase in LERF is very small, which is taken as being less than 10-7 per reactor year, the change will be considered regardless of whether there is a calculation of the total LERF. When the calculated increase in LERF is in the range of 10-7 per reactor year to 10-6 per reactor year, applications will be considered only if it can be reasonably shown that the total LERF is less than 10-5 per reactor year. Applications that result in increases to LERF above 10-6 per reactor year would not normally be considered. These guidelines are intended to provide assurance that proposed increases in CDF and LERF are small and are consistent with the intent of the Commission's Safety Goal Policy Statement. WCAP-16168-NP-A June 2008 Revision 2

4-6 Principle 5: Use performance-measurementstrategies to monitor the change. Performance-based implementation and monitoring strategies are also addressed as part of the key elements of the evaluation as described previously. Risk-Acceptance Criteria for Analysis For the purposes of this bounding analysis of the risk impact of the proposed change in RV inspection frequency, the following criteria are applied with respect to Principle 4 (small change in risk):

• Change in CDF < 1 x 10-6 per reactor year
• Change in LERF < 1 x 10- per reactor year These values are selected so that the proposed change may be later considered on a plant-specific basis regardless of the plant's baseline CDF and LERF.

To conservatively simplify these acceptance criteria, it will be assumed that through-wall crack growth is equivalent to reactor vessel failure, and that reactor vessel failure results in both core damage and a large early release. It is also conservatively assumed that the conditional probability of a large early release given core damage is 1.0 (See Section 4.3). Therefore, the simplified conservative/bounding acceptance criterion becomes: Increase in frequency of Change in CDF through-wall crack <x 10-l per

                        =    Change in LERF growth due to increase in             reactor year inspection interval 4.2      FAILURE MODES AND EFFECTS Failure Modes The failure mode of concern was thermal fatigue crack growth due typical plant operation. The growth of an existing undetected fabrication flaw in the RV base metal, cladding, or weld metal was assumed to reach a critical size that would lead to reactor vessel through-wall fracture if a PTS-type transient would occur.

Failure Effects A through-wall flaw failure of the RV was assumed to result in core damage and a large early release. June 2008 WCAP- 16168-NP-A WCAP-16168-NP-A June 2008 Revision 2

4-7 4.3 CORE DAMAGE RISK EVALUATION The objective of the risk assessment was to evaluate the core damage risk from the extension of the examination of the RV relative to other plant risk contributors through a qualitative and quantitative evaluation. NRC RG 1.174 [4] provided the basis for this evaluation as well as the acceptance guidelines to make a change to the current licensing basis. Risk was defined as the combination of likelihood of an event and severity of consequences of an event. Therefore, the following two questions were addressed:

• What was the likelihood of the event?
• What would the consequences be?

The following sections describe the likelihood and postulated consequences. The likelihood and consequences were then combined in the risk calculation and the results of the evaluation are presented in this report. What is the Likelihood of the Event? The likelihood of the event was addressed by identifying the plant transients or operational events that might lead to failure of the RV, and estimating the frequency of these events. What are the Consequences? The consequences were defined in terms of the CDF and LERF risk metrics. For this evaluation, the conditional core damage probability given the failure of the RV was assumed to be 1.0 (no credit for safety system actuation to mitigate the consequences of the failure). Since this was intended as a bounding assessment, it was also conservatively assumed that the conditional probability of a large early release given core damage for this scenario is 1.0 (i.e., no credit for consequence mitigation via the containment and related systems). Note that this was a simplifying assumption, and a specific mechanism for LERF was not implied or defined here. WCAP-16168-NP-A June 2008 Revision 2

4-8 Risk Calculation For this evaluation, the CDF and LERF were calculated by: N TWCF = LERF = CDF= IE, *CPFI where: TWCF = Through wall cracking frequency CDF = Core damage frequency from all vessel failures due to PTS events (events per year) LERF = Large early release frequency from all vessel failures due to PTS events (events per year) lE, = Initiating event frequency (events per year) for a given PTS transient, i CPFj = Conditional probability of reactor vessel failure for a given PTS transient i, and N = The total number of postulated PTS transients for a given plant. The transient initiating frequency distributions were identified in the NRC PTS Risk Study [8, 9] and are included in Appendices D, H, and L for the pilot plants. The probability of failure was calculated by the FAVPFS module of FAVOR. The FAVPOST module of FAVOR combined the transient initiating frequency distribution with the reactor vessel conditional failure probability distribution to determine a reactor vessel failure frequency distribution for each transient. From these failure frequency distributions, FAVPOST determined a mean reactor vessel failure frequency. In addition to this mean failure frequency a standard error was reported. To account for uncertainties, Upper and Lower Bounds are determined. The Upper Bound was determined by adding 2 times the standard error from the "10-Year ISI-Only" case. The Lower Bound was determined by subtracting 2 times the standard error from the "ISI Every 10 Years" case. The change in reactor vessel failure frequency was determined by subtracting the Lower Bound from the Upper Bound. The mean reactor vessel failure frequencies, Upper and Lower Bounds, and change in failure frequency are given in Sections 3.2 and 3.3. As previously stated, reactor vessel failure results in core damage which results in large early release. Therefore, the large early release frequencies were equal to the reactor vessel failure frequencies. The large early release frequencies, Upper and Lower Bounds, and change in large early release frequency are summarized in Table 4-1, based on FAVOR 06.1 evaluations. Table 4-1 Large Early Release Frequencies BV1 Palisades OCi (per year) (per year) (per year) 10-Year ISI Only 5.04E-09 7.62E-08 3.11 E-08 Upper Bound 5.55E-09 8.44E-08 3.62E-08 ISI Every 10 Years 5.23E-09 7.39E-08 2.62E-08 Lower Bound 4.61E-09 6.63E-08 2.36E-08 Bounding Change in Large 9.37E-l0 1.81E-08 1.26E-08 Early Release Frequency WCAP-16168-NP-A June 2008 Revision 2

4-9 Risk Results and Conclusions The -analysis described above demonstrates that changes in CDF and LERF do not exceed the NRC's RG-1.174 [4] acceptance guidelines for a small change in CDF and LERF (<10-6 per year for CDF, <107 per year for LERF). As part of this evaluation, the key principles identified in RG- 1.174 were reviewed and the responses based on the evaluation are provided in Table 4-2. This evaluation concluded that extension of the RV in-service examination from 10 to 20 years would not be expected to result in an unacceptable increase in risk. Given this outcome, and the fact that other key principles listed in RG- 1.174 continue to be met, the proposed change in inspection interval from 10 to 20 years is acceptable. Table 4-2 Evaluation with Respect to Regulatory Guide 1.174 14] Key Principles Key Principles Evaluation Response Change meets current regulations unless it is Change to current RG 1.150 [2] requirements is proposed. explicitly related to a requested exemption or rule change. Change is consistent with defense-in-depth Potential for failure of the RV is acceptably small during normal philosophy, or accident conditions, and does not threaten plant barriers. See the discussion below for additional information on defense in depth. Maintain sufficient safety margins. No safety analysis margins are changed. Proposed increases in CDF or risk are small and Proposed increase in risk is estimated to be acceptably small. are consstent with the Commission's Safety Goal Policy Statement. Useperformance-measurement strategies to NDE examinations still conducted, but on less frequent basis not monitor the change. to exceed 20 years. Other indications of potential degradation of RV are available (e.g., foreign experience and periodic testing with visual examinations) Defense-in-Depth While the results presented in this report demonstrate that the contribution of eliminating future inspections after the initial 10 year ISI meets prescribed regulatory criteria for assessing risk, the proposed course of action is to extend the inspection interval requirements from 10 to 20 years while not eliminating any portion of the current inspection requirements. This provides additional margin for defense-in-depth and contributes directly toward maintaining plant safety. Extending the RV ISI interval does not imply that generic degradation mechanisms will be ignored for 20 years. (With the number of PWR nuclear power plants in operation in the U.S. and globally, a sampling of plants inevitably undergo examinations in a given year.) This provides for early detection of WCAP- 16168-NP-A June 2008 Revision 2

4-10 any potential emerging generic degradation mechanisms, and would permit the industry to react with more frequent examinations if needed. In addition, it must be recognized that all reactor coolant pressure boundary failures occurring to date have been identified as a result of leakage, and were discovered by visual examination. The proposed RV ISI interval extension does not alter the visual examination interval. The reactor vessel would undergo, as a minimum, the Section XI Examination Category B-P pressure tests and visual examinations conducted at the end of each refueling before plant start-up, as well as leak tests with visual examinations that precede each start-up following maintenance or repair activities. Page 4-4 identifies from Regulatory Guide 1. 174 that: "Defense-in-depth philosophy is not expected to change unless:

     "   A significant increase in the existing challenges to the integrity of the barriers occurs.

• The probability of failure of each barrier changes significantly.
• New or additional failure dependencies are introduced that increase the likelihood of failure compared to the existing conditions.

     "   The overall redundancy and diversity in the barriers changes."

The extension in inspection interval will not result in any of the changes identified above. Also identified on page 4-4 are six elements for maintaining defense in depth. Due to the fact that the interval extension will not result in any of the changes identified above, it is expected that the defense in depth elements listed on page 4-4 will not be impacted. Additional assessment of the impact on each of the defense-in-depth elements from page 4-4 is provided below:

• A reasonable balance among prevention of core damage, prevention of containment failure, and consequence mitigation is preserved:

The proposed increase in inspection would not cause an increased reliance on any of the identified elements. Therefore, the interval increase would not change the existing balance among prevention of core damage, prevention of containment failure, and consequence mitigation.

• Over-reliance on programmatic activities to compensate for weaknesses in plant design is avoided:

The change in inspection interval does not change the robustness of the vessel design in any way. It is because of this robustness that the inspection interval can be doubled with no significant change in failure frequency.

      "  System redundancy, independence, and diversity are preserved commensurate with the expected frequency and consequences to the system (e.g., no risk outliers):

The proposed increase in inspection interval does not impact system redundancy, independence, or diversity in any way since it is not changing the plant design or how it is operated. WCAP- 16168-NP-A June 2008 Revision 2

4-11

• Defenses against potential common cause failures are preserved and the potential for introduction of new common cause failure mechanisms is assessed:

The proposed increase in inspection interval does not impact any defenses against any comrumon cause failures and there is no reason to expect the introduction of any new common cause failure mechanisms. This requirement applies to multiple active components. There is only one reactor vessel per plant and it is a passive component.

• Independence of barriers is not degraded (the barriers are identified as the fuel cladding, reactor coolant pressure boundary, and containment structure):

The increase in inspection interval does change the relationship between the barriers in anyway and therefore does not degrade the independence of the barriers. The change in inspection interval does not change the robustness of the vessel design in any way. It is because of this robustness that the inspection interval can be doubled with no significant change in failure frequency. Defenses against human errors are preserved: The increase in the RV inspection interval does not impact any defenses against human errors in any way. The increase in the inspection interval reduces the frequency for which the lower internals need to be removed. Reducing this frequency reduces the possibility for human error and damaging the core. WCAP- 16168-NP-A June 2008 Revision 2

5-1 5 CONCLUSIONS Based on the results of this analysis, it is concluded that:

1. The beltline is the most limiting region for the evaluation of risk.

2. RV inspections performed to date have not detected any significant flaws.

3. Crack extension due to fatigue crack growth during service is small.

4. The man-rem exposure can be reduced by extending the inspection interval.

5. The failure frequencies for PWR RVs due to the dominant PTS transients are well below 10-7 per year.

6. The change in risk meets the RG 1.174 [4] acceptance guidelines for a small change in LERF.

7. The increase in the RV ISI interval from 10 to 20 years satisfies all the RG 1.174 criteria, including other considerations, such as defense-in-depth.

Based on the above conclusions, the ASME Section XI [1] 10-year inspection interval for examination categories B-A and B-D welds in PWR RVs can be extended to 20 years. In-service inspection intervals of 20 years for FENOC's Beaver Valley Unit 1, NMC's Palisades, and Duke Energy's Oconee Unit 1 are acceptable for implementation. The methodology in WCAP-16168-NP Revision 1 is applicable to plants other than the pilot plants by confirming the applicability of the parameters in Appendix A on a plant specific basis. Since the 10 year inspection interval is required by Section XI, IWB-2412, as codified in 10 CFR 50.55a, an exemption request must be submitted and approved by the NRC to extend the inspection interval to 20 years, unless 10 CFR 50.55a is amended to incorporate ASME Code Case N-691. June 2008 WCAP-16 168-NP-A WCAP-16168-NP-A June 2008 Revision 2

6-1 6 REFERENCES

1. ASME Boiler andPressure Vessel Code, Section XI, 1989 Edition with the 1989 Addenda up to and including the 2004 Edition with the 2005 Addenda, American Society of Mechanical Engineers, New York.

2. NRC Regulatory Guide 1.150, Revision 1, Ultrasonic Testing of Reactor Vessel Welds During PreserviceandInserviceExaminations,February 1983 (ADAMS Accession # ML003739996).

3. 10CFR50.55a, Codes and Standards, 36 FR 11424, June 12, 1971.

4. NRC Regulatory Guide 1.174, Revision 1, An Approachfor Using ProbabilisticRisk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, November 2002 (ADAMS Accession #ML023240437).

5. ASME Boiler and Pressure Vessel Code, Code Case N-691, "Application of Risk-Informed Insights to Increase the Inspection Interval for Pressurized Water Reactor Vessels," Section XI, Division 1, 2003.

6. WCAP-14572 Revision 1-NP-A, "Westinghouse Owners Group Application of Risk-Informed Methods to Piping Inservice Inspection Topical Report," February 1999.

7. WCAP- 15666-A Revision 1, "Extension of Reactor Coolant Pump Motor Flywheel Examination," October 2003.

8. NUREG-1806, Technical Basisfor Revision of the PressurizedThermal Shock (PTS) Screening Limit in the PTS Rule (IOCFR50.61):Summary Report, May 2006 (ADAMS Accession

     #ML061580318).

9. NUREG- 1874, Recommended Screening Limits for PressurizedThermal Shock (PTS), March 2007 (ADAMS Accession #ML070860156).

10. ASME Boiler andPressure Vessel Code, Code Case N-481, "Alternative Examination Requirements for Cast Austenitic Pump Casings," Section XI, Division 1, 1990

11. ASME Boiler and Pressure Vessel Code, Code Case N-560, "Alternative Examination Requirements for Class 1, Category B-J Piping Welds," Section XI, Division 1, 1996.

12. ASME Boiler and Pressure Vessel Code, Code Case N-577, "Risk-Informed Requirements for Class 1, 2, and 3 Piping, Method A," Section XI, Division 1, 1997.

13. ASME Boiler and Pressure Vessel Code, Code Case N-578, "Risk-Informed Requirements for Class 1,2, and 3 Piping, Method B," Section XI, Division 1, 1997.

WCAP- 16168-NP-A June 2008 Revision 2

6-2

14. ASME Boiler and Pressure Vessel Code, Code Case N-613, "Ultrasonic Examination of Full Penetration Nozzles in Vessels, Examination Category B-D Item No's B3.10 and B3.90, Reactor Vessel-To-Nozzle Welds, Fig. IWB-2500-7 (a), (b), and (c)," Section XI, Division 1, 1998.

15. ASME Boiler andPressure Vessel Code, Code Case N-552, "Alternative Methods - Qualification for Nozzle Inside Radius Section from Outside Surface," Section XI, Division 1, 1995.

16. ASME Boiler and Pressure Vessel Code, Code Case N-610, "Alternative Reference Stress Intensity Factor (KIR) Curve for Class Components," Section III, Division 1, 1998.

17. ASME Boiler and Pressure Vessel Code, Code Case N-409, "Procedure and Personnel Qualification for Ultrasonic Detection and Sizing of Intergranular Stress Corrosion Cracking in Austenitic Piping Welds," Section XI, Division 1, 1984.

18. ASME Boiler andPressure Vessel Code, Code Case N-512, "Assessment of Reactor Vessels with Low Upper Shelf Charpy Impact Energy Levels," Section XI, Division 1, 1993.

19. ASME Boiler andPressure Vessel Code, Code Case N-557, "In-Place Dry Annealing of a PWR Nuclear Reactor Vessel," Section XI, Division 1, 1996.

20. WCAP-15912, Extension of Inservice Inspection Interval for PWR Reactor Vessels, November 2002.

21. EPRI NP-719-SR, Flaw Evaluation Procedures - Background and Application of ASME Section XI Appendix A, Electric Power Research Institute, 1978.

22. A.M. Kolaczkowski, SAIC, et. al., "Oconee Pressurized Thermal Shock (PTS) Probabilistic Risk Assessment (PRA)," September 28, 2004 (ADAMS Accession #ML042880452).

23. D.W. Whitehead, SNL, et. al., "Beaver Valley Pressurized Thermal Shock (PTS) Probabilistic Risk Assessment (PRA)," September 28, 2004 (ADAMS Accession #ML042880454).

24. D.W. Whithead, et. al., "Palisades Pressurized Thermal Shock (PTS) Probabilistic Risk Assessment (PRA)," October 6, 2004 (ADAMS Accession #ML042880473).

25. NUREG- 1809, Thermal Hydraulic Evaluationof PressurizedThermal Shock, 2005 (ADAMS Accession #ML050390012).

26. D.W. Whitehead, SNL, et. al., Generalizationof Plant-Specific PressurizedThermal Shock (PTS)

    ,'RiskResults to AdditionalPlants, December 14, 2004 (ADAMS Accession #ML042880482).

27. NRC Regulatory Guide 1.190, "CalculationalandDosimetry Methodsfor DeterminingPressure Vessel Neutron Fluence," March 2001 (ADAMS Accession #ML010890301).

WCAP- 16168-NP-A June 2008 Revision 2

6-3

28. Draft NUREG/CR (ORNL/TM-2007/0030), FractureAnalysis Qf Vessels - Oak Ridge FAVOR, v06. 1, Computer Code: Theory and Implementation ofAlgorithms, Methods, and Correlations, issued by Oak Ridge National Laboratory, 2007.

29. NUREG/CR-5864, Theoretical and User's Manual for pc-PRAISE, A ProbabilisticFracture Mechanics Computer Code for PipingReliability Analysis, 1999.

30. WCAP-14572 Supplement 1, Westinghouse StructuralReliability and Risk Assessment (SRRA)

Model for PipingRisk-Informed In-Service Inspection, Rev. 1-NP-A, February 1999.

31. BWRVIP- 108: B WR Vessel and InternalsProject, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii, EPRI, Palo Alto, CA 1003557, 2002.

32. ASME Boiler andPressure Vessel Code, Code Case N-697, "Pressurized Water Reactor (PWR)

Examination and Alternative Examination Requirements for Pressure Retaining Welds in Control Rod Drive and Instrumentation Nozzle Housings," Section XI, Division 1, 2003.

33. ASME Boiler and Pressure Vessel Code, Code Case N-700, "Alternative Rules for Selection of Classes 1, 2, and 3 Vessel Welded Attachments for Examination," Section XI, Division 1, 2003.

34. ASME Boiler and Pressure Vessel Code, Code Case N-648, "Alternative Requirements for Inner Radius Examination of Class 1 Reactor Vessel Nozzles," Section XI, Division 1, 2001.

35. ASME Boiler andPressure Vessel Code, Code Case N-624,"Successive Inspections," Section XI, Division 1, 1998.

36. ASME Boiler and Pressure Vessel Code, Code Case N-623, "Deferral of Inspections of Shell-to-Flange and Head-to-Flange Welds of a Reactor Vessel," Section XI, Division 1, 1998.

37. ASME Boiler and Pressure Vessel Code, Code Case N-615, "Ultrasonic Examination as a Surface Examination Method for Category B-F and B-J Piping Welds," 2001.

38. ASME Boiler and Pressure Vessel Code, Code Case N-613, "Ultrasonic Examination of Full Penetration Nozzles, Examination Category B-D, Item Nos. B3.10 and B3.90, Reactor Vessel-to-Nozzle Welds," 2002.

39. ASME Boiler and Pressure Vessel Code, Code Case N-598, "Alternative Requirements to Required Percentages of Examinations," Section XI, Division 1, 1995.

40. NRC Regulatory Guide 1.175, An Approachfor Plant-Specific,Risk-Informed Decisionmaking:

Inservice Testing, August 1998 (ADAMS Accession #ML003740149).

41. NRC Regulatory Guide 1.176, An Approach for Plant-Specific,Risk-Informed Decisionmaking:

Graded Quality Assurance, August 1998 (ADAMS Accession #ML003740172). WCAP- 16168-NP-A June 2008 Revision 2

6-4

42. NRC Regulatory Guide 1.177, An Approachfor Plant-Specific, Risk-informedDecisionmaking:

Technical Specifications,August 1998 (ADAMS Accession #ML003740176).

43. NRC Regulatory Guide 1.178, An ApproachforPlant-Specific,Risk-InformedDecisionmaking Inservice Inspection of Piping,August 1998 (ADAMS Accession #ML032510128).

44. NRC Memorandum, Thadani to Collins, Technical Basis for Revision of the Pressurized Thermal Shock (PTS) Screening Criteria in the PTS Rule (10CFR50.61), December 31, 2002 (ADAMS Accession #ML030090626).

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A-I APPENDIX A PLANT SPECIFIC APPLICATION WCAP-16168-NP Revision 2 describes the methodology used to demonstrate the feasibility of extending the reactor vessel inspection interval required by the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section XI, as supplemented by Nuclear Regulatory Commission (NRC) Regulatory Guide 1.150. This methodology was used to perform risk analysis for pilot plants representing the Westinghouse and Combustion Engineering designs. It is an extension of work done as part of the NRC PTS Risk Study. Table A-I identifies critical parameters to be used to determine if the pilot plant evaluations documented in this report bound a plant specific application. If the plant-specific parameter is not bounded by the pilot plant analysis, additional evaluations or sensitivity studies may be required to support the use of the pilot plant risk studies. Additional information relative to plant specific reactor vessel inspection is to be provided in Table A-2. Information required to calculate the plant specific through wall cracking frequency is to be provided in Table A-3. Additional information and requirements are provided following each table. Examples of plant specific use of these tables for Wolf Creek and Waterford 3 are contained in Appendices A-1 and A-2 respectively. Table A-1 Critical Parameters for the Application of the Bounding Analysis Additional Evaluation Pilot Plant Plant Specific Required? Parameter Basis Basis (Y/N) Dominant PTS Transients in the NRC PTS Risk Study are applicable Through Wall Cracking Frequency (TWCF) Frequency and Severity of Design Basis Transients Cladding Layers (Single/Multiple) For each of the four parameters in Table A-1, the licensee must identify the pilot plant basis and the plant specific basis. If the plant specific basis is not bounded by the pilot plant basis, additional evaluation is required. Dominant PTS Transients in the NRC PTS Risk Study are applicable The transients evaluated in the WCAP pilot plant analyses were the PTS transients from the NRC PTS Risk Re-evaluation (NUREG-1806 or NUREG-1874). For this parameter, it is necessary to demonstrate that the PTS transients used in the pilot plant analyses are applicable for a specific plant. As stated in the last paragraph in Section 3.2.1 of NUREG-1874, the "PTS Generalization Study demonstrates that risk-significant PTS transients do not have any appreciable plant-specific differences within the population of PWRs currently operating in the United States." Based on this statement, plant specific analyses are not needed for this criterion. A licensee may enter the "PTS Generalization Study" in Table A-1 as the basis for the applicability of the pilot plant PTS transients to their specific plant. WCAP-16168-NP-A June 2008 Revision 2

A-2 Through Wall Cracking Frequency (TWCF) Each licensee shall calculate their plant specific TWCF95-TOTAL value using the correlations in NUREG-1874. The calculated plant specific TWCF95-TOTAL value must be lower than the applicable pilot plant TWCF95-TOTAL value calculated using the correlations in NUREG-1874. The TWCF is essentially a measure of the embrittlement of the reactor vessel components weighted by their contribution to PTS failure. By demonstrating that the pilot plant has a higher TWCF95-TOTAL value it follows that the pilot plant change in risk calculation is bounding of that for the specific plant. The pilot plant TWCF 95-TOTAL values calculated using the NUREG- 1874 correlations depend upon the applicable pilot plant design: Westinghouse: Beaver Valley Unit 1: 1.76E-08 Events per year CE: Palisades: 3.16E-07 Events per year B&W: Oconee Unit 1: 4.42E-07 Events per year The applicable correlations from NUREG-l1874 are as follows: TWCF95-TOTAL = aAWTWCF 95-AW + apLTWCF 95-PL + acwTWCF95_cw + aFoTWCF95_FO Where, ax is determined as follows: IfRTmAx-xx< 625°R, then (x = 2.5 If 625R < RTmAx-< 875°R then a = 2.5 - (1.5/250)(RTm*x_., - 625) If RTMAx.> 875'R, then cx = I and TWCF 95.- ,r values are calculated as follows: TWCF 95_AW = exp { 5.5198*ln(R TmAxAw- 616) - 40.542}*,6 TWCF9 5-PL exp{23.737*ln(RTqAxpL - 300) - 162.38}*,8 TWCF95 -cW = exp {9.13 63*ln(R TMAxcw - 616) - 65.066}*,6 TWCF9 5 -FO = exp{23.737*ln(R Tma- 1 Fo - 300) - 162.38}*l +...

                                +il*{1.3 x 10-137*1 0.185*RTMAX°F0}*g ri is equal to "0" for ring forged vessels fabricated compliant with Regulatory Guide 1.43 and equal to "1" for ring forged vessels not fabricated compliant with Regulatory Guide 1.43.
        ,8 is determined as follows irrespective of the set of TWCF formulas used.

If TwALL < 91/2 -in, then ,8=l If 91/2 < TWALL < 111/2/2 -in, then 86= I+ 8 (TwALL- 9/2) If TWALL > 11/2 -in, then fl= 17 WCAP- 16168-NP-A June 2008 Revision 2

A-3 Where TWALL is the thickness of the RPV wall (inches), including the cladding. RTMAx-xx values are calculated in degrees Rankin ('R) as follows: nAWFL (RTadi-aw(i) +/- ATadia`(i)(OtFL) RTMAXAWxV- MAX MAX AWFL(i) j(RTdJT(V)+ AT 3 o0_pIC,)(*,FL)) Where: nAWFL is the number of axial weld fusion lines in the beltline region of the vessel, i is a counter that ranges from 1 to nAWFL, btFL is the maximum fluence occurring on the vessel ID along a particular axial weld fusion line, RTadaw(i) is the unirradiated RTNDT of the weld adjacent to the ith axial weld fusion line, RT -di*(.i) is the unirradiated RTNDT of the plate adjacent to the ith axial weld fusion line AT odg-aO(i) is the shift in the Charpy V-Notch 30-foot-pound (ft-lb) energy produced by irradiation to btFL of the weld adjacent to the ith axial weld fusion line, and ATaoj-Pl(i) is the shift in the Charpy V-Notch 30-foot-pound (ft-lb) energy produced by irradiation due to tFL of the plate adjacent to the ith axial weld fusion line. nPL RTMAPLRTMX-L*-X*'ZXeXt'NDT~u) MAX[RTrPL(i) + T*30 ATPL(i) \rMAX (tPL(i))] Where: nPL is the number of plates in the beltline region of the vessel, i is a counter that ranges from 1 to npL,

               )tPL(i)

AUX is the maximum fluence occurring over the vessel ID occupied by a particular plate, RTPLND() is the unirradiated RTNDT of a particular plate, and AT 30L() is the shift in the Charpy V-Notch 30-foot-pound (ft-lb) energy produced by irradiation to PiA) of a particular plate. RTMAX-FO = MAX [RT- ) + AT O(V)FO(i) i~=1 Where: nFO is the number of forgings in the beltline region of the vessel, i is a counter that ranges from 1 to nFO Ot is the maximum fluence occurring over the vessel ID occupied by a particular forging, RTFO(i) is the unirradiated RTNDT of a particular forging, and AT3*() is the shift in the Charpy V-Notch 30-foot-pound (ft-lb) energy produced by irradiation to Ot of a particular forging. WCAP- 16168-NP-A June 2008 Revision

A-4

                                                               'RTa'di'cn(i) +~   di-cw(i)(,iY ncwFLNDT(u)
                                              ]k/[A'**    *a~aj-1(i)            30        ()tF  ))

AcwFL(i +A RTMAX-CW

                        -MAX~MAXcFi)                                ;T u    +A(t    P        FL Where:

Where: i=1. **'R TadJ- '(, + AT3jdJ fo(i)(OtFL)) I nCWFL is the number of circumferential weld fusion lines in the beltline region of the vessel, I is a counter that ranges from 1 to nCWFL, bt FL is the maximum fluence occurring on the vessel ID along a particular circumferential weld fusion line, RT adj-c ,(i) is the unirradiated RTNDT of the weld adjacent to the ith circumferential weld fusion line, RTad-pl(i) is the unirradiated RTNDT of the plate adjacent to the ith circumferential weld fusion line (if there is no adjacent plate this term is ignored), RT adj-fo(i) is the unirradiated RTNDT of the forging adjacent to the ith circumferential weld fusion line (if there is no adjacent forging this term is ignored), AT3ofcw(i) is the shift in the Charpy V-Notch 30-foot-pound (ft-ib) energy produced by irradiation due to OfFL of the weld adjacent, to the itb circumferential weld fusion line, AT 3o-p4 is the shift in the Charpy V-Notch 30-foot-pound (ft-lb) energy produced by irradiation to ObtFL of the plate adjacent to the ith axial weld fusion line (if there is no adjacent plate this term is ignored), and AT ad-folil is the shift in the Charpy V-Notch 30-foot-pound (ft-lb) energy produced by irradiation to fbtFL of the forging adjacent to the ith axial weld fusion line (if there is no adjacent forging this term is ignored). The AT30 shift shall be determined using the latest approved methodology in Regulatory Guide 1.99 or other NRC-approved methodology (Equations in Section 3.5.2 of NUREG-1874 were used to calculate pilot plant values). All material properties used to determine the TWCF 9 5-TOTAL value shall be documented in Table A-3. The plant specific TWCF9S5TOTAL value shall be re-evaluated any time fluence is re-projected to increase as a result of core reloading, core configuration, power uprating, or when a surveillance capsule is pulled from the reactor vessel. In the case that the calculated plant specific TWCF95-TOTAL value exceeds the pilot plant value (at any time the evaluation is performed), additional evaluation shall be performed to demonstrate that the change-in-risk associated with the extension in the inservice inspection is acceptable. Frequency and Severity of Design Basis Transients It is necessary to demonstrate that the amount of fatigue crack growth considered in the pilot plant analyses is bounding for a specific plant. Since the amount of fatigue crack growth used in the pilot plant analyses was calculated based on the design basis transients, a comparison of design basis transients shall WCAP-16168-NP-A June 2008 Revision 2

A-5 be performed to ensure that the assumed number of heatup-coodown transients per year is also applicable for the specific plant. The pilot plant basis depends on the applicable design: Westinghouse: 7 Cooldowns per year CE: 13 Cooldowns per year B&W: 12 Cooldowns per year (assumed) ForCE and Westinghouse designs, if the specific plant has operated within its design basis, the amount of fatigue crack growth in the pilot plant will be bounding. If a plant has operated outside of its design basis, an additional evaluation is required to demonstrate that the pilot plant fatigue crack growth is still bounding. For B&W plants, an evaluation must be performed to show that the fatigue crack growth used in the pilot plant analysis is bounding of that which may occur at the specific plant. If the plant specific fatigue crack growth is projected to be greater than pilot plant fatigue crack growth considered, additional evaluation is required to demonstrate that the change-in-risk associated with the extension in the inservice inspection is acceptable. Cladding Layers The pilot plant analyses were performed assuming a single layer of cladding because the probability of having a surface breaking flaw in multi-layer cladding is much less than that of single-layer cladding. The licensee shall identify whether their reactor vessel was fabricated with single or multi-layer cladding. Since the pilot plant analyses were performed with single-layer, all plants are bounded by this parameter. Table A-2 Additional Information Pertaining to the Reactor Vessel Inspection Inspection methodology: Numberof past inspections: Number of indications found: Proposed inspection schedule for balance of plant life: Table A-2 is to be completed with plant specific reactor vessel inservice inspection data to meet the requirements stated as follows. Inspection Methodology The licensee shall identify the methodology used for the most recent inservice inspection performed on the ASME Category B-A and B-D welds that are included in this evaluation. Typically the methodology used will be either Regulatory Guide 1.150 or ASME Section XI Appendix VIII. WCAP-16168-NP-A June 2008 Revision 2

A-6 Number of Past Inspections The licensee shall identify the number of past inspections that that have been perfonried on the ASME Category B-A and B-D welds that are included in this evaluation. Number of Indications Found The licensee shall identify the number of flaws of concern (as defined in part c of the requirements above) found during the most recent inservice inspection. Proposed Inspection Schedule for Balance of Plant Life The licensee shall identify the years in which future inservice inspections will be performed. The dates identified must be within plus or minus one refueling cycle of the dates identified in the implementation plan provided to the NRC in PWR Owners Group letter OG-06-356, "Plan for Plant Specific Implementation of Extended Inservice Inspection Interval per WCAP-16 168-NP, Revision 1, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval." MUHP 5097-99, Task 2059," dated October 31, 2006. If the licensee identifies dates in Table A-2 that are not within plus or minus one refueling cycle of the dates in the plan, the licensee will be required to have additional discussion with the NRC Staff. After implementation of the extended interval the following requirement must be met: All data on embeddedflaws of concern with a through-wall extent (TWE) greater than 0.1 inch shall be provided to NRC within one year of completing the next vessel beltline inservice inspection per ASME Section XI, Appendix VIII, Supplement 4. Forpotential vessel failure due to PTS, embeddedflaws of concern are axially orientedplanarflaws in the vessel beltline within the inner 12.5% (1/81h) of the vessel wall thickness. An assessment of the inservice inspection results relative to the flaw distributionsused in the pilot plant analyses shall also be provided. This assessment shall be performed in accordance with the requirements of Section (d)in the final published version of the voluntary PTS rule, 10 CFR 50.61(a). WCAP-16168-NP-A June 2008 Revision 2

A-7 Table A-3 Details of TWCF Calculation Inputs Reactor Coolant System Temperature, TRcs[°F]: Twan [inches]: P CuMaterialNeutron/cm Ni Mn Un-Irradiated 2 Fluence [1019 Region/Component Material , Description [wt%] [wt%] [wt%] [wt%] RTNDT(u) [F] E>1 MeV] Outputs Methodology Used to Calculate AT 30 : Region # Fluence [1019 AT3o (From RTMAx-xx [R] Neutron/cm 2, i (flux) ATF] TWCF 95-xx E> 1 MeV] [OF] Above) Limiting Axial Weld - AW Limiting Plate - PL Forging - FO Circumferential Weld - CW TWCF95-TOTAL (cLAwTWCF 95-AW + ctPLTWCF95-PL + ctcwTWCF 9s cw + CaFoTWCF 9 5-Fo): The infornation used to calculate the through wall cracking frequency (TWCF) shall be included in Table A-3. The fields are defined in the TWCF correlation definition for Table A-1. Additional rows should be added to the input section for each beltline region/component where region/component is a particular weld, plate, or forging. Refer to Appendices A- I and A-2 for examples of how Table A-3 are to be completed. WCAP- 16168-NP-A June 2008 Revision 2

A-8 APPENDIX A-I WOLF CREEK PLANT IMPLEMENTATION EXAMPLE Table 1 Critical Parameters for Application of Bounding Analysis Additional Evaluation Required? Parameter Pilot Plant Basis Plant Specific Basis (Y/N) Dominant Pressurized Thermal Shock NRC PTS Risk Study PTS Generalization Study No (PTS) Transients in the NRC PTS Risk (Reference 8) (Reference 26) Study are applicable Through Wall Cracking Frequency 1.76-08 Events per year 2.5 1E-13 Events per year No Frequency and Severity of Design 7 heatup/cooldowns per Bounded by 7 No Basis Transients year heatup/cooldowns per year Cladding Layers (Single/Multiple) Single Single (assumed) No Table 2 Additional Information Pertaining to Reactor Vessel Inspection Inspection methodology: Past inspections have been performed to Regulatory Guide 1.150. Inspections performed during RF13 and RF 14 were also performed to ASME Section XI Appendix VIII. Number of past inspections: - Category B-A welds (reactor vessel): 2 inspections, RF8 - Spring 1996 and RF14 - Spring 2005 with the exception of weld RV-101-121 which was also inspected in RF2 - Spring 1987 and RF10 - Spring 1999

                                     -   Category B-A welds (closure head): 2 inspections, Interval 1 examinations in RFl - Fall 1986, RF4 - Spring 1990, and RF6
                                        - Spring 1993. Interval 2 examinations were performed in RF9 - Fall 1997, RFll - Fall 2000, and RF13 - Fall 2003. 2 welds were examined each outage.
                                     - Category B-D welds (outlet nozzles): 3 inspections RF3 - Fall 1988, RF8 - Spring 1996, RF14 - Spring 2005
                                     -  Category B-D welds (inlet nozzles): 2 inspections, RF8 -

Spring 1996, RF14 - Spring 2005 Number of indications found: Zero reportable indications have been found to date. Any' recordable indications have been acceptable per ASME Section XI IWB-3500. No flaws of concern were detected. Proposed inspection schedule for Third inservice inspection currently scheduled for 2015. The third balance of plant life: inservice inspection is proposed to be performed in 2025. The fourth inservice inspection interval is proposed to be performed in 2045. WCAP- 16168-NP-A June 2008 Revision 2

A-9 Table 3 Details of TWCF Calculation Inputs 0 Reactor Coolant System Temperature, TRCs[ F]: 550 Tw,.j [inches]: 8.62 Fluence [1019 Region/Component Material Cu Ni P Mn Un-Irradiated Neutron/cm 2, Description [wt%] [wt%] [wt%] [wt%] RTNDT(u) [OF] E>1 MeV] 1 Lower Shell Plate A 533B 0.070 0.620 0.008 1.35 40.0 3.90 2 Lower Shell Plate A 533B 0.090 0.670 0.009 1.35 0.0 3.90 3 Lower Shell Plate A 533B 0.060 0.640 0.008 1.35 10.0 3.90 4 Intermediate Shell Plate A 533B 0.040 0.640 0.007 1.35 -20.0 3.90 5 Intermediate Shell Plate A 533B 0.050 0.630 0.007 1.35 -20.0 3.90 6 Intermediate Shell Plate A 533B 0.040 0.660 0.008 1.35 -20.0 3.90 7 Lower Shell Axial Weld Linde 0091 0.040 0.080 0.005 1.35 -50.0 1.76 8 Lower Shell Axial Weld Linde 0091 0.040 0.080 0.005 1.35 -50.0 3.42 9 Lower Shell Axial Weld Linde 0091 0.040 0.080 0.005 1.35 -50.0 3.42 10 Inter. Shell Axial Weld Linde 0091 0.040 0.080 0.005 1.35 -50.0 1.76 11 Inter. Shell Axial Weld Linde 0091 0.040 0.080 0.005 1.35 -50.0 3.42 12 Inter. Shell Axial Weld Linde 0091 0.040 0.080 0.005 1.35 -50.0 3.42 13 Inter. - Lower Circ Weld Linde 124 0.040 0.080 0.007 1.35 -50.0 3.90 Outputs Methodology Used to Calculate AT 30 : NUREG-1874 Region # Fluence [1019 AT30 (From RTMAx-xx [R] Neutron/cm 2, 4 (flux) [T30 [OF] Above) E> 1 MeV] Limiting Axial Weld - AW 1 561.96 3.42 1.81E+10 62.27 2.47E-18 Limiting Plate - PL 1 565.05 3.90 2.06E+10 65.36 1.OOE-13 Forging - FO N/A N/A N/A N/A N/A N/A Circumferential Weld - CW 1 565.05 3.90 2.06E+10 65.36 5.52E-29 TWCF95-TOTAL (cLAwTWCF 95-AW + ItPLTWCF 95-PL + axcwTWCF 95-cw + cLFoTWCF95-FO): 2.51E-13 WCAP- 16168-NP-A June 2008 Revision 2

A-10 APPENDIX A-2 WATERFORD 3 PLANT IMPLEMENTATION EXAMPLE Table 1 Critical Parameters for Application of Bounding Analysis Additional Evaluation Required? Parameter Pilot Plant Basis Plant Specific Basis (Y/N) Dominant Pressurized Thermal Shock NRC PTS Risk Re- PTS Generalization Study No (PTS) Transients in the NRC PTS Risk Evaluation (Reference 8) (Reference 26) Study are applicable Through Wall Cracking Frequency 3.16E-07 Events per year 2.87E-14 Events per year No Frequency and Severity of Design 13 heatup/cooldowns per Bounded by 13 No Basis Transients year heatup/cooldowns per year Cladding Layers (Single/Multiple) Single Single No Table 2 Additional Information Pertaining to Reactor Vessel Inspection Inspection methodology: Past inspections have been performed to Regulatory Guide 1.150 Number of past inspections: - Category B-A welds (reactor vessel): 1 inspection - 1995, with the exception of weld 0 1-020 which was also inspected in 1988.

                                      -  Category B-A welds (closure head): 4 inspections with 3 welds inspected 1986, 3 welds inspected 1989, 1 weld inspected 1994, 3 welds inspected 2000
                                      -  Category B-D welds (outlet nozzles): 2 inspections - 1988 and 1995, with the exception of weld 01-021 which was also inspected in 1989.
                                      - Category B-D welds (inlet nozzles): 1 inspection - 1995 Number of indications found:        Zero reportable indications have been found to date. Any recordable indications have been acceptable per ASME Section XI IWB-3500. No flaws of concern have been detected.

Proposed inspection schedule for Second inservice inspection currently scheduled for Spring 2008. balance of plant life: The second inservice inspection is proposed to be performed in 2015. The third inservice inspection is proposed to be performed in 2035. WCAP- 16168-NP-A June 2008 Revision 2

A-lI Table 3 Details of TWCF Calculation Inputs Reactor Coolant System Temperature, TRCS[°F]: 553 Twall [inches]: 8.62 Fluence [1019 Region/Component Cu Ni P Mn Un-Irradiated Feun/c 2, Description [wt%] [wt%] [wt%] [wt%] RTNDT(u) [°F] E>I MeV] 1 Lower Shell Plate A 533B 0.030 0.580 0.005 1.35 22.0 4.49 2 Lower Shell Plate A 533B 0.030 0.620 0.006 1.35 -15.0 4.49 3 Lower Shell Plate A 533B 0.030 0.620 0.007 1.35 -10.0 4.49 4 Intermediate Shell Plate A 533B 0.020 0.700 0.007 1.35 -42.0 4.49 5 Intermediate Shell Plate A 533B 0.020 0.710 0.004 1.35 -30.0 4.49 6 Intermediate Shell Plate A 533B 0.020 0.670 0.006 1.35 -50.0 4.49 7 Lower Shell Axial Weld Linde 0091 0.030 0.200 0.007 1.35 -80.0 4.49 8 Lower Shell Axial Weld Linde 0091 0.030 0.200 0.007 1.35 -80.0 4.49 9 Lower Shell Axial Weld Linde 0091 0.030 0.200 0.007 1.35 -80.0 4.49 10 Inter. Shell Axial Weld E 8018 0.020 0.960 0.010 1.35 -60.0 4.50 11 Inter. Shell Axial Weld E 8018 0.020 0.960 0.010 1.35 -60.0 4.50 12 Inter. Shell Axial Weld E 8018 0.020 0.960 0.010 1.35 -60.0 4.50 13 Inter. - Lower Circ. Weld Linde 0091 0.050 0.160 0.008 1.35 -70.0 4.49 Outputs Methodology Used to Calculate AT 30 : NUREG-1874 Region # Fluence [1019 AT30 (From RTMAx-XX [R] Neutron/cm 2, 4 (flux) 0 [ F] TWCFsxx Above) E>1 MeV][ Limiting Axial Weld - AW 1 541.91 4.49 2.37E10 57.93 2.47E-18 Limiting Plate - PL 1 541.91 4.49 2.37E10 57.93 5.52E-29 Forging - FO N/A N/A N/A N/A N/A N/A Circumferential Weld - CW 1 541.91 4.49 2.37E10 57.93 1.15E-14 TWCF95-TOTAL ((XAwTWCF 9 5-AW + CQpLTWCF 9 5-PL + oQcwTWCF 95-cw + cLFOTWCF 9 5-FO): 2.87E-14 WCAP- 16168-NP-A June 2008 Revision 2

B-1 APPENDIX B INPUTS FOR THE BEAVER VALLEY UNIT 1 PILOT PLANT EVALUATION June 2008 WCAP-16168-NP-A WCAP-16168-NP-A June 2008 Revision 2

B-2 A summary of the NDE inspection history based on Regulatory Guide 1.150 and pertinent input data for BV1 is as follows:

1. Number of ISIs performed (relative to initial pre-service and 10-year interval inspections) for full penetration Category B-A and B-D reactor vessel welds assuming all of the candidate welds were inspected: 2 (covering all welds of the specified categories).

2. The inspections performed covered: A total of 34 items. 15 Category B-A items had coverage of

       <90%. 1 Category B-A item had coverage > 90% but <100%. 6 Category B-A items had coverage of 100%. 6 Category B-D items had coverage of 90% and 6 had coverage of 100%.

3. Number of indications found during the most recent inservice inspection: 42 This number includes consideration of the following additional information.

a. Indications found that were reportable: 0

b. Indications found that were within acceptable limits: 42 C. Indications/anomalies currently being monitored: 0

4. Full penetration relief requests for the RV were submitted and accepted by the NRC for 15 items.

5. Fluence distribution at inside surface of RV beltline until end of life (EOL): see Figure B-1 taken from the NRC PTS Risk Study [44], Figure 4.2.

WCAP- 16168-NP-A June 2008 Revision 2

B-3

                                  .1 K

INM g. Figure B-1 Rollout Diagram of Beltline Materials and Representative Fluence Maps for BV1

6. Reactor vessel cladding details:

a. Thickness: 0.156 inches

b. Material properties are identified in Table B-1. This is consistent with the NRC PTS Risk Study [8, 9]:

WCAP-16168-NP-A June 2008 Revision 2

B-4 Table B-1 Cladding Material Properties Specific Young's Thermal Thermal Heat Modulus of Expansion Conductivity (Btu/LBM- Elasticity Coefficient 0F (KI1S' Density Poisson's (Bu/r-t-F) Temperature (Btu/hr-ft-°F) F) (KS1) (OF- (LBM/ft3) Ratio (OF) "K C "E".c a. p V" 0 - 489 .3 68 - - 22045.7 489 .3 70 8.1 0.1158 - 489 .3 100 8.4 0.1185 - 8.55E-06 489 .3 150 8.6 0.1196 - 8.67E-06 489 .3 200 8.8 0.1208 - 8.79E-06 489 .3 250 9.1 0.1232 - 8.9E-06 489 .3 300 9.4 0.1256 - 9.OE-06 489 .. 3 302 - - 20160.2 - 489 .3 350 9.6 0.1258 - 9.1E-06 489 .3 400 9.9 0.1281 9.19E-06 489 .3 450 10.1 0.1291 - 9.28E-06 489 .3 482 - - 18419.8 - 489 .3 500 10.4 0.1305 - 9.37E-06 489 .3 550 10.6 0.1306 - 9.45E-06 489 .3 600 10.9 0.1327 - 9.53E-06 489 .3 650 11.1 0.1335 - 9.61E-06 489 .3 700 11.4 0.1348 - 9.69E-06 489 .3 750 11.6 0.1356 9.76E-06 489 .3 800 11.9 0.1367 9.82E-06 489 .3

c. Material including copper and nickel content: Material properties assigned to clad flaws are that of the underlying material be it base metal or weld. These properties are identified in Table B-3. This is consistent with the NRC PTS Risk Study [8, 9].

d. Material property uncertainties:

1) Bead width: 1 inch - bead widths vary for all plants. Based on the NRC PTS Risk Study [8, 9], a nominal dimension of 1 inch is selected for all analyses because this parameter is not expected to influence significantly the predicted reactor vessel failure probabilities.

2) Truncation limit: Cladding thickness rounded to the next 1/100th of the total reactor vessel thickness to be consistent with the NRC PTS Risk Study [8, 9].

3) Surface flaw depth: 0.161 inch WCAP- 16168-NP-A June 2008 Revision 2

B-5

4) All cladding flaws are surface-breaking. Only flaws in cladding that would influence brittle fracture of the reactor vessel are brittle. This is consistent with the NRC PTS Risk Study [8, 9].

e. Additional cladding properties are identified in Table B-4.

7. Base metal:

a. Wall thickness: 7.875 inches

b. Material properties are identified in Table B-2 and B-3. This is consistent with the NRC PTS Risk Study [8, 9]:

Table B-2 Base Metal Material Properties Specific Young's Thermal Thermal Heat Modulus of Expansion Conductivity (Btu/LBM- Elasticity Coefficient Temperature (Btu/hr-ft-°F) °F) (KSI) (OF-I) (LBM/fDensity Poisson's (TF) "K".. "" "" BMC..a*Ci " 0 - - - 489 .3 70 24.8 0.1052 29200 - 489 .3 100 25 0.1072 - 7.06E-06 489 .3 150 25.1 0.1101 - 7.16E-06 489 .3 200 25.2 0.1135 28500 7.25E-06 489 .3 250 25.2 0.1166 - 7.34E-06 489 .3 300 25.1 0.1194 28000 7.43E-06 489 .3 350 25 0.1223 - 7.5E-06 489 .3 400 25.1 0.1267 27400 7.58E-06 489 .3 450 24.6 0.1277 - 7.63E-06 489 .3 500 24.3 0.1304 27000 7.7E-06 489 .3 550 24 0.1326 - 7.77E-06 489 .3 600 23.7 0.135 26400 7.83E-06 489 .3 650 23.4 0.1375 - 7.9E-06 489 .3 700 23 0.1404 25300 7.94E-06 489 .3 750 22.6 0.1435 - 8.OE-06 489 .3 800 22.2 0.1474 23900 8.05E-06 489 .3 WCAP-16168-NP-A June 2008 Revision 2

B-6 Table B-3 BV1-Specific Material Values Drawn from the RVID (see Ref. 44, Table 4.1) Un-Major Material Region Description Cu Ni CU P Mn Irradiated RND Type Heat Location [wt%J Iwt%] Iwt%] [wt%] RTNDT 1 Axial Weld 305414A Lower 0.337 0.609 0.012 1.440 - 56 2 Axial Weld 305414B Lower 0.337 0.609 0.012 1.440 -56 3 Axial Weld 305424A Upper 0.273 0.629 0.013 1.440 - 56 4 Axial Weld 305424B Upper 0.273 0.629 0.0 13 1.440 -56 5 Circ Weld 90136 Intermediate 0.269 0.070 0.013 0.964 - 56 6 Plate C6317-1 Lower 0.200 0.540 0.010 1.310 27 7 Plate C6293-2 Lower 0.140 0.570 0.015 1.300 20 8 Plate C4381-2 Upper 0.140 0.620 0.015 1.400 73 9 Plate C4381-1 Upper 0.140 0.620 0.015 1.400 43

8. Weld metal details: Details of information used in addressing weld-specific information are taken directly from the NRC PTS Risk Study [44], Table 4.2. Summaries are reproduced as Table B-4.

Values for SAW Weld Volume fraction and Repair Weld Volume fraction in Table B-4 were changed to 96.7% and 2.3% respectively per NUREG-1874 [9]. WCAP- 16168-NP-A June 2008 Revision 2

B-7 Table B-4 Summary of Reactor Vessel-Specific Inputs for Flaw Distribution Inner Radius (to cladding) i[!n] 85.5 [ 78.5* 86 1 86 1Vessel specific info Metal Thickness [in] 8438 7.875 8.5 8675 Vessel specific info Total Wall Thickness [in] ....... 801 8.62 875 8.8 eslsecific info B-Volure fraction ((%I 97% o00%- SMAW%- REPAIR% Thru-Wall Bead. All plants report plant:specific [in] 0.1875 0.1875 1 0.1875 0.1875 dimensions of 3116-in. Thickness Judgment. Approx. 2X the size of the largest non-repair Truncation Limit I flaw observed in PVRUF & _t Shoreham. Buried or Surface All flaws are buried Observation Observation: Virtually all of the weld flaws in PVRUF & Circ flaws in Circ welds, axial flaws in axial Shoreham werealigned with Orientation welds. the welding direction because they were tack.of sidewall SAW 4 fusibn defects. Density basis 4- a --.--- ,,---.--.--.--------. Weld Shoreham density Highest of observations Statistically similar distributions from Shoreham and PVRUF were combined to provide more robust Aspect ratio. Shoreham PVRUF observations estimates, when based on basis, judgment the amount data were limited andlor insufficient to identify different trends for aspect ratios for flaws in the twoyvessels. Statistically similar distributions combined to Depth basis Shoreham & PVRUF observations provide more robust estimates WCAP- 16168-NP-A June 2008 Revision 2

B-8 1 Table B-4 Summary of Reactor Vessel-Specific Inputs for Flaw Distribution (cont.) upper Douna to au piant specific info provided by Volume fraction [%] 1% Steve Byrne (Westinghouse - Windsor).

                                     -4                                                               1-Oconee is generic value based on average of all plants specific vatues Thru-Wall Bead fin]    021          0,20          0122            0.25         .(including Shoreham &

Thickness PVRUF data). Other values are plant specific as' reported by Steve Bvrne. Judgment. Approx. 2X the size of the:largestnon-repair Truncation Limit [in] flaw observed in PVRUF & Shoreham, Buried or Surface All flaws are buried Observation Observation: Virtually all of SMAW the weld flaws in PVRUF & Weld Orientation Circ flaws in circ welds, axial flaws in axial Shoreham were aligned with Ot welds. the welding direction because they were lack of sidewall fusion defects. Density basis .. Shoreham density Highest of observations

                                                                                                         -Statistically similar distributionsfrom Shoreham and PVRUF were combined to provide more robust Aspect ratio                                                                           estimates, when based on basis                              h     a...      .      bsra.n
                                                                                  ..                      judgment the 'amount data were limited and/or insufficient to identify different trends for aspect.ratios for flaws in the two vessels.

Statistically similar distributions combined to Depth basis Shoreham &,PVRUF observations provide more robust estimates A-- ~----.--- A A A -_______________ a - - -- A WCAP- 16168-NP-A June 2008 Revision 2

B-9 Table B-4 Summary of Reactor Vessel-Specific Inputs for Flaw Distribution (cont.) juogmen[. A rounoea integral percentage that Repair Volume fraction 2% exceeds the repaired volume Weld observed for Shoreham and for PVRUF, whichwas 1 5%. Thru-Wall Bead Generic value: As observed Thick Beas [in] 0,14 in PVRUF and Shoreham by Thickness PNNL. Judgment. Approx. 2Xthe largest repair flaw found in, Truncation Limit fin] 2 PVRUF & Shoreham. Also based on maximum expected width of repair cavity. Buried or Surface I- All flaws are buried Observation All flaws are buried Observation The repair flaws had complex shapes and orientations that were not aligned.with either the axial or circumferential welds; for consistency With Orientation Circ flaws in circ welds, axial flaws in axial the available treatments of welds. flaws bythe FAVOR code, a common treatmentof! orientations was adopted for flaws in SAW/SMAW and. reoair welds. i I Density basis, Shoreham density Highest of observations Statistically similar distributions from ShOreham and .PVRUF were combined to provide more robust AsPectrati - Shoreham.& PVRUF observations estimates, when based on basis judgment the amount data were limited and/or insufficient to identify different trends for aspect ratios for flaws in the two.vessels. Statistically similar distributions combined to Depth basis Shoreham & PVRUF observations provide mord robust estimates WCAP-16168-NP-A June 2008 Revision 2

B-10 Table B-4 Summary of Reactor Vessel-Specific Inputs for Flaw Distribution (cont.) Actual Thickness I [in] 0.188 1 0.156 1 0.25 1 0.313 Vessel specific info I I

         # of Layers        [#]              1,             2             2          2 Vessel specific info Bead widths of 1to 5-in.

characteristic of machine deposited cladding. Bead widths down to 1/22-in. can occur over welds. Nominal dimension of 1-in. selected Bead Width [in] 1 for all analyses because this parameter .is not expected to influence significantly the predicted vessel failure probabilities. May need to refine this estimate later, particularly for Oconee who Truncation Limit [in) IActual clad thickness rounded to the nearest 11100d of the total vessel wall thickness reported a 5-in bead width. Judgment & computational Surface flaw [ convenience depth in FAVOR .. 0.263 0360 Judgment. Only flaws in cladding that would influence brittle fracture of the vessel Buried or Surface All flaws are surface breaking are brittle. Material properties assigned to clad flaws are that of. the -underlying material, be it base or weld. Observation: All flaws observed in PVRUF & Shoreham were lack of0inter-Orientation All circumferential. run fusion defects, and cladding is~always deposited circumferentially No surface flaws observed. Density is 111000 that of the observed buried flaws in Density basis cladding of vessels examined bY .PNNL. If Judgment there is more than one clad layer then there are no cladflaws. Aspect bas. ratio ais Observations on buried flaws Judgment Depth of all surface flaws is the actual clad Depth basis thickness rounded up to the nearest i/i001" Judgment. of the total vessel wall thickness. June2008 WCAP- 16168-NP-A WCAP-16168-NP-A June 2008 Revision 2

B-11 Table B-4 Summary of Reactor Vessel-Specific Inputs for Flaw Distribution (cont.) lt;F Imli, WI.twl01iut; u5j1i Truncatiorn L:imit [in] 0433 of the largesl flaw observed in all PNNL ate insp*ctions. Buried or Surface All flaws are buried Observation 4-A. Observalion & Physics: No Half of the:simulated flaws are observed orientation Orientation preference, and no reason to circumferential, half are axial. Plate suspect one (other than laminations which are beniqn. Density basis - 1/10 of small weld flaw density, 1/40 of large Judgment. Supported by weld flaw density of the PVRUF data limited data. Aspect ratio Same as for PVRUF welds Judgment basis___- Depth basis Same as for PVRUF welds Judgment. Supported by limited data.

9. TWCF95-TOTAL value calculated at 60 EFPY using correlations from NUREG-1874 (Reference 9):

1.76E-08 Events per year WCAP- 16168-NP-A June 2008 Revision 2

C-1 APPENDIX C BEAVER VALLEY UNIT 1 PROBSBFD OUTPUT June 2008 WCAP- 161 68-NP-A WCAP-16168-NP-A June 2008 Revision 2

C-2 C-i: 10 Year ISI Only STRUCTURAL RELIABILITY AND RISK ASSESSMENT (SRRA) WESTINGHOUSE MONTE-CARLO SIMULATION PROGRAM PROBSBFD VERSION 1.0 INPUT VARIABLES FOR CASE 3: BV1 HUCD 10 YR ISI ONLY NCYCLE = 80 NFAILS = 1001 NTRIAL = 1000 NOVARS = 19 NUMSET = 2 NUMISI = 5 NUMSSC = 4 NUMTRC = 4 NUMFMD = 4 VARIABLE DISTRIBUTION MEDIAN DEVIATION SHIFT USAGE NO. NAME TYPE LOG VALUE OR FACTOR MV/SD NO. SUB 1 FIFDepth - CONSTANT - 2.OOOOD-02 1 SET 2 IFlawDen - CONSTANT - 3.6589D-03 2 SET 3 ICy-ISI - CONSTANT - 1. OOOOD+01 1 ISI 4 DCy-ISI - CONSTANT - 8. OOOOD+01 2 ISI 5 MV-Depth - CONSTANT - 1.5000D-02 3 ISI 6 SD-Depth - CONSTANT - 1.8500D-01 4 ISI 7 CEff-ISI - CONSTANT - 1.OOOOD+00 5 ISI 8 Aspectl - CONSTANT - 2.OOOOD+00 1 SSC 9 Aspect2 - CONSTANT - 6.OOOOD+00 2 SSC 10 Aspect3 - CONSTANT - 1. OOOOD+01 3 SSC 11 Aspect4 - CONSTANT - 9 .9000D+01 4 SSC 12 NoTr/Cy - CONSTANT - 7.OOOOD+00 1 TRC 13 FCGTh1d - CONSTANT - 1.5000D+00 2 TRC 14 FCGR-UC NORMAL NO 0.OOOOD+00 1.OOOOD+00 .00 3 TRC 15 DKINFile - CONSTANT - 1.OOOOD+00 4 TRC 16 Percentl - CONSTANT - 5.6175D+01 1 FMD 17 Percent2 - CONSTANT - 3. 0283D+01 2 FMD 18 Percent3 - CONSTANT - 3.9086D+00 3 FMD 19 Percent4 - CONSTANT - 9.6333D+00 4 FMD INFORMATION GENERATED FROM FAVLOADS.DAT FILE AND SAVED IN DKINSAVE.DAT FILE: WALL THICKNESS = 8.0360 INCH FLAW DEPTH MINIMUM K AND MAXIMUM K FOR TYPE 1 WITH AN ASPECT RATIO OF 2. 8.0360OD-02 2 .41927D+00 1.03655D+01 1.47862D-01 3. 22858D+00 1.40170D+01 4.01800D-01 1. 29279D+01 1.75751D+01 6.02700D-01 1. 41327D+01 2.09080D+01 8.03600D-01 1. 49423D+01 2.33544D+01 1.60720D+00 1. 45812D+01 2.72710D+01 2.41080D+00 1. 02448D+01 2.63600D+01 4.01800D+00 2. 35823D+00 2.78623D+01 WCAP- 16168-NP-A June 2008 Revision 2

C-3 C-1: 10 Year ISI Only (cont.) TYPE 2 WITH AN ASPECT RATIO OF 6.

8. 03600D-02 3.63673D+00 1. 56338D+01 1 .47862D-01 4.95557D+00 2. 15454D+01 4.0180OD-01 1.90999D+01 2 .63794D+01

6. 02700D-01 2.31650D+01 3 .16223D+01

8. 03600D-01 2.48064D+01 3. 60464D+01

1. 60720D+00 2.65025D+01 4. 51155D+01

2. 41080D+00 2.31198D+01 4. 76172D+01

4. 01800D+00 1.54934D+01 5 .27667D+01 TYPE 3 WITH AN ASPECT RATIO OF 10.

8.03600D-02 3. 98451D+00 1. 71374D+01 1.47862D-01 5. 29827D+00 2.30393D+01 4.01800D-01 2. 02922D+01 2. 81955D+01 6.02700D-01 2 .51750D+01 3. 36684D+01 8.03600D-01 2.69393D+01 3. 84779D+01 1.60720D+00 2. 92755D+01 4 .91684D+01 2.41080D+00 2. 74642D+01 5.45509D+01 4.01800D+00 2. 02195D+01 6.28814D+01 TYPE 4 WITH AN ASPECT RATIO OF 99. 8.03600D-02 6.51796D+00 1.75511D+01 1.60720D-01 1.01756D+01 2. 28059D+01 2.41080D-01 1.54398D+01 2 .23553D+01 4.01800D-01 2' 18696D+01 2 .94323D+01 6.02700D-01 2.69582D+01 3 .66108D+01 8.03600D-01 2.88204D+01 4. 17713D+01 1.60720D+00 3.37365D+01 5 .67413D+01 2.41080D+00 3.35927D+01 6.64759D+01 AVERAGE CALCULATED VALUES FOR: Surface Flaw Density with FCG and ISI NUMBER FAILED = 0 NUMBER OF TRIALS = 1000 DEPTH (WALL/400) AND FLAW DENSITY FOR ASPECT RATIOS OF 2, 6, 10 AND 99 8 4.4254D-04 1. 4320D-04 1.4728D-05 4.7035D-05 9 0.OOOOD+00 8. 8686D-05 1.4703D-05 2.7347D-05 10 0.0000D+00 4 .4175D-06 9.2631D-07 7.2598D-07 11 0.0000D+00 2 .2821D-07 5. 9150D-08 7. 0131D-08 12 0.OOOOD+00 2 .2665D-07 2.9099D-08 0.OOOOD+00 13 0.OOOOD+00 0 OOOOD+00 2. 8861D-08 0.OOOOD+00 WCAP- 16168-NP-A *. June 2008 Revision 2

C-4 C-2: ISI Every 10 Years STRUCTURAL RELIABILITY AND RISK ASSESSMENT (SRRA) WESTINGHOUSE MONTE-CARLO SIMULATION PROGRAM PROBSBFD VERS ION 1.0 INPUT VARIABLES FOR CASE 2: BV1 HUCD 10 YR ISI INT NCYCLE 80 NFAILS = 1001 NTRIAL = 1000 NOVARS 19 NUMSET = 2 NUMISI = 5 NUMSSC 4 NUMTRC = 4 NUMFMD = 4 VARIABLE DISTRIBUTION MEDIAN DEVIATION SHIFT USAGE NO. NAME TYPE LOG VALUE OR FACTOR MV/SD NO. SUB 1 FIFDepth - CONSTANT - 2.OOOOD-02 1 SET 2 IFlawDen - CONSTANT - 3.6589D-03 2 SET 3 ICy-ISI - CONSTANT - 1. OOOOD+01 1 ISI 4 DCy-ISI - CONSTANT - 1. OOOOD+01 2 ISI 5 MV-Depth - CONSTANT - 1.500OD-02 3 ISI 6 SD-Depth - CONSTANT - 1.8500D-01 4 ISI 7 CEff-ISI - CONSTANT - 1.OOOOD+00 5 ISI 8 Aspectl - CONSTANT - 2 .OOOOD+00 1 SSC 9 Aspect2 - CONSTANT - 6 .OOOOD+00 2 SSC 10 Aspect3 - CONSTANT - 1. OOOOD+01 3 SSC 11 Aspect4 - CONSTANT - 9 9000D+01 4 SSC 12 NoTr/Cy - CONSTANT - 7. OOOOD+00 1 TRC 13 FCGThld - CONSTANT - 1. 5000D+00 2 TRC 14 FCGR-UC NORMAL NO 0 .OOOOD+00 1.OOOOD+00 .00 3 TRC 15 DKINFile - CONSTANT - 1 .OOOOD+00 4 TRC 16 Percentl - CONSTANT - 5. 6175D+01 1 FMD 17 Percent2 - CONSTANT - 3. 0283D+01 2 FMD 18 Percent3 - CONSTANT - 3.9086D+00 3 FMD 19 Percent4 - CONSTANT - 9.6333D+00 4 FMD INFORMATION GENERATED FROM FAVLOADS.DAT FILE AND SAVED IN DKINSAVE.DAT FILE: WALL THICKNESS = 8.0360 INCH FLAW DEPTH MINIMUM K AND MAXIMUM K FOR TYPE 1 WITH AN ASPECT RATIO OF 2. 8.03600D-02 2.41927D+00 1.03655D+01 1.47862D-01 3.22858D+00 1.40170D+01 4.01800D-01 1.29279D+01 1.75751D+01 6.02700D-01 1.41327D+01 2.09080D+01 8.03600D-01 1.49423D+01 2.33544D+01 1.60720D+00 1.45812D+01 2.72710D+01 2.41080D+00 1.02448D+01 2.63600D+01 4.01800D+00 2.35823D+00 2.78623D+01 WCAP- 16168-NP-A June 2008 Revision 2

C-5 C-2: ISI Every 10 Years (cont.) TYPE 2 WITH AN ASPECT RATIO OF 6.

8. 03600D-02 3.63673D+00 1.56338D+01 1.47862D-01 4.95557D+00 2. 15454D+01 4.0180OD-01 1.90999D+01 2. 63794D+01

6. 02700D-01 2.31650D+01 3 .16223D+01

8. 03600D-01 2.48064D+01 3. 60464D+01 1.60720D+00 2.65025D+01 4. 51155D+01 2.41080D+00 2.31198D+01 4. 76172D+01

4. 01800D+00 1.54934D+01 5.27667D+01 TYPE 3 WITH AN ASPECT RATIO OF 10.

8.03600D-02 3.98451D+00 1.71374D+01 1.47862D-01 5.29827D+00 2.30393D+01 4.01800D-01 2.02922D+01 2.81955D+01 6.02700D-01 2.51750D+01 3.36684D+01 8.03600D-01 2.69393D+01 3.84779D+01 1.60720D+00 2.92755D+01 4.91684D+01 2.41080D+00 2.74642D+01 5.45509D+01 4.01800D+00 2.02195D+01 6.28814D+01 TYPE 4 WITH AN ASPECT RATIO OF 99. 8.03600D-02 6. 51796D+00 1.75511D+01 1.60720D-01 1. 01756D+01 2.28059D+01 2.41080D-01 1. 54398D+01 2.23553D+01 4.01800D-01 2. 18696D+01 2.94323D+01 61.02700D-01 2. 69582D+01 3.66108D+01 8.03600D-01 2. 88204D+01 4.17713D+01 1.60720D+00 3. 37365D+01 5.67413D+01 2.41080D+00 3. 35927D+01 6.64759D+01 AVERAGE CALCULATED VALUES FOR: Surface Flaw Density with FCG and ISI NUMBER FAILED = 0 NUMBER OF TRIALS = 1000 DEPTH (WALL/400) AND FLAW DENSITY FOR ASPECT RATIOS OF 2, 6, 10 AND 99 8 4.3486D-08 1.2355D-08 1.2447D-09 4. 0471D-09 9 0.OOOOD+00 6.1902D-09 9. 9626D-10 1.8380D-09 10 0.OOOOD+00 1. 8825D-10 3.7663D-11 2. 6218D-11 11 0.OOOOD+00 4.7355D-12 1. 6752D-12 1.3302D-12 12 0.OOOOD+00 3.5199D-12 4.3837D-13 0.OOOOD+00 13 0.OOOOD+00 0.OOOOD+00 3. 0423D-13 0. OOOD+00 WCAP- 16168-NP-A June 2008 Revision 2

D-1 APPENDIX D BEAVER VALLEY UNIT 1 PTS TRANSIENTS Table D-1 PTS Transient Descriptions for BV1 Count TH System Failure Operator Action Mean 1E HZP Dominant* Case Frequency 1 002 3.59 cm [1.414 in] surge line None. 1.23E-04 No No break 22 __ 003 003 5.08 cm [2 in] surge line break None. Noe_97E_5No 9.76E-05 No No N 3 007 None. 2.11E-05 No Yes at 32, 60, 100, 2.54 cm [8 in] surge line break 200 EFPY 4 009 None. 6.99E-06 No Yes at 32, 60, 100, 2.54 cm [16 in] hot leg break 200 EFPY 5 014 Reactor/turbine trip w/one stuck None. 2.22E-04 No No open pressurizer SRV 6 031 Reactor/turbine trip w/feed and None. 3.1 OE-07 No No bleed (Operator open all pressurizer PORVs and use all charging/HHSI pumps) 7 034 Reactor/turbine trip w/two stuck None. 4.95E-07 No No open pressurizer SRV's 8 056 None. 1.23E-04 Yes Yes at 32, 10.16 cm [4.0 in] surge line break 60, 100, 200 EFPY 9 059 Reactor/turbine trip w/one stuck None. 3.46E-04 No No open pressurizer SRV which recloses at 3,000 s. 10 060 Reactor/turbine trip w/one stuck None. 2.15E-05 No Yes at 32, open pressurizer SRV which 60, 100 recloses at 6,000 s. EFPY 11 061 Reactor/turbine trip w/two stuck None. 1.79E-06 No No open pressurizer SRV which recloses at 3,000 s. 12 062 Reactor/turbine trip w/two stuck None. 1.08E-07 No No open pressurizer SRV which recloses at 6,000 s. 13 064 Reactor/turbine trip w/two stuck None. 8.67E-08 Yes No open pressurizer SRV's 14 065 Reactor/turbine trip w/two stuck Operator opens all ASDVs 1.04E-09 No No open pressurizer SRV's and HHSI 5 minutes after HHSI failure would have come on. WCAP- 16168-NP-A June 2008 Revision 2

D-2 Table D-1 PTS Transient Descriptions for BVI Count TH System Failure Operator Action Mean IE HZP Dominant* Case Frequency 15 066 Reactor/turbine trip w/two stuck None. 1.18E-06 No No open pressurizer SRV's. One valve recloses at 3000 seconds while the other valve remains open. 16 067 Reactor/turbine trip w/two stuck None. 1.18E-06 No No open pressurizer SRV's. One valve recloses at 6000 seconds while the other valve remains open. 17 068 Reactor/turbine trip w/two stuck Operator opens all ASDVs 1.33E-08 No No open pressurizer SRV's that 5 minutes after HHSI reclose at 6000 s with HHSI would have come on. failure. 18 069 Reactor/turbine trip w/two stuck None. 2.09E-08 Yes No open pressurizer SRVs which reclose at 3,000 s. 19 070 Reactor/turbine trip w/two stuck None. 2.09E-08 Yes No open pressurizer SRVs which reclose at 6,000 s. 20 071 Reactor/turbine trip w/one stuck None. 3.74E-06 Yes Yes at 32 open pressurizer SRV which EFPY recloses at 6,000 s. 21 072 Reactor/turbine trip w/one stuck Operator opens all ASDVs 5.14E-07 No No open pressurizer SRV with HHSI 5 minutes after HHSI failure. would have come on. 22 073 Reactor/turbine trip w/one stuck Operator open all ASDVs 5 6.56E-08 Yes No open pressurizer SRV with HHSI minutes after HHSI would failure have come on. 23 074 Main steam line break with AFW None. 1:46E-06 No No continuing to feed affected generator 24 076 Reactor/turbine trip w/full MFW Operator trips reactor 1.06E-04 Yes No to all 3 SGs (MFW maintains SG coolant pumps. level near top). 25 078 Reactor/turbine trip with failure Operator opens all ASDVs 3.25E-08 No No of MFW and AFW. to let condensate fill SGs. 26 081 Main Steam Line Break with Operator opens ADVs (on 2.65E-06 No No AFW continuing to feed affected intact generators). HHSI is generator and with HHSI failure restored after CFTs initially, discharge 50%. 27 082 Reactor/turbine trip w/one stuck Operator opens all ASDVs 1.51E-06 No No open pressurizer SRV (recloses at 5 minutes after HHSI 6000 s) and with HHSI failure. would have started. WCAP- 16168-NP-A June 2008 Revision 2

D-3 Table D-1 PTS Transient Descriptions for BVI Count TH System Failure Operator Action Mean IE HZP Dominant* Case Frequency 28 083 2.54 cm [1.0 in] surge line break Operator trips RCPs. 3.5 1E-06 No No with HHSI failure and motor Operator opens all ASDVs driven AFW failure. MFW is 5 minutes after HHSI tripped. Level control failure would have come on. causes all steam generators to be overfed with turbine AFW, with the level maintained at top of SGs. 29 092 Reactor/turbine trip w/two stuck None. 2.13E-07 Yes No open pressurizer SRV's, one recloses at 3000 s. 30 093 Reactor/turbine trip w/two stuck None. 2.13E-07 Yes No open pressurizer SRV's. One valve recloses at 6000 seconds while the other valve remains open. 31 094 Reactor/turbine trip w/one stuck None. 4.10E-05 Yes No open pressurizer SRV. 32 097 Reactor/turbine trip w/one stuck None. 3.74E-06 Yes Yes at 32, open pressurizer SRV which 60 EFPY recloses at 3,000 s. 33 102 Operator controls HHSI 30 1.02E-04 No Yes at 100, minutes after allowed. 200 EFPY Break is assumed to occur inside containment so that Main steam line break with AFW the operator trips the RCPs continuing to feed affected due to adverse containment generator for 30 minutes. conditions. 34 103 Operator controls HHSI 30 1.07E-05 Yes Yes at 60, minutes after allowed. 100, 200 Break is assumed to occur EFPY inside containment so that Main steam line break with AFW the operator trips the RCPs continuing to feed affected due to adverse containment generator for 30 minutes. conditions. 35 104 Operator controls HHSI 60 1.09E-04 No Yes at 100, minutes after allowed. 200 EFPY Break is assumed to occur inside containment so that Main steam line break with AFW the operator trips the RCPs continuing to feed affected due to adverse containment generator for 30 minutes. conditions.

       ýWCAP- 16168-NP-A                                                                       June 2008 Revision 2

D-4 Table D-1 PTS Transient Descriptions for BV1 Count TH System Failure Operator Action Mean IE HZP Dominant* Case Frequency 36 105 Operator controls HHSI 60 1.07E-05 Yes No minutes after allowed. Break is assumed to occur inside containment so that Main steam line break with AFW the operator trips the RCPs continuing to feed affected due to adverse containment generator for 30 minutes. conditions. 37 106 Operator controls HHSI 30 2.21E-05 No No minutes after allowed. Break is assumed to occur inside containment so that Main steam line break with AFW the operator trips the RCPs continuing to feed affected due to adverse containment generator. conditions. 38 107 Operator controls HHSI 30 4.31E-07 Yes No minutes after allowed. Break is assumed to occur inside containment so that Main steam line break with AFW the operator trips the RCPs continuing to feed affected due to adverse containment generator. conditions. 39 108 Small steam line break Operator controls HHSI 30 6.46E-04 Yes No (simulated by sticking open all minutes after allowed. SG-A SRVs) with AFW continuing to feed affected generator for 30 minutes. 40 109 Operator controls HHSI 30 6.81E-05 Yes No minutes after allowed. Small steam line break Break is assumed to occur (simulated by sticking open all inside containment so that SG-A SRVs) with AFW the operator trips the RCPs continuing to feed affected due to adverse containment generator for 30 minutes. conditions. 41 110 Small steam line break Operator controls HHSI 60 6.91E-04 No Yes at 200 (simulated by sticking open all minutes after allowed. EFPY SG-A SRVs) with AFW continuing to feed affected generator for 30 minutes 42 111 Operator controls HHSI 60 6.82E-05 Yes No minutes after allowed. Small steam line break Break is assumed to occur (simulated by sticking open all inside containment so that SG-A SRVs) with AFW- the operator trips the RCPs continuing to feed affected due to adverse containment generator for 30 minutes. conditions. WCAP- 16168-NP-A June 2008 Revision 2

D-5 Table D-1 PTS Transient Descriptions for BV1 Count TH System Failure Operator Action Mean IE HZP Dominant* Case Frequency 43 112 Operator controls HHSI 30 1.41E-05 No No minutes after allowed. Small steam line break Break is assumed to occur (simulated by sticking open all inside containment so that SG-A SRVs) with AFW the operator trips the RCPs continuing to feed affected due to adverse containment generator. conditions. 44 113 Operator controls HHSI 30 2.74E-06 Yes No minutes after allowed. Small steam line break Break is assumed to occur (simulated by sticking open all inside containment so that SG-A SRVs) with AFW the operator trips the RCPs continuing to feed affected due to adverse containment generator. conditions. 45 114 7.18 cm [2.828 in] surge line None. 9.76E-05 No No break, summer conditions (HHSI, LHSI temp = 557F, Accumulator Temp = 1057F), heat transfer coefficient increased 30% (modeled by increasing heat transfer surface area by 30% in passive heat structures). 46 115 None.. 9.76E-05 No No 46_____ 115____ 7.18 cm [2.828 in] cold leg break 47 116 14.366 cm [5.657 in] cold leg None. 1.81E-05 No No break with break area increased 30% 48 117 14.366 cm [5.657 in] cold leg None. 2.11E-05 No No break, summer conditions (HHSI, LHSI temp = 55°F, Accumulator Temp = 105°7) 49 118 Small steam line break None. 9.30E-06 No No (simulated by sticking open all SG-A SRVs) with AFW continuing to feed affected generator 50 119 Reactor/turbine trip w/two stuck Operator controls HHSI (1 6.84E-07 No No open pressurizer SRV which minute delay). Updated recloses at 6,000 s control logic. 51 120 Reactor/turbine trip w/two stuck Operator controls HHSI (10 9.98E-07 No No open pressurizer SRV which minute delay). Updated recloses at 6,000 s control logic. 52 121 Reactor/turbine trip w/two stuck Operator controls HHSI (1 1.33E-07 Yes No open pressurizer SRV which minute delay). Updated recloses at 3,000 s control logic. WCAP- 16168-NP-A June 2008 Revision 2

D-6 Table D-1 PTS Transient Descriptions for BVI Count TH System Failure Operator Action Mean IE HZP Dominant* Case Frequency 53 122 Reactor/turbine trip w/two stuck Operator controls HHSI (1 1.33E-07 Yes No open pressurizer SRVs which minute delay). Updated reclose at 6,000 s control logic. 54 123 Reactor/turbine trip w/two stuck Operator controls HHSI (10 1.65E-07 Yes Yes at 32 open pressurizer SRVs which minute delay). Updated EFPY reclose at 3,000 s control logic. 55 124 Reactor/turbine trip w/two stuck Operator controls HHSI (10 1.65E-07 Yes No open pressurizer SRVs which minute delay). Updated reclose at 6,000 s control logic. 56 125 Reactor/turbine trip w/one stuck Operator controls HHSI (1 1.34E-04 No No open pressurizer SRV which minute delay). Updated recloses at 6,000 s control logic. 57 126 Reactor/turbine trip w/one stuck Operator controls HHSI (10 1.87E-04 No Yes at 32, open pressurizer SRV which minute delay). Updated 60, 100 recloses at 6,000 s control logic. EFPY 58 127 Reactor/turbine trip w/one stuck Operator controls HHSI (1 2.59E-05 Yes No open pressurizer SRV which minute delay). Updated recloses at 6,000 s control logic. 59 128 Reactor/turbine trip w/one stuck Operator controls HHSI (1 2.59E-05 Yes No open pressurizer SRV which minute delay). Updated recloses at 3,000 s control logic. 60 129 Reactor/turbine trip w/one stuck Operator controls HHSI (10 3.09E-05 Yes Yes at 32, open pressurizer SRV which minute delay). Updated 60 EFPY recloses at 6,000 s control logic. 61 130 Reactor/turbine trip w/one stuck Operator controls HHSI (10 3.09E-05 Yes Yes at 32, open pressurizer SRV which minute delay). Updated 60, 100 recloses at 3,000 s control logic. EFPY Notes:

1. TH - Thermal hydraulics

2. LOCA - Loss-of-coolant accident

3. SBLOCA - Small-break loss-of-coolant accident

4. MBLOCA - Medium-break loss-of-coolant accident

5. LBLOCA - Large-break loss-of-coolant accident

6. HZP - Hot-zero power

7. SRV - Safety and relief valve

8. MSLB - Main steam line break

9. AFW - Auxiliary feedwater

10. HPI - High-pressure injection

11. RCPs - Reactor coolant pumps

• The arbitrary definition of a dominant transient is a transient that contributes 1% or more of the total Through-Wall Cracking Failure (TWCF).

WCAP- 16168-NP-A June 2008 Revision 2

E-1 APPENDIX E BEAVER VALLEY UNIT 1 FAVPOST OUTPUT June 2008 WCAP- 16168-NP-A WCAP-16168-NP-A June 2008 Revision 2

E-2 E-1: 10 Year ISI Only

• WELCOME TO FAVOR
• FRACTURE ANALYSIS OF VESSELS: OAK RIDGE *
• VERSION 06.1
• FAVPOST MODULE: POSTPROCESSOR MODULE
• COMBINES TRANSIENT INITIAITING FREQUENCIES
• WITH RESULTS OF PFM ANALYSIS
• PROBLEMS OR QUESTIONS REGARDING FAVOR
• SHOULD BE DIRECTED TO *
• TERRY DICKSON *
• OAK RIDGE NATIONAL LABORATORY
• e-mail: dicksontl@ornl.gov
• This computer program was prepared as an account of *
• work sponsored by the United States Government *
• Neither the United States, nor the United States *
• Department of Energy, nor the United States Nuclear *
• Regulatory Commission, nor any of their employees, *
• nor any of their contractors, subcontractors, or their *
• employees, makes any warranty, expressed or implied, or *
• assumes any legal liability or responsibility for the *
• accuracy, completeness, or usefulness of any *
• information, apparatus, product, or process disclosed, *
• or represents that its use would not infringe *
• privately-owned rights.
• DATE: 03-May-2007 TIME: 16:02:21 Begin echo of FAVPost input data deck 16:02:21 03-May-2007 End echo of FAVPost input data deck 16:02:21 03-May-2007 FAVPOST INPUT FILE NAME = postbv.in FAVPFM OUTPUT FILE CONTAINING PFMI ARRAY = INITIATE.DAT FAVPFM OUTPUT FILE CONTAINING PFMF ARRAY = FAILURE.DAT FAVPOST OUTPUT FILE NAME = 70000.out WCAP- 16168-NP-A June 2008 Revision 2

E-3 E-1: 10 Year ISI Only (cont.)

• NUMBER OF SIMULATIONS = 70000
• CONDITIONAL PROBABILITY CONDITIONAL PROBABILITY OF INITIATION CPI=P(IjE) OF FAILURE CPF=P(FIE)

TRANSIEN T MEAN 95th % 99th % MEAN 95th % 99th % RATIO NUMBER CPI CPI CPI CPF CPF CPF CPFmn/CPImn

                                 -                       I I-2    0.OOOOE+00         0.OOOOE+00    0.OOOOE+00       0.OOOOE+00      0.OOOOE+00    0.000OE+00    0.0000 3     3.4447E-07         0 .OOOOE+00   0.OOOOE+00       1.0487E-08      0.OOOOE+00    0.OOOOE+00    0.0304 7    2.4538E-03         5. 7840E-03   2.9648E-02       8. 9261E-06     1.7129E-04    1.6542E-04    0.0036 9    3. 5917E-03        8. 8320E-03   4.7025E-02       9. 1001E-06     1.5080E-04    1.8094E-04    0.0025 14     2.8062E-09         0 .OOOOE+00   0.OOOOE+00       3. 9534E-11     0.OOOOE+00    0.OOOOE+00    0.0141 31     3. 1040E-06        0 OOOOE+00    9.1224E-07       6.5429E-09      0 OOOOE+00    0.OOOOE+00    0.0021 34     3.6725E-06         0 OOOOE+00    8.8530E-06       1.3780E-08      0 OOOOE+00    0.OOOOE+00    0.0038 56     3. 6233E-03        8.4587E-03    4. 0163E-02      1.5372E-05      2. 1788E-04   2.9507E-04    0.0042 59     0. 0000E+00        0.OOOOE+00    0.OOOOE+00       0.OOOOE+00      0 OOOOE+00    0.OOOOE+00    0.0000 60     1. 8872E-05        2 . 1105E- 04 1.4508E-04       1.8606E-05      2. 1105E-04   1.4063E-04    0.9859 61     2. 7682E-05        3.6093E-04    3.3554E-04       5.9860E-06      9. 8116E-05   6.4286E-05    0.2162 62     8. 8381E-06        0.OOOOE+00    3.9900E-05       5.3279E-06      0 OOOOE+00    5.4892E-06    0.6028 64     3 .0356E-04        2.4273E-03    4.1264E-03       3.3431E-07      0 OOOOE+00    8. 1349E-08   0.0011 65     1 .2194E-08        0.OOOOE+00    0.OOOOE+00       0.OOOOE+00      0 OOOOE+00    0.OOOOE+00    0.0000 66     2 .6208E-05        3.6093E-04    2. 9981E-04      1 .4196E-07     0 .OOOOE+00   7.0609E-08    0 .0054 67     4 .2327E-06        0.OOOOE+00    4.7032E-05       2. 0694E-07     0 OOOOE+00    1.1577E-06    0.0489 68     4. 0785E-07        0.OOOOE+00    0.OOOOE+00       3. 6585E-07     0 OOOOE+00    0.OOOOE+00    0.8970 69     6.2806E-04         1.6800E-03    9.3269E-03       4. 6732E-04     1. 6799E-03   6.5428E-03    0.7441 2.4275E-03    4.9362E-03       4. 7246E-05     6.4369E-04    6.3369E-04    0.1398 70     3. 3801E-04 71     1.7881E-05         3.7377E-04    1.5558E-04       1. 6288E-05     3.7377E-04    1. 0611E-04   0.9109 72     0.OOOOE+00         0.OOOOE+00    0.OOOOE+00       0 OOOOE+00      0.OOOOE+00    0.OOOOE+00    0.0000 73     4.1554E-06         6.8843E-05    4.5561E-05       8.3562E-08      0.OOOOE+00    3.6136E-07    0.0201 1.1710E-07         0.OOOOE+00    0.OOOOE+00       4.4526E-09      0.OOOOE+00    0.OOOOE+00    0.0380 74 76     0.OOOOE+00         0.OOOOE+00    0.OOOOE+00       0.OOOOE+00      0.OOOOE+00    0.OOOOE+00    0.0000 78     0.OOOOE+00         0.OOOOE+00    0.OOOOE+00       0.OOOOE+00      0.OOOOE+00    0.OOOOE+00    0.0000 81     0.OOOOE+00         0.OOOOE+00    0.000OE+00       0.OOOOE+00      0.OOOOE+00    0.OOOOE+00    0.0000 0.OOOOE+00         0.OOOOE+00    0.OOOOE+00       0.OOOOE+00      0.OOOOE+00    0.OOOOE+00    0.0000 82 WCAP- 16168-NP-A                                                                                                   June 2008 Revision 2

E-4 E-1: 10 Year ISI Only (cont.) 83 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0 .OOOOE+00 0.0000 92 2.3070E-04 1.1915E-03 3. 0316E-03 1.0747E-06 6.0187E-05 8. 1838E-06 0.0047 93 2.3070E-04 1.1915E-03 3. 0316E-03 1.0747E-06 6. 0187E-05 8. 1838E-06 0.0047 94 0.OOOOE+00 0.OOOOE+00 0 .OOOOE+00 0.OOOOE+00 0.OOOOE+00 0 .OOOOE+00 0.0000 97 7.7573E-05 6.7177E-04 1. 2231E-03 7.4960E-05 6. 7177E-04 1.1540E-03 0.9663 102 1.6387E-06 0.OOOOE+00 0 .OOOOE+00 2.8950E-08 0.OOOOE+00 0 OOOOE+00 0.0177 103 2.5650E-05 3 9193E-04 2. 2038E-04 2. 0631E-06 0.OOOOE+00 2. 3905E-06 0.0804 104 1.6387E-06 0 OOOOE+00 0 OOOOE+00 2.8950E-08 0 OOOOE+00 0.OOOOE+00 0. 0177 105 1.8207E-07 0 OOOOE+00 0 .OOOOE+00 9.2460E-09 0 OOOOE+00 0.OOOOE+00 0. 0508 106 1.5553E-06 0 OOOOE+00 0 OOOOE+00 3.0059E-08 0 OOOOE+00 0.OOOOE+00 0. 0193 107 2. 5612E-05 3. 5810E-04 1. 8762E-04 2. 6496E-06 0 OOOOE+00 3. 9040E-06 0.1035 108 5. 8945E-10 0 OOOOE+00 0.OOOOE+00 0 OOOOE+00 0 OOOOE+00 0 OOOOE+00 0.0000 109 5. 1071E-08 0 OOOOE+00 0.OOOOE+00 2. 9873E-09 0 OOOOE+00 0 OOOOE+00 0.0585 110 5. 8945E-10 0 .OOOOE+00 0.OOOOE+00 0 .OOOOE+00 0.OOOOE+00 0 OOOOE+00 0 .0000 1il 5 1071E-08 0 .OOOOE+00 0.OOOOE+00 2. 9873E-09 0.OOOOE+00 0 OOOOE+00 0.0585 112 4. 9435E-10 0.OOOOE+00 0.OOOOE+00 0 .OOOOE+00 0.OOOOE+00 0 OOOOE+00 0 .0000 113 3.8359E-07 0.OOOOE+00 0.OOOOE+00 1.1730E-09 0.OOOOE+00 0 OOOOE+00 0.0031 114 3 .2501E-05 2.9625E-04 5.5873E-04 1.8948E-07 0.OOOOE+00 1. 0536E-06 0.0058 115 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0 OOOOE+00 0.0000 116 3.2756E-05 1.4620E-04 7.1227E-04 2.0959E-07 0.000OE+00 1. 8495E-06 0.0064 117 1.3498E-04 8.6111E-04 1.6687E-03 6. 1235E-07 1. 7625E-05 5. 7374E-06 0.0045 118 1.0922E-08 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0 OOOOE+00 0.OOOOE+00 0.0000 119 5. 2979E-06 0.OOOOE+00 2. 6217E-05 1.2116E-07 0 OOOOE+00 5.0497E-08 0.0229 120 2 .2859E-05 6.2652E-04 1.8086E-04 1. 8474E-05 6. 2025E-04 9. 6291E-05 0.8082 121 1. 9161E-04 1.0949E-03 2.4748E-03 1. 6094E-06 5. 2227E-05 5. 0140E-06 0. 0084 122 1. 9161E-04 1.0949E-03 2.4748E-03 5. 0299E-07 0 .OOOOE+00 1.0741E-07 0. 0026 123 4. 9718E-04 1.5265E-03 7. 3133E-03 3. 7011E-04 1. 5265E-03 4.8876E-03 0.7444 124 2. 1942E-04 1.0951E-03 2 . 8781E-03 3. 2633E-05 1. 0262E-03 2. 0102E-04 0.1487 125 2.2644E-08 0.OOOOE+00 0.OOOOE+00 1. 4703E-12 0 OOOOE+00 0.OOOOE+00 0. 0001 126 3.2134E-06 0.OOOOE+00 1.0696E-05 3. 0296E-06 0.OOOOE+00 9. 4015E-06 0. 9428 127 3.3065E:05 4.2287E-04 4. 9126E-04 7. 0810E-08 0.OOOOE+00 2.4099E-08 0.0021 128 3.3065E-05 4.2287E-04 4.9126E-04 7. 0810E-08 0.OOOOE+00 2.4099E-08 0. 0021 129 3.5962E-05 4.2289E-04 5.6359E-04 4.2114E-06 1.0047E-04 2.8955E-05 0.1171 130 6.4214E-05 4.8542E-04 9.0405E-04 3.7700E-05 4.8057E-04 4.3728E-04 0.5871 NOTES: CPI IS CONDITIONAL PROBABILITY OF CRACK INITIATION, P(IIE) CPF IS CONDITIONAL PROBABILITY OF TWC FAILURE, P(FIE) June 2008 WCAP- 161 68-NP-A WCAP- 16168-NP-A June 2008 Revision 2

E-5 E-1: 10 Year ISI Only (cont.)

• PROBABILITY DISTRIBUTION FUNCTION (HISTOGRAM) *
• FOR THE FREQUENCY OF CRACK INITIATION
• FREQUENCY OF RELATIVE CUMULATIVE CRACK INITIATION DENSITY DISTRIBUTION (PER REACTOR-OPERATING YEAR) (%) (%)

0.OOOOE+00 16.2600 16.2600 1.3009E-06 78.7386 94.9986 3.9026E-06 2.5114 97.5100 6,.5043E-06 1.0343 98.5443 9.1060E-06 0.5171 99.0614 1.1708E-05 0.2743 99.3357 1.4309E-05 0.1857 99.5214 1.6911E-05 0.1057 99.6271 1.9513E-05 0.0786 99.7057 2.2115E-05 0.0529 99.7586 2.4716E-05 0.0357 99.7943 2.7318E-05 0.0286 99.8229 2.9920E-05 0.0171 99.8400 3.2521E-05 0.0243 99.8643 3.5123E-05 0.0171 99.8814 3.7725E-05 0.0143 99.8957 4.0327E-05 0.0114 99.9071 4.2928E-05 0.0086 99.9157 4.5530E-05 0.0100 99.9257 4.8132E-05 0.0057 99.9314 5.0733E-05 0.0043 99.9357 5.3335E-05 0.0043 99.9400 5.5937E-05 0.0029 99.9429 5.8539E-05 0.0014 99.9443 6.1140E-05 0.0014 99.9457 6.3742E-05 0.0086 99.9543 6.6344E-05 0.0014 99.9557 6.8946E-05 0.0071 99.9629 7.1547E-05 0.0029 99.9657 7.4149E-05 0.0014 99.9671 7.6751E-05 0.0014 99.9686 7.9352E-05 0.0029 99.9714 8.7158E-05 0.0029 99.9743 8.9759E-05 0.0014 99.9757 9.2361E-05 0.0014 99.9771 9.4963E-05 0.0014 99.9786 9.7564E-05 0.0014 99.9800 1.0797E-04 0.0029 99.9829 1.1057E704 0.0029 99.9857 1.1838E-04 0.0014 99.9871 1.2098E-04 0.0029 99.9900 1.2879E-04 0.0014 99.9914 1.3919E-04 0.0029 99.9943 1.5480E-04 0.0029 99.9971 WCAP- 16168-NP-A June 2008 Revision 2

E-6 E-1: 10 Year ISI Only (cont.) 1.8082E-04 0.0014 99.9986 2.3546E-04 0.0014 100.0000

                             ==       Summary Descriptive Statistics             ==

Minimum = 0.OOOOE+00 Maximum = 2.3451E-04 Range = 2.3451E-04 Number of Simulations - 70000 5th Percentile = 0.OOOOE+00 Median = 1.0978E-08 95.0th Percentile = 1.3009E-06 99.0th Percentile = 8.7970E-06 99.9th Percentile = 3.8701E-05 Mean = 5.9461E-07 Standard Deviation = 3.3139E-06 Standard Error = 1.2525E-08 Variance (unbiased) = 1.0982E-11 Variance (biased) = 1.0982E-11 Moment Coeff. of Skewness = 2.4476E+01 Pearson's 2nd Coeff. of Skewness = 5.3829E-01 Kurtosis = 1.0028E+03

• PROBABILITY DISTRIBUTION FUNCTION (HISTOGRAM) *
• FOR THROUGH-WALL CRACKING FREQUENCY (FAILURE)
• FREQUENCY OF RELATIVE CUMULATIVE TWC FAILURES DENSITY DISTRIBUTION (PER REACTOR-OPERATING YEAR) (%) (%)

0.OOOOE+00 31.5414 31.5414 4.5123E-08 67.4686 99.0100 1.3537E-07 0.5000 99.5100 2.2561E-07 0.1786 99.6886 3.1586E-07 0.0843 99.7729

4. 0611E-07 0 .0543 99.8271 4.9635E-07 0.0400 99.8671 5.8660E-07 0.0257 99.8929 6.7684E-07 0. 0186 99.9114 7.6709E-07 0 .0157 99.9271 8.5733E-07 0.0100 99.9371 9.4758E-07 0.0100 99.9471 WCAP- 16168-NP-A June 2008 Revision 2

E-7 E-1: 10 Year ISI Only (cont.) 1.0378E-06 0.0100 99.9571 1.1281E-06 0.0029 99.9600 1.2183E-06 0.0029 99.9629 1.3086E-06 0 .0014 99.9643 1 .4891E-06 0. 0014 99.9657 1.5793E-06 0.0043 99.9700 1.6695E-06 0. 0014 99.9714 1.7598E-06 0.0043 99.9757

1. 8500E-06 0.0014 99.9771

1. 9403E-06 0.0014 99.9786

2. 0305E-06 0.0043 99.9829 2 .2110E-06 0.0014 99 .9843

2. 3013E-06 0.0029 99 .9871

2. 7525E-06 0. 0014 99. 9886 2 .8427E-06 0.0043 99.9929 3 .2940E-06 0 .0014 99.9943 3 .4745E-06 0. 0014 99.9957 4.1964E-06 0. 0014 99. 9971 5.3696E-06 0 .0014 99.9986 8.8892E-06 0 .0014 100.0000

                              -=       Summary Descriptive    Statistics              ==

Minimum = 0.OOOOE+00 Maximum = 8.9343E-06 Range = 8.9343E-06 Number of Simulations = 70000 5th Percentile = 0.OOOOE+00 Median = 3.2629E-13 95.0th Percentile = 4.5123E-08 99.0th Percentile = 8.9609E-08 99.9th Percentile = 6.2131E-07 Mean = 5.0396E-09 Standard Deviation = 6.7097E-08 Standard Error = 2.5360E-10 Variance (unbiased) = 4.5020E-15 Variance (biased) = 4.5019E-15 Moment Coeff. of Skewness = 5.9654E+01 Pearson's 2nd Coeff. of Skewness =-2.6549E-01 Kurtosis = 5.8120E+03

• FRACTIONALIZATION OF FREQUENCY OF CRACK INITIATION *
• AND THROUGH-WALL CRACKING FREQUENCY (FAILURE) - *
• WEIGHTED BY TRANSIENT INITIATING FREQUENCIES
• WCAP-16168-NP-A June2008 Revision 2

E-8 E-1: 10 Year ISI Only (cont.)

                                      % of total      % of total frequency of    frequency of crack initiation of TWC failure 2       0.00             0.00 3       0.01             0.03 7       2.62             1.25 9       1.18             0.37 14        0.00             0.00 31        0.00             0.00 34        0.00             0.00 56       93.99            43.32 59        0.00             0.00 60        0. 08            8.80 61        0  .01            0.23 62        0.00              0. 02 64        0.00              0.00 65        0.00              0.00 66        0.01              0.00 67        0.00              0.01 68        0.00              0.00 69        0.00              0.23 70        0.00              0. 03 71        0.01              1.29 72        0.00              0.00 73        0.00              0.00 74        0.00              0.00 76        0.00              0.00 78        0.00              0.00 81        0.00              0.00 82        0.00              0.00 83        0.00              0.00 92        0.01              0.00 93        0.01              0. 01 94        0.00              0.00 97        0.05              6.07 102         0.03              0.08 103         0.05              0.44 104          0.03             0.04 105         0.00              0.00 106          0.00             0.00 107          0.00             0.02 108          0.00             0.00 109          0.00             0.00 110          0.00             0.00 ill          0.00             0.00 112          0.00             0.00 113          0. 00            0.00 114          0.72             0.46 115          0.00             0.00 116          0.03             0. 03 117          0  .14           0.08 118          0.00             0.00 119          0.00             0.00 120          0.00             0.42 WCAP- 16168-NP-A                                                    June 2008 Revision 2

E-9 E-1: 10 Year ISI Only (cont.) 121 0.00 0.01 122 0.00 0.00 123 0.02 1.34 124 0.01 0 .14 125 0.00 0.00 126 0.10 10.65 127 0.18 0.04 128 0.15 0.04 129 0.21 2.56 130 0.32 22.00 TOTALS 100.00 100.00 DATE: 03-May-2007 TIME: 16:03:48 June 2008 WCAP-16168-NP-A June 2008 Revision 2

E-10 E-2: ISI Every 10 Years WELCOME TO FAVOR *

• FRACTURE ANALYSIS OF VESSELS: OAK RIDGE
• VERSION 06.1 *
• FAVPOST MODULE: POSTPROCESSOR MODULE
• COMBINES TRANSIENT INITIAITING FREQUENCIES *
• WITH RESULTS OF PFM ANALYSIS
• PROBLEMS OR QUESTIONS REGARDING FAVOR
• 0 SHOULD BE DIRECTED TO
• TERRY DICKSON
• OAK RIDGE NATIONAL LABORATORY
• e-mail: dicksontl@ornl.gov *
• This computer program was prepared as an account of *
• work sponsored by the United States Government *
• Neither the United States, nor the United States *
• Department of Energy, nor the United States Nuclear *
• Regulatory Commission, nor any of their employees, *
• nor any of their contractors, subcontractors, or their *
• employees, makes any warranty, expressed or implied, or *
• assumes any legal liability or responsibility for the *
• accuracy, completeness, or usefulness of any *
• information, apparatus, product, or process disclosed, *
• or represents that its use would not infringe *
• privately-owned rights.
• DATE: 03-May-2007 TIME: 15:03:08 Begin echo of FAVPost input data deck 15:03:08 03-May-2007 End echo of FAVPost input data deck 15:03:08 03-May-2007 FAVPOST INPUT FILE NAME = postbv.in FAVPFM OUTPUT FILE CONTAINING PFMI ARRAY = INITIATE.DAT FAVPFM OUTPUT FILE CONTAINING PFMF ARRAY = FAILURE.DAT FAVPOST OUTPUT FILE NAME = 70000.out WCAP- 16168-NP-A June 2008 Revision 2

B-Il E-2: ISI Every 10 Years (cont.)

• NUMBER OF SIMULATIONS = 70000
• CONDITIONAL PROBABILITY CONDITIONAL PROBABILITY OF INITIATION CPI=P(IIE) OF FAILURE CPF=P(FIE)

MEAN 95th % 99th % MEAN 95th % 99th % RATIO TRANS IENT CPF CPF CPF CPFmn/CPImn NUMBER CPI CPI CPI -------

                                                                                                        ----- --- ------I
            ------------------------------                                I 2      0.OOOOE+00                         0.OOOOE+00     0.OOOOE+00       0.OOOOE+00      0.OOOOE+00      0.OOOOE+00    0.0000 3      2.9800E-07                         0.OOOOE+00     0.OOOOE+00       1. 0009E-11     0 OOOOE+00      0 OOOOE+00    0.0000 7         2.3214E-03                      5.3036E-03     3. 1228E-02      9.5073E-06      1. 6615E-04     1. 5375E-04   0.0041 9     23.4412E-03                         8.2533E-03     4.6458E-02       9.7557E-06      1. 2533E-04     1. 6701E-04   0.0028 14      9. 1393E-09                        0.OOOOE+00     0.OOOOE+00       0.OOOOE+00      0 .OOOOE+00     0 OOOOE+00    0.0000 31     :2.9830E-06                         0.OOOOE+00     5. 6087E-07      6. 0712E-09     0 OOOOE+00      0 OOOOE+00    0.0020 34     23.5119E-06                         0.OOOOE+00     1. 7572E-06      8. 9484E-10     0 OOOOE+00      0 .OOOOE+00   0.0003 56      "

3.4601E-03 7. 9216E-03 3. 9985E-02 1.6170E-05 2. 0726E-04 2. 7235E-04 0.0047 59 C .0000E+00 0.OOOOE+00 0 OOOOE+00 0.OOOOE+00 0.OOOOE+00 0 OOOOE+00 0.0000 60 2 .0168E-05 3 .2408E-04 1.3589E-04 1.9908E-05 3.0787E-04 1. 3395E-04 0.9871 61 2 2.7685E-05 8 .4773E-04 3 .2463E-04 6.6560E-06 1.7053E-04 6. 1162E-05 0.2404 62 7 7.5949E-06 0 OOOOE+00 1. 9721E-05 4.0857E-06 0.OOOOE+00 8. 8105E-07 0.5380 64 2 *.2973E-04 2 .4655E-03 3. 8380E-03 1.4364E-07 0.OOOOE+00 1.7605E-09 0.0006 65 2 *.2001E-08 0 .OOOOE+00 0 OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.0000 66 2 *.6002E-05 8 .4773E-04 2. 8721E-04 2.3979E-07 0.OOOOE+00 1.4232E-08 0.0092 67 2 *.0237E-06 0 OOOOE+00 1.0314E-08 4.1104E-10 0.OOOOE+00 0.OOOOE+00 0.0002 68 3 3.1241E-07 0 OOOOE+00 0.OOOOE+00 2.5031E-07 0 OOOOE+00 0.OOOOE+00 0.8012 69 E .2793E-04 2 3821E-03 8. 8419E-03 4.7965E-04 2. 3081E-03 6. 1921E-03 0.7639 70 2 .6032E-04 2 .4655E-03 4. 5217E-03 3.5399E-05 4. 5620E-04 3. 0528E-04 0.1360 71 1 . 1361E-05 0.OOOOE+00 2. 1871E-05 1.1359E-05 0 OOOOE+00 2. 1795E-05 0.9998 72 C .OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0 OOOOE+00 0 OOOOE+00 0.0000 73 2 . 6310E-06 0.OOOOE+00 6.8862E-06 1.1986E-08 0 .OOOOE+00 0 .OOOOE+00 0.0046 74 1 . 9735E-07 0.0000E+00 0.OOOOE+00 4. 1257E-09 0 OOOOE+00 0 OOOOE+00 0.0209 76 C OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0 OOOOE+00 0 OOOOE+00 0.0000 78 C .OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0 OOOOE+00 0 OOOOE+00 0.0000 81 C . OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.0000 82 2 .3896E-12 0.OOOOE+00 0.OOOOE+00 1.4423E-12 0.OOOOE+00 0.OOOOE+00 0.6036 WCAP- 16168-NP-A June 2008 Revision 2

E- 12 E-2: ISI Every 10 Years (cont.) 83 0.OOOOE+00 0 OOOOE+00 0 OOOOE+00 0.OOOOE+00 0 OOOOE+00 0.000OE+00 0.0000 92 2.2151E-04 2. 3171E-03 3. 9602E-03 1.4445E-06 6. 6444E-05 6.9896E-06 0.0065 93 2.2151E-04 2. 3171E-03 3. 9602E-03 1.4435E-06 6. 6444E-05 6.9896E-06 0.0065 94 5.0580E-13 0 .OOOOE+00 0 .OOOOE+00 0.OOOOE+00 0 .OOOOE+00 0.OOOOE+00 0.0000 97 5.6862E-05 1. 055E-03 5. 9676E-04 5.6592E-05 1 1055E-03 5. 9102E-04 0.9953 102 2. 3159E-06 0 OOOOE+00 0 OOOOE+00 1.7235E-08 0 OOOOE+00 0 OOOOE+00 0. 0074 103 2.7193E-05 1 .3062E-03 2. 3097E-04 2.2622E-06 0 OOOOE+00 2. 7032E-06 0.0832 104 2 3159E-06 0 OOOOE+00 0 OOOOE+00 1.7235E-08 0 OOOOE+00 0 OOOOE+00 0. 0074 105 3. 0512E-07 0.OOOOE+00 0.OOOOE+00 6 . 1501E-09 0.OOOOE+00 0 OOOOE+00 0.0202 106 2 .2119E-06 0.OOOOE+00 0.OOOOE+00 2.0044E-08 0.OOOOE+00 0 OOOOE+00 0.0091 107 2. 6872E-05 1.4079E-03 1.8487E-04 2.9860E-06 0.OOOOE+00 4 3421E-06 0.1111 108 6. 2467E-10 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0 OOOOE+00 0.0000 109 1. 1837E-07 0.OOOOE+00 0 .OOOOE+00 3.2204E-11 0 OOOOE+00 0 OOOOE+00 0.0003 110 6. 2467E-10 0.OOOOE+00 0 OOOOE+00 0.OOOOE+00 0 OOOOE+00 0 OOOOE+00 0.0000 ill 1. 1837E-07 0.OOOOE+00 0 .OOOOE+00 3.2204E-11 0 OOOOE+00 0 OOOOE+00 0.0003 112 6. 2651E-10 0.OOOOE+00 0 .OOOOE+00 2. 9991E-18 0 OOOOE+00 0 OOOOE+00 0.0000 113 6. 1401E-07 0.OOOOE+00 0 .OOOOE+00 2. 1350E-09 0 OOOOE+00 0.000OE+00 0.0035 114 2.4706E-05 7.7763E-04 3 1629E-04 5 .7733E-08 0 OOOOE+00 0.OOOOE+00 0.0023 115 0.OOOOE+00 0.OOOOE+00 0 .OOOOE+00 0 .OOOOE+00 0 OOOOE+00 0.OOOOE+00 0.0000 116 5.5954E-06 0.OOOOE+00 1 .1859E-05 2. 5625E-09 0 .OOOOE+00 0.OOOOE+00 0.0005 117 1 .2355E-04 1.5759E-03 2.1069E-03 8. 5564E-07 3. 3720E-05 3.4261E-06 0.0069 118 2. 7278E-08 0.OOOOE+00 0.OOOOE+00 0 OOOOE+00 0 OOOOE+00 0.OOOOE+00 0.0000 119 4. 5739E-06 0.OOOOE+00 3.4377E-06 1. 5432E-08 0 .OOOOE+00 0.OOOOE+00 0.0034 120 2. 0556E-05 4.1262E-04 1.0487E-04 1. 6115E-05 3. 6562E-04 5.0320E-05 0.7839 121 1. 8168E-04 2 .1666E-03 3. 1673E-03 1 .5012E-06 7.2990E-05 2.5174E-06 0. 0083 122 1. 8168E-04 2 .1666E-03 3. 1673E-03 4.0757E-07 0.OOOOE+00 7. 5126E-08 0. 0022 123 5 0111E-04 2. 1906E-03 6.5400E-03 3.8406E-04 2.0773E-03 4.9104E-03 0.7664 124 2. 0576E-04 2. 1666E-03 3. 6134E-03 2 . 8451E-05 6.1944E-04 1.8793E-04 0.1383 125 5.2120E-08 0 .OOOOE+00 0 .OOOOE+00 2. 5396E-13 0.OOOOE+00 0.OOOOE+00 0.0000 126 3.4102E-06 0 .OOOOE+00 7. 9349E-07 3.2667E-06 0.OOOOE+00 6.8717E-07 0.9579 127 2.6738E-05 2. 5191E-04 3. 1581E-04 2. 6252E-12 0.OOOOE+00 0.OOOOE+00 0.0000 128 2.6738E-05 2. 5191E-04 3. 1581E-04 2.6252E-12 0.OOOOE+00 0.OOOOE+00 0.0000 129 2.9639E-05 2. 5191E-04 3. 7032E-04 2. 9210E-06 0.OOOOE+00 7.9206E-07 0.0986 130 5.5626E-05 9. 5051E-04 7. 2434E-04 3.2049E-05 9. 4101E-04 2.4706E-04 0.5762 NOTES: CPI IS CONDITIONAL PROBABILITY OF CRACK INITIATION, P(IIE) CPF IS CONDITIONAL PROBABILITY OF TWC FAILURE, P(FIE) WCAP- 16168-NP-A June 2008 Revision 2

E-13 E-2: ISI Every 10 Years (cont.)

• PROBABILITY DISTRIBUTION FUNCTION (HISTOGRAM) *
• FOR THE FREQUENCY OF CRACK INITIATION
• FREQUENCY OF RELATIVE CUMULATIVE CRACK INITIATION DENSITY DISTRIBUTION (PER REACTOR-OPERATING YEAR) (%) (%)

0.0000E+00 16.2857 16.2857 1.2622E-06 78.7129 94.9986 3.7866E-06 2.5743 97.5729 6.3111E-06 0.9743 98.5471 8.8355E-06 0.4871 99.0343 1.1360E-05 0.2614 99.2957 1.3884E-05 0.1829 99.4786 1.6409E-05 0.1029 99.5814 1.8933E-05 0.0900 99.6714 2.1458E-05 0.0643 99.7357 2.3982E-05 0.0400 99.7757 2.6506E-05 0.0314 99.8071 2.9031E-05 0.0329 99.8400 3.1555E-05 0.0171 99.8571 3.4080E-05 0.0200 99.8771 3.6604E-05 0.0143 99.8914 3.9129E-05 0.0086 99.9000 4.1653E-05 0.0114 99.9114 4.4177E-05 0.0086 99.9200 4.6702E-05 0.0129 99.9329 4.9226E-05 0.0014 99.9343 5.1751E-05 0.0071 99.9414 5.4275E-05 0.0043 99.9457 5.6800E-05 0.0043 99.9500 5.9324E-05 0.0043 99.9543 6.1848E-05 0.0043 99.9586 6.6897E-05 0.0029 99.9614 6.9422E-05 0.0014 99.9629 7.1946E-05 0.0029 99.9657 7.6995E-05 0.0014 99.9671 8.4568E-05 0.0029 99.9700 8.7093E-05 0.0014 99.9714 8.9617E-05 0.0029 99.9743 9.2141E-05 0.0029 99.9771 9.4666E-05 0.0014 99.9786 9.7190E-05 0.0029 99.9814 1.0224E-04 0.0014 99.9829 1.0729E-04 0.0014 99.9843 1.0981E-04 0.0014 99.9857 1.1739E-04 0.0014 99.9871 1.2243E-04 0.0014 99.9886 1.3506E-04 0.0014 99.9900 1.5020E-04 0.0014 99.9914 1.5525E-04 0.0014 99.9929 WCAP- 16168-NP-A June 2008 Revision 2

E-14 E-2: ISI Every 10 Years (cont.) 1.5778E-04 0.0014 99.9943 1.7797E-04 0.0014 99.9957 2.2089E-04 0.0014 99.9971 2.2846E-04 0.0014 99.9986 2.3098E-04 0.0014 100.0000

                            ==        Summary Descriptive Statistics            ==

Minimum = 0.OOOOE+00 Maximum = 2.3200E-04 Range = 2.3200E-04 Number of Simulations = 70000 5th Percentile = 0.OOOOE+00 Median = 9.7197E-09 95.0th Percentile = 1.2622E-06 99.0th Percentile = 8.6578E-06 99.9th Percentile = 3.9129E-05 Mean = 5.8049E-07 Standard Deviation = 3.4364E-06 Standard Error = 1.2989E-08 Variance (unbiased) = 1.1809E-11 Variance (biased) = 1.1809E-11 Moment Coeff. of Skewness = 2.8192E+01 Pearson's 2nd Coeff. of Skewness = 5.0677E-01 Kurtosis = 1.3237E+03

• PROBABILITY DISTRIBUTION FUNCTION (HISTOGRAM) *
• FOR THROUGH-WALL CRACKING FREQUENCY (FAILURE)
• FREQUENCY OF RELATIVE CUMULATIVE TWC FAILURES DENSITY DISTRIBUTION (PER REACTOR-OPERATING YEAR) (96) (%)

0.OOOOE+00 32.3229 32.3229 3.7808E-08 66.6543 98.9771 1.1342E-07 0.4457 99.4229 1.8904E-07 0.2186 99.6414 2.6466E-07 0.0914 99.7329 3.4027E-07 0.0514 99.7843

4. 1589E-07 0.0343 99.8186 4.9151E-07 0.0271 99.8457

5. 6712E-07 0.0171 99.8629 WCAP- 16168-NP-A June 2008 Revision 2

E-15 E-2: ISI Every 10 Years (cont.) 6.4274E-07 0.0186 99.8814

7. 1836E-07 0. 0171 99.8986 7.9397E-07 0. 0143 99.9129 8.6959E-07 0.0043 99 .9171 9.4521E-07 0.0086 99.9257 1.0208E-06 0. 0043 99.9300 1.0964E-06 0.0043 99.9343 1.1721E-06 0.0071 99. 9414 1.2477E-06 0.0043 99.9457 1.3989E-06 0.0029 99.9486 1.4745E-06 0.0057 99.9543

1. 5501E-06 0.0014 99. 9557 1.6258E-06 0. 0057 99.9614 1.7014E-06 0.0029 99.9643 1.7770E-06 0. 0043 99.9686 2.0038E-06 0.0029 99. 9714 2.1551E-06 0. 0014 99.9729

2. 3063E-06 0. 0014 99.9743

2. 6844E-06 0. 0014 99. 9757

2. 9869E-06 0.0014 99 .9771

3. 0625E-06 0.0014 99 .9786

3. 1381E-06 0.0014 99.9800 3 .2893E-06 0.0029 99.9829 3 .5162E-06 0.0014 99.9843

3. 6674E-06 0.0014 99.9857 3.8186E-06 0.0014 99.9871 3.9699E-06 0.0029 99.9900 4.0455E-06 0.0014 99.9914 4.2723E-06 0. 0014 99.9929 5.2554E-06 0. 0014 99.9943 5.5578E-06 0.0029 99.9971

6. 0115E-06 0. 0014 99.9986 7.4482E-06 0.0014 100.0000

                             ==         Summary Descriptive    Statistics               ==

Minimum = 0.OOOOE+00 Maximum = 7.4860E-06 Range = 7.4860E-06 Number of Simulations = 70000 5th Percentile = 0.OOOOE+00 Median = 2.5574E-13 95.0th Percentile = 3.7808E-08 99.0th Percentile = 4.1686E-08 99.9th Percentile = 7.2592E-07 Mean = 5.2336E-09 Standard Deviation = 8.2534E-08 WCAP-16168-NP-A June 2008 Revision 2

E-16 E-2: ISI Every 10 Years (cont.) Standard Error = 3.1195E-10 Variance (unbiased) = 6.8118E-15 Variance (biased) = 6.8117E-15 Moment Coeff. of Skewness = 4.7286E+01 Pearson's 2nd Coeff. of Skewness =-1.6306E-01 Kurtosis = 2.9147E+03

• FRACTIONALIZATION OF FREQUENCY OF CRACK INITIATION *
• AND THROUGH-WALL CRACKING FREQUENCY (FAILURE) *
• WEIGHTED BY TRANSIENT INITIATING FREQUENCIES *

                                          % of total        % of total frequency of      frequency of crack initiation    of TWC failure 2           0.00                0.00 3           0.00                0.00 7           2.58                1.01 9           1.13                0.39 14            0.00                0.00 31            0.00                0.00 34            0.00                0.00 56           94.32               43.78 59            0.00                0.00 60            0.10               11.13 61            0.01                0.28 62            0.00                0.01 64            0.00                0.00 65            0.00                0.00 66            0.01                0.01 67            0.00                0.00 68            0.00'               0.00 69            0.00                0.22 70            0.00                0.02 71            0.01                0.78 72            0.00                0.00 73            0.00                0.00 74            0.00                0.00 76            0.00                0.00 78            0.00                0.00 81            0.00                0.00 82            0.00                0.00 83            0.00                0.00 92            0.01                0.01 93            0.01                0.01 94            0.00                0.00 97           0.05                5.07 102             0.07                0.12 103             0.06                0.51 104             0.05                0.04 105             0.00                0.00 WCAP- 16168-NP-A                                                           June 2008 Revision 2

E-17 E-2: ISI Every 10 Years (cont.) 106 0.00 0.00 107 0.00 0.03 108 0.00 0.00 109 0.00 0.00 110 0.00 0.00 ill 0.00 0.00 112 0.00 0.00 113 0.01 0.00 114 0.47 0.13 115 0.00 0.00 116 0.01 0.00 117 0.13 0.09 118 0.00 0.00 119 0.00 0.00 120 0.00 0.34 121 0.00 0.00 122 0.00 0.00 123 0 .02 1.33 124 0 .01 0.10 125 0.00 0.00 126 0.09 10.04 127 0.13 0.00 128 0.15 0.00 129 0.20 1.27 130 0.36 23.28 TOTALS 100.00 100.00

• FRACTIONALIZATION OF FREQUENCY OF CRACK INITIATION *
• AND THROUGH-WALL CRACKING FREQUENCY (FAILURE) -
• BY
• RPV BELTLINE MAJOR REGION
• BY PARENT SUBREGION *
• WEIGHTED BY % CONTRIBUTION OF EACH TRANSIENT *
• TO FREQUENCY OF CRACK INITIATION AND *
• THROUGH-WALL CRACKING FREQUENCY (FAILURE) *

                                                                       % of total 6    of        % of total             through-wall crack MAJOR      RTndt     total          frequency of                   frequency REGION      (MAX)    flaws        crack initiation      cleavage ductile total 1     174.15      2.29                0.11              2.82        0.02     2.84 2     174.15      2.29                0.10              2.15        0.03     2.18 3     164.48      3.69                4.16             25.64        4.45   30.09 4      164.48     3.69                3.65             20.74        4.15   24.90 5      89.30    19.28               89.66              2.88        0.05     2.93 WCAP- 16168-NP-A                                                                          June 2008 Revision 2

E- 18 E-2: ISI Every 10 Years (cont.) 6 220.82 13.16 0.08 0.82 0.07 0.89 7 192.65 13.16 0.00 0.00 0.00 0.00 8 253.03 21.22 2.06 32.01 2.57 34.58 9 223.03 21.22 0.18 1.36 0.24 1.60 TOTALS 100.00 100.00 88.43 11.57 100.00

• FRACTIONALIZATION OF FREQUENCY OF CRACK INITIATION *
• AND THROUGH-WALL CRACKING FREQUENCY (FAILURE) - *
• BY *
• RPV BELTLINE MAJOR REGION *
• BY CHILD SUBREGION *
• WEIGHTED BY % CONTRIBUTION OF EACH TRANSIENT *
• TO FREQUENCY OF CRACK INITIATION AND *
• THROUGH-WALL CRACKING FREQUENCY (FAILURE) *

                                                               % of total
                            % of       % of total          through-wall crack MAJOR        RTndt     total     frequency of              frequency REGION        (MAX)    flaws   crack initiation   cleavage ductile total 1       174.15     2.29           0.00           0.00      0.00       0.00 2       174 .15    2.29           0.00           0.00      0.00       0.00 3       164.48     3.69           0.00           0.00      0.00       0.00 4       164.48     3.69           0.00           0.00      0.00       0.00 5        89.30    19.28           0.00           0.00      0.00       0.00 6       220.82    13 .16          2.89           5.80      0.12       5 .92 7       192.65    13 .16           0.00          0.00      0.00        0.00 8       253 .03   21.22          86.26         81.11      11.21    92 .32 9      223.03    21.22          10.85           1.51      0.25        1.76 TOTALS 100.00           100.00         88.43      11.57  100.00

• FRACTIONALIZATION OF FREQUENCY OF CRACK INITIATION *
• AND THROUGH-WALL CRACKING FREQUENCY (FAILURE) - *
• MATERIAL, FLAW CATEGORY, AND FLAW DEPTH *
• WEIGHTED BY % CONTRIBUTION OF EACH TRANSIENT *
• TO FREQUENCY OF CRACK INITIATION AND *
• THROUGH-WALL CRACKING FREQUENCY (FAILURE)
• I WCAP- 16168-NP-A June 2068 Revision 2

B-i19 E-2: ISI Every 10 Years (cont.)

• WELD MATERIAL *

                            % of total     frequency             % of total      through-wall of crack initiation                       crack frequency FLAW DEPTH      CAT I        CAT 2        CAT 3        CAT 1        CAT 2    CAT 3 (in)       flaws        flaws        flaws        flaws        flaws    flaws 0.080        0.00         1.68         0.00         0.00         0.06       0.00 0.161        0.00       38.20          0.00         0.00         2.93       0.00 0.241        0.00        18.39         0.00         0.00         2.38       0.00 0.321        0.00         9.55         0.00         0.00         1.39       0.01 0.402        0.00         7.76         0.00         0.00         2.20       0.03 0.482        0.00         6.12         0.00         0.00         2.57       0.08 0.563        0.00         4.14         0.00         0.00         3.19       0. 08 0.643        0.00         2.73         0.00         0.00         2 .74      0 .13 0.723        0.00         2.03         0.00         0.00         2.78       0 .12 0.804        0.00         1.35         0.00         0.00         3 .51      0.30 0.884        0.00         1.02         0.00         0.00         3.28       0.26

0. 964 0.00 0.67 0.00 0.00 2.93 0.19

1. 045 0.00 1.08 0.00 0.00 2.91 0.07 1.125 0.00 0.40 0.00 0.00 2.84 0.07 1.205 0.00 0.30 0.00 0.00 2.57 0.06 1.286 0.00 0.21 0.00 0 .00 4.21 0 .44 1.366 0.00 0.34 0.01 0.00 1.35 0.80 1.446 0.00 0.16 0.00 0.00 4.13 0.16

1. 527 0.00 0.42 0.00 0.00 2.18 0.05 1.607 0.00 0.21 0.00 0.00 4.06 0.29 1.688 0.00 0.29 0.00 0.00 1.86 0 .14 1.768 0.00 0.02 0.00 0.00 0.61 0.39 1.848 0.00 0.01 0.01 0.00 1.15 0.66 1.929 0.00 0.53 0.00 0.00 0.69 0.10 TOTALS 0.00 97.63 0.05 0.00 58.51 4.42

                                               *P********  **E****

• PLATE MATERIAL *

                             % of total     frequency             % of total      through-wall of crack initiation                       crack frequency FLAW DEPTH       CAT I        CAT 2        CAT 3        CAT 1       CAT 2     CAT 3 (in)       flaws        flaws        flaws        flaws        flaws    flaws 0.080        0.00        0.01         0.00          0.00         0.13       0.00 0.161        0.00        0.48         0.00          0.00         7.33       0.00 0.241        0.00        0.69         0.00          0.00       10.42        0.00 0.321        0.00        0.61         0.00          0.00         9.99       0.00 WCAP- 16168-NP-A                                                                               June 2008 Revision 2

E-20 E-2: ISI Every 10 Years (cont.) 0.402 0.00 0.54 0.00 0.00 9.20 0.00 0.482 0.00 0.00 0.00 0.00 0.00 0.00 0.563 0.00 0.00 0.00 0.00 0.00 0.00 0.643 0.00 0.00 0.00 0.00 0.00 0.00 0.723 0.00 0.00 0.00 0.00 0.00 0.00 0.804 0.00 0.00 0.00 0.00 0.00 0.00 0.884 0.00 0.00 0.00 0.00 0.00 0.00 0.964 0.00 0.00 0.00 0.00 0.00 0.00

1. 045 0.00 0.00 0.00 0.00 0.00 0.00 TOTALS 0.00 2.33 0.00 0.00 37.06 0.00 DATE: 03-May-2007 TIME: 15:04:07 WCAP- 16168-NP-A June 2008 Revision 2

F-1 APPENDIX F INPUTS FOR THE PALISADES PILOT PLANT EVALUATION WCAP- 16168-NP-A June 2008 Revision 2

F-2 A summary of the NDE inspection history based on Regulatory Guide 1.150 and pertinent input data for Palisades is as follows:

1. Number of ISis performed (relative to initial pre-service and 10-year interval inspections) for full penetration Category B-A and B-D reactor vessel welds assuming all of the candidate welds were inspected: 2 (covering all welds of the specified categories).

2. The inspections performed covered: 100% for 13 Category B-A welds, >90% but <100% for 6 Category B-A welds, <90% for 8 Category B-A welds, and 100% of all Category B-D welds.

3. Number of indications found during most recent inservice inspection: 11 This number includes consideration of the following additional information:

a. Indications found that were reportable: 0

b. Indications found that were within acceptable limits: 11 C. Indications/anomalies currently being monitored: 0

4. Full penetration relief requests for the RV submitted and accepted by the NRC: 2 relief requests for limited converage for 12 welds

5. Fluence distribution at inside surface of RV beltline until end of life (EOL): see Figure F-I taken from the NRC PTS Risk Study [9], Figure 4.3.

h~25 AS< 930 1:3 180 25 271 315 ý361

               .4      ,.
                          -11  .4P$*

Figure F-I Rollout Diagram of Beltline Materials and Representative Fluence Maps for Palisades WCAP-16168-NP-A June 2008 Revision 2

F-3

6. Reactor vessel cladding details:

a. Thickness: 0.25 inches

b. Material properties are identified in Table F-i:

Table F-I Cladding Material Properties Specific Young's Thermal Thermal Heat Modulus of Expansion Conductivity (Btu/LBM- Elasticity Coefficient Temperature (Btu/hr-ft-°F) °F) (KSI) (OF-l) (LBM/fDensity Poisson'S (OF ) "K ' .. C' '"E" "acl.... 'Iv,, 0 - 489 .3 68 - - 22045.7 489 .3 70 8.1 0.1158 - 489 .3 100 8.4 0.1185 - 8.55E-06 489 .3 150 8.6 0.1196 - 8.67E-06 489 .3 200 8.8 0.1208 - 8.79E-06 489 .3 250 9.1 0.1232. - 8.9E-06 489 .3 300 9.4 0.1256 - 9.OE-06 489 .3 302 - - 20160.2 489 .3 350 9.6 0.1258 - 9.1E-06 489 .3 400 9.9 0.1281 9.19E-06 489 .3 450 10.1 0.1291 - 9.28E-06 489 .3 482 - - 18419.8 489 .3 500 10.4 0.1305 - 9.37E-06 489 .3 550 10.6 0.1306 - 9.45E-06 489 .3 600 10.9 0.1327 - 9.53E-06 489 .3 650 11.1 0.1335 - 9.61E-06 489 .3 700 11.4 0.1348 - 9.69E-06 489 .3 750 11.6 0.1356 - 9.76E-06 489 .3 800 11.9 0.1367 - 9.82E-06 489 .3

c. Material including copper and nickel content: Material properties assigned to clad flaws are that of the underlying material be it base metal or weld. These properties are identified in Table F-3. This is consistent with the NRC PTS Risk Study [8, 9].

d. Material property uncertainties:

1) Bead width: 1 inch - bead widths vary for all plants. Based on the NRC PTS Risk Study [8, 9], a nominal dimension of 1 inch is selected for all analyses because this parameter is not expected to influence significantly the predicted vessel failure probabilities.

2) Truncation limit: Cladding thickness rounded to the next 1/100th of the total reactor vessel thickness to be consistent with the NRC PTS Risk Study [8, 9].

3) Surface flaw depth: 0.263 inch

4) All flaws are surface-breaking. Only flaws in cladding that would influence brittle fracture of the reactor vessel are brittle. This is consistent with the NRC PTS Risk Study [8, 9].

WCAP-.16168-NP-A June 2008 Revision 2

F-4

e. Additional cladding properties are identified in Table F-4. This is consistent with the NRC PTS Risk Study [8, 9].

7. Base metal:

a. Wall thickness: 8.5 inches

b. Material properties are identified in Tables F-2 and F-3:

Table F-2 Base Metal Material Properties Specific Young's Thermal Thermal Heat Modulus of Expansion Conductivity (Btu/LBM- Elasticity Coefficient Temperature (Btu/hr-ft-°F) OF) (KSI) (OF-') Density Poisson's 3 Temeraure(LBM~ft ) Ratio (OF) "K' ' "E".. .. ' I 0 - - 489 .3 70 24.8 0.1052 29200 489 .3 100 25 0.1072 7.06E-06 489 .3 150 25.1 0.1101 - 7.16E-06 489 .3 200 25.2 0.1135 28500 7.25E-06 489 .3 250 25.2 0.1166 - 7.34E-06 489 .3 300 25.1 0.1194 28000 7.43E-06 489 .3 350 25 0.1223 - 7.5E-06 489 .3 400 25.1 0.1267 27400 7.58E-06 489 .3 450 24.6 0.1277 - 7.63E-06 489 .3 500 24.3 0.1304 27000 7.7E-06 489 .3 550 24 0.1326 - 7.77E-06 489 .3 600 23.7 0.135 26400 7.83E-06 489 .3 650 23.4 0.1375 - 7.9E-06 489 .3 700 23 0.1404 25300 7.94E-06 489 .3 750 22.6 0.1435 - 8.OE-06 489 .3 800 22.2 0.1474 23900 8.05E-06 489 .3 WCAP- 16168-NP-A June 2008 Revision 2

F-5 Table F-3 Palisades-Specific Material Values Drawn from the RVID (see Ref. 44 Table 4.1) Major Material Region Description Un-Irradiated RTNDT Cu Ni P Mn

     #         Type          Heat        Location    Iwt%J    Iwt%I     Iwt%]      Iwt%1         [OF]

1 Axial Weld 3-112A lower 0.213 1.010 0.019 1:315 -56 2 Axial Weld 3-112B lower 0.213 1.010 0.019 1.315 -56 3 Axial Weld 3-112C lower 0.213 1.010 0.019 1.315 -56 4 Axial Weld 2-112A upper 0.213 1.010 0.019 1.315 -56 5 Axial Weld 2-112B upper 0.213 1.010 0.019 1.315 - 56 6 Axial Weld 2-112C upper 0.213 1.010 0.019 1.315 -56 7 Circ Weld 9-112 intermediate 0.203 1.018 0.013 1.147 -56 8 Plate D3804-1 lower 0.190 0.480 0.016 1.235 0 9 Plate D3804-2 lower 0.190 0.500 0.015 1.235 -30 10 Plate D3804-3 lower 0.120 0.550 0.010 1.270 -25 11 Plate D3803-1 upper 0.240 0.510 0.009 1.293 -5 12- Plate D3803-2 upper 0.240 0.520 0.010 1.350 -30 13 Plate D3803-3 upper 0.240 0.500 0.011 1.293 -5

8. Weld metal details: Details of information used in addressing weld-specific information are taken directly from the NRC PTS Risk Study [44], Table 4.2. Summaries are reproduced as Table F-4.

Values for SAW Weld Volume fraction and Repair Weld Volume fraction in Table F-4 were changed to 96.7% and 2.3% respectively per NUREG-1874 [9]. WCAP-16168-NP-A .June 2008 Revision 2

F-6 Table F-4 Summary of Reactor Vessel-Specific Inputs for Flaw Distribution Base Metal Thickness [in] 8,438 7.875 8.5 8.675 Vessel specific info Total Wall Thickness [in] 8.626 8.031 8.75 8988 Vessel specific info Volume fraction [%1 97% 100% - SMAW% - REPAIR% Thru-Wall Bead in] 0.1875 01875 0.1875 0.1875 All plants report plant specific Thickness dimensions of 3!16:in. Judgment. Approx. 2X the Truncation Limit [in] 1size of the largest non-repair flaw observed in PVRUF & Shoreham. Buried or Surface -- All flaws are buried Observation Observation: Virtually all of the weld flaws in PVRUF & Orientation Circ flaws in circ welds, axial flaws in axial Shoreham were'aligned with welds, the welding direction because they were tack of sidewall SAW fusion defects. Weld Density basis -- Shoreham density Highest of observations Statistically similar distributions from Shoreham and PVRUF were combined to provide more robust Aspect ratio Shoreham & PVRUF observations estimates, when based on basis judgment theamount-data were limited andlor insufficient to identifydifferent trends for aspect ratios for flaws in the two vessels. Statistically similar distributions combined to Depth basis Shoreham & PVRUF observations provide more robust

                                                    ..                                     estimates June 2008 WCAP- 16168-NP-A WCAP-16168-NP-A                                                                                                      June 2008 Revision 2

F-7 Table F-4 Summary of Reactor Vessel-Specific Inputs for Flaw Distribution (cont.) upper bound to all plant specific info provided by Volume fraction 1% Steve Byme (Westinghouse - Windsor),

                                              -Oconee                                          is generic value based on average of all Thru-Wall Bead                                                            plants specific values ThickWnes Bsa        [in] 0.21           020          0.22         0.25   (including Shoreham &

PVRUF data), Other values are plant specific as reported by Steve Byrne. Judgment. Approx. 2X the Trunc nLimit [i] size of the largest non-repair Truncation iflaw observed in PVRUF & Shoreham, Buried or Surface All flaws are buried Observation Observation: Virtually all of the weld flaws in PVRUF & SMAW O'ientation Circ flaws in circ welds, axial flaws in axial Shoreham were aligned with Weld - welds. the welding direction because they were lack of sidewall fusion defects. Density basis - Shoreham density Highest of observations Statistically similar distributions from Shoreham and PVRRUF were combined

                                                                                    'to provide more robust Aspect ratio                  'Shoreham & PVRUF observations              estimates, when based on basis"                         Shh....vi                                  judgment the amount data were liriied and/or
                                                                                    *insufficient to identify different trends for aspect ratios for
            .....                                                                    flaws .inthe two Vessels.

Statistically similar distributions combined to Depth basis Shoreham & PVRUF observations provide more robust estimates WCAP- 16168-NP-A *. June 2008 Revision 2

F-8 Table F-4 Summary of Reactor Vessel-Specific Inputs for Flaw Distribution (cont.) jUUylwlgt. MI IUUIIUeU Repair integral percentage that Volume fraction [%] exceeds the. repaired volume Weld observed for Shoreham and for PVRUF. which was 1.5%. Thru-Wall Bead Generic value: As observed Thruckn ead [in] 0,14 in PVRUF and Shoreham by Thickness PNNL. Judgment. Approx. 2X the largest repair flaw found in Truncation Limit tin] .2 PVRUF &Shoreham. Also basedon maximum expected

                                                    ..                            width of repair cavity.

Buried or Surface All flaws are buried Observation The repair flaws had complex shapes and orientations that were not aligned, with either the axial or circumferential Cire flaws in circ welds, axial flaws in axial welds; for consistency with Orientation Csthe welds.. available treatments

                                                                                                          -    of flaws by the FAVOR code,.a common treatment of orientations was adopted for flaws in SAW/SMAW and
                             .__   _ __.........repair                                    welds.

Density basis - . Shoreham density Highest of observations Statistically similar distributions from Shoreham and PVRUFwere combined to provide more irobust Aspect ratio Shoreham & PVRUF observations estimates, when based on* basis judgment the amount data were limited and/or insufficient to identify different trends for aspect ratios for flaws in the two vessels. Statistically similar distributions combined to Depth basis Shoreham & PVRUF observations provide more robust estimates. WCAP-16168-NP-A June 2008 Revision'2

F-9 Table F-4 Summary of Reactor Vessel-Specific Inputs for Flaw Distribution (cont.) Cladding TActual Thickness [in) 0.188 0.156 0.25 0.313 Vessel specific info I '-+------------ ~

              ##ofLayers
                                +/-L]          i-      .       21              2               2   Vessel specific info Bead widths of 1 to 5-in, characteristic of machine deposited cladding, Bead widths down to 24-in, can occur over welds, Nominal dimension of 1-in. selected for all analyses because this Bead Width        fin]                            1 parameter is not expected to influence significantly the predicted vessel failure probabilities. May need to refine this estimate later, particularly for Oconee who reo~orted a 5-in bead width.

Actual clad thickness rounded to the nearest Truncation Limit [in] 1/100& of the total vessel wall thickness Judgment & computational. Surface flaw convenience depth in FAVOR [in] 0.259 0.161 0.263 0360 Judgment. Only flaws in cladding that would influence brittle fracture of the vessel Buried or Surface All flaws are surface breaking are brittle. Material properties assigned to clad flaws are that of the underlying material, be it base or weld. Observation:. All flaws observed in PVRUF & Shoreham were lack of inter-Orientation All circumferential. run fusion defects, and cladding is always deposited circumferentiallv No surface flaws observed. Density is 111000 that of the .observed buried flaws in Density basis -- cladding of vessels examined by PNNL. If Judgment there is more than one clad layer then there are.no clad flaws. Aspect ratio: -O Observations on buried flaws Judgment ba sis . . ... . . . .. . . .. Depth of all surfaceflaws is the actual clad Depth basis thickness rounded up to the nearest 1/1100 Judgment. of the total.vessel wall thickness. WCAP- 16168-NP-A June 2008 Revision 2

F-10 Table F-4 Summary of Reactor Vessel-Specific Inputs for Flaw Distribution (cont.) Truncation Limit tin] 0,433 of the largest flaw observed in

                                                                                        .all P.NNLpttje inspejtrons-_

Buried or Surface All flaws are buried Observation 4 ~ Observation &Physics: No Half of the simulated flaws are observed orientation Orientation preference, and no reason to circumferential, half are axial. Plate suspect one (other than laminations which are benicin. 1110Dweld of small weld flaw density, 1/40.of.large flaw density of the PVRUF data Judgment. Supported by limited data. Aspect ratio Same as for PVRUF welds Judgment basis . ... Judgment. Supported by Depth basis Same as for PVRUF welds limited data. limited data.

9. TWCF95-TOTAL value calculated at 60 EFPY using correlations from NUREG-1874 (Reference 9):

3.16E-7 Events per year WCAP- 16168-NP-A June 2008 Revision 2

G-1 APPENDIX G PALISADES PROBSBFD OUTPUT WCAP- 16168-NP-A June 2008 Revision 2

G-2 G-1: 10 Year ISI Only STRUCTURAL RELIABILITY AND RISK ASSESSMENT (SRRA) WESTINGHOUSE MONTE-CARLO SIMULATION PROGRAM PROBSBFD VERSION 1.0 INPUT VARIABLES FOR CASE 2: PAL 10 YEAR ISI ONLY NCYCLE = 80 NFAILS = 1001 NTRIAL = 1000 NOVARS = 19 NUMSET = 2 NUMISI = 5 NUMSSC = 4 NUMTRC = 4 NUMFMD = 4 VARIABLE DISTRIBUTION MEDIAN DEVIATION SHIFT USAGE NO. NAME TYPE LOG VALUE OR FACTOR MV/SD NO. SUB 1 FIFDepth - CONSTANT - 3.OOOOD-02 1 SET 2 IFlawDen - CONSTANT - 3j6589D-03 2 SET 3 ICy-ISI - CONSTANT - 1. OOOOD+01 1 ISI 4 DCy-ISI - CONSTANT - 8 . OOOOD+01 2 ISI 5 MV-Depth - CONSTANT - 1.5000D-02 3 ISI 6 SD-Depth - CONSTANT - 1. 8500D-01 4 ISI 7 CEff-ISI - CONSTANT - 1.000OD+00 5 ISI 8 Aspectl - CONSTANT - 2.000OD+00 1 SSC 9 Aspect2 - CONSTANT - 6 .OOOOD+00 2 SSC 10 Aspect3 - CONSTANT - 1. OOOOD+01 3 SSC 11 Aspect4 - CONSTANT - 9 9000D+01 4 SSC 12 NoTr/Cy - CONSTANT - 1. 3000D+01 1 TRC 13 FCGTh1d - CONSTANT - 1 .5000D+00 2 TRC 14 FCGR-UC NORMAL NO 0 OOOOD+00 1.OOOOD+00 .00 3 TRC 15 DKINFile - CONSTANT - 1.0000D+00 4 TRC 16 Percentl - CONSTANT - 7. 8870D+01 1 FMD 17 Percent2 - CONSTANT - 1. 0720D+01 2 FMD 18 Percent3 - CONSTANT - 4. 3807D+00 3 FMD 19 Percent4 - CONSTANT - 6. 0298D+00 4 FMD INFORMATION GENERATED FROM FAVLOADS.DAT FILE AND SAVED IN DKINSAVE.DAT FILE: WALL THICKNESS = 8.7500 INCH FLAW DEPTH MINIMUM K AND MAXIMUM K FOR TYPE 1 WITH AN ASPECT RATIO OF 2. 8.7500OD-02 2.69285D+00 1.08492D+01 1 .61000D-01 3.60064D+00 1.46562D+01 4 .37500D-01 1.26609D+01 2.00367D+01

6. 56250D-01 1.49279D+01 2.39231D+01

8. 75000D-01 1.53491D+01 2.67406D+01

1. 75000D+00 1.37876D+01 3.14212D+01

2. 62500D+00 8.13906D+00 3.01520D+01

4. 37500D+00 -2.32655D+00 2.91175D+01 TYPE 2 WITH AN ASPECT RATIO OF 6.

8.75000D-02 4.04516D+00 1.64003D+01 June 2008 WCAP- 16168-NP-A WCAP-16168-NP-A June 2008 Revision 2

G-3 C-I: 10 Year ISI Only (cont.) 1.61000D-01 5.52109D+00 2 .25832D+01 4.37500D-01 1.80126D+01 3. 03772D+01 6.5625OD-01 2.31235D+01 3 .61026D+01 8.75000D-01 2.65795D+01 4 .11957D+01 1.75000D+00 2.62424D+01 5. 18633D+01 2.62500D+00 2.10650D+01 5.45640D+01 4.37500D+00 9.61580D+00 5.85179D+01 TYPE 3 WITH AN ASPECT RATIO OF 10. 8.7500OD-02 4.43154D+00 1.79837D+01 1.61000D-01 5.90218D+00 2.41564D+01 4.37500D-01 1.90406D+01 3.24750D+01 6.56250D-01 2.45354D+01 3.85918D+01 8.7500OD-01 2.87821D+01 4.40958D+01 1.75000D+00 2.91774D+01 5.64674D+01 2.62500D+00 2.54877D+01 6.25646D+01 4.37500D+00 1.38132D+01 7.03917D+01 TYPE 4 WITH AN ASPECT RATIO OF 99. 8.75000D-02 7.10780D+00 1.85180D+01 1.75000D-01 1.00487D+01 2.59141D+01

2. 62500D-01 1.38195D+01 2.86661D+01

4. 37500D-01 2.16458D+01 3.45538D+01 6.56250D-01 2.85157D+01 4.23747D+01

8. 75000D-01 3.03911D+01 4.83133D+01 1.75000D+00 3.36289D+01 6.57043D+01

2. 62500D+00 3.16032D+01 7.68320D+01 "AVERAGE CALCULATED VALUES FOR: Surface Flaw Density with FCG and ISI NUMBER FAILED = 0 NUMBER OF TRIALS = 1000 DEPTH (WALL/400) AND FLAW DENSITY FOR ASPECT RATIOS OF 2, 6, 10 AND 99 12 2.5402D-04 4.5317D-06 1.3489D-06 2. 1739D-06 13 5.8986D-06 1. 9521D-05 7.3792D-06 9. 7312D-06 14 0.0000D+00 7.0234D-06 3.2977D-06 4.3086D-06 15 0 OOOOD+00 1.9775D-06 1. 1450D-06 1.7029D-06 16 0 .0000D+00 5.8037D-07 4.0809D-07 6. 4975D-07 17 0 OOOOD+00 3.4736D-07 1. 5441D-07 2. 2919D-07 18 0 OOOOD+00 1.5414D-07 8.8627D-08 1. 7208D-07 19 0.0000D+00 9. 1024D-08 6.1738D-08 5. 0696D-08 20 0. 0000D+00 0.OOOOD+00 3.6375D-08 8. 2449D-08 21 0 .0000D+00 0.OOOOD+00 0.OOOOD+00 3.2256D-08 23 0 OOOOD+00 2.7971D-08 0.OOOOD+00 0 .OOOOD+00 25 0.OOOOD+00 0.OOOOD+00 1.1041D-08 0.OOOOD+00 26 0.0000D+00 2 .6821D-08 0.OOOOD+00 0.OOOOD+00 27 0.OOOOD+00 0.0000D+00 0.OOOOD+00 1.4338D-08 29 0.0000D+00 0.OOOOD+00 1.0518D-08 0.OOOOD+00 31 0.OOOOD+00 0.0000D+00 0.0000D+00 1.3440D-08 WCAP-16168-NP-A June 2008 Revision 2

G-4 G-2: ISI Every 10 Years STRUCTURAL RELIABILITY AND RISK ASSESSMENT (SRRA) WESTINGHOUSE MONTE-CARLO SIMULATION PROGRAM PROBSBFD VERSION 1.0 INPUT VARIABLES FOR CASE 2: PAL 10 YEAR INT NCYCLE = 80 NFAILS = 1001 NTRIAL = 1000 NOVARS = 19 NUMSET = 2 NUMISI = 5 NUMSSC = 4 NUMTRC = 4 NUMFMD = 4 VARIABLE DISTRIBUTION MEDIAN DEVIATION SHIFT USAGE NO. NAME TYPE LOG VALUE OR FACTOR MV/SD NO. SUB 1 FIFDepth - CONSTANT - 3.OOOOD-02 1 SET 2 IFlawDen - CONSTANT - 3.6589D-03 2 SET 3 ICy-ISI - CONSTANT - 1. 0000D+01 1 ISI 4 DCy-ISI - CONSTANT - 1. OOOOD+01 2 ISI 5 MV-Depth - CONSTANT - 1.500OD-02 3 IS11 6 SD-Depth - CONSTANT - 1. 8500D-01 4 ISI 7 CEff-ISI - CONSTANT - 1. 0000D+00 5 ISI 8 Aspectl - CONSTANT - 2 .0000D+00 1 SSC 9 Aspect2 - CONSTANT - 6. OOOOD+00 2 SSC 10 Aspect3 - CONSTANT - 1. 0000D+01 3 SSC 11 Aspect4 - CONSTANT - 9. 9000D+01 4 SSC 12 NoTr/Cy - CONSTANT - 1. 3000D+01 1 TRC 13 FCGThld - CONSTANT - 1.5000D+00 2 TRC

   .14    FCGR-UC         NORMAL     NO       0 .OOOOD+00   1.0000D+00       .00     3  TRC 15    DKINFile         -  CONSTANT  -     1 .0000D+00                            4  TRC 16    Percentl         -  CONSTANT  -     7. 8870D+01                            1  FMD 17    Percent2         -  CONSTANT  -     1.0720D+01                             2   FMD:

18 Percent3 - CONSTANT - 4.3807D+00 3 FMDi 19 Percent4 - CONSTANT - 6.0298D+00 4 FMD INFORMATION GENERATED FROM FAVLOADS.DAT FILE AND SAVED IN DKINSAVE.DAT FILE: WALL THICKNESS = 8.7500 INCH FLAW DEPTH MINIMUM K AND MAXIMUM K FOR TYPE 1 WITH AN ASPECT RATIO OF 2. 8.75000D-02 2.69285D+00 1.08492D+01 1.61000D-01 3.60064D+00 1.46562D+01 4.37500D-01 1.26609D+01 2.00367D+01 6.56250D-01 1.49279D+01 2. 39231D+01 8.75000D-01 1.53491D+01 2 .67406D+01 1.75000D+00 1.37876D+01 3. 14212D+01 2.62500D+00 8.13906D+00 3. 01520D+01 4.37500D+00 -2.32655D+00 2. 91175D+01 TYPE 2 WITH AN ASPECT RATIO OF 6. 8.75000D-02 4.04516D+00 1.64003D+01 WCAP-16168-NP-A June 2008 Revision 2

G-5 G-2: ISI Every 10 Years (cont.)

1. 61000D-01 5.52109D+00 2.25832D+01

4. 37500D-01 1.80126D+01 3.03772D+01

6. 56250D-01 2.31235D+01 3.61026D+01

8. 75000D-01 2.65795D+01 4.11957D+01 1.75000D+00 2.62424D+01 5.18633D+01

2. 62500D+00 2.10650D+01 5.45640D+01 4 .37500D+00 9.61580D+00 5.85179D+01 TYPE 3 WITH AN ASPECT RATIO OF 10.

8.75000D-02 4.43154D+00 1.79837D+01 1.61000D-01 5.90218D+00 2.41564D+01 4.37500D-01 1.90406D+01 3.24750D+01 6.56250D-01 2.45354D+01 3.85918D+01 8.75000D-01 2.87821D+01 4.40958D+01 1.75000D+00 2.91774D+01 5.64674D+01 2.62500D+00 2.54877D+01 6.25646D+01 4.37500D+00 1.38132D+01 7.03917D+01 TYPE 4 WITH AN ASPECT RATIO OF 99. 8.75000D-02 7.10780D+00 1.85180D+01 1.75000D-01 1 .00487D+01 2.59141D+01 2.62500D-01 1. 38195D+01 2.86661D+01 4.37500D-01 2 .16458D+01 3.45538D+01 6.56250D-01 2 .85157D+01 4.23747D+01 8.75000D-01 3. 03911D+01 4.83133D+01 1.75000D+00 3.36289D+01 6.57043D+01 2.62500D+00 3. 16032D+01 7.68320D+01 AVERAGE CALCULATED VALUES FOR: Surface Flaw Density with FCG and ISI NUMBER FAILED = 0 NUMBER OF TRIALS = 1000 DEPTH (WALL/400) AND FLAW DENSITY FOR ASPECT RATIOS OF 2, 6, 10 AND 99 12 1.2465D-I0 1.8940D-12 5. 5678D-13 9. 1111D-13 13 1.9983D-12 5. 5048D-12 2. 0459D-12 2. 7226D-12 14 0.OOOOD+00 9. 6570D-13 4. 5289D-13 5. 8811D-13 15 0.OOOOD+00 1.2835D-13 7. 5032D-14 1. 0930D-13 16 0.OOOOD+00 1. 8170D-14 1.2594D-14 1.8759D-14 17 0.OOOOD+00 5.2179D-15 2.1701D-15 2.9926D-15 18 0.OOOOD+00 9.4118D-16 6.6938D-16 9. 6145D-16 19 0.OOOOD+00 2. 9809D-16 1.7580D-16 1.4879D-16 20 0.OOOOD+00 0.OOOOD+00 4.8987D-17 9. 2976D-17 21 0.OOOOD+00 0.OOOOD+00 0.OOOOD+00 1.4658D-17 23 0.OOOOD+00 2.2110D-18 0.OOOOD+00 0 OOOOD+00 25 0.OOOOD+00 0.OOOOD+00 1.5152D-19 0 OOOOD+00 26 0.OOOOD+00 2.1470D-19 0.OOOOD+00 0 OOOOD+00 27 0.OOOOD+00 0.OOOOD+00 0.OOOOD+00 2 .4461D-20 29 0.OOOOD+00 0.OOOOD+00 7. 9308D-21 0 OOOOD+00 31 0.OOOOD+00 0.000OD+00 0.OOOOD+00 5.2922D-22 WCAP-16168-NP-A June 2008 Revision 2

H-1 APPENDIX H PALISADES PTS TRANSIENTS Table H-I PTS Transient Descriptions for Palisades Count TH System Failure Operator Action Mean IE HZP HiK Dominant* Case Frequency 1 2 3.59 cm (1.414 in) surge line None 2.66E-04 No Yes No break. Containment sump recirculation included in the analysis. 2 16 Turbine/reactor trip with 2 Operator starts second AFW 1.23E-04 No No No stuck-open ADVs on SG-A pump. Operator isolates AFW combined with controller to affected SG at 30 minutes failure resulting in the flow after initiation. Operator from two AFW pumps into assumed to throttle HPI if affected steam generator. 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 %. 3 18 Turbine/reactor trip with 1 Operator does not isolate AFW 4.71E-03 No No No stuck-open ADV on SG-A. on affected SG. Normal AFW Failure of both MSIVs (SG- flow assumed (200 gpm). A and SG-B) to close. 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 %. 4 19 Reactor trip with 1 stuck- None. Operator does not. 2.29E-03 Yes No Yes at 60, open ADV on SG-A. throttle HPI. 200, 500 EFPY 5 22 Turbine/reactor trip with loss Operator depressurizes through 6.67E-05 No No No of MFW and AFW. ADVs and feeds SG's using condensate booster pumps. Operators maintain a cooldown rate within technical specification limits and throttle condensate flow at 84 % level in the steam generator. 6 24 Main steam line break with None 2.43E-06 No No No the break assumed to be inside containment causing containment spray actuation. WCAP- 16168-NP-A June 2008 Revision 2

H-2 Table H-1 PTS Transient Descriptions for Palisades Count TH System Failure Operator Action Mean IE HZP HiK Dominant Case Frequency 7 26 Main steam line break with Operator isolates AFW to 5.69E-04 No No No the break assumed to be affected SG at 30 minutes after inside containment causing initiation. containment spray actuation. 8 27 Main steam line break with Operator starts second AFW 3.65E-05 No No No controller failure resulting in pump. the flow from two AFW pumps into affected steam generator. Break assumed to be inside containment causing containment spray actuation. 9 29 Main steam line break with None. Operator does not 4.20E-08 Yes No No break assumed to be inside throttle HPI. containment causing containment spray actuation. 10 31 Turbine/reactor trip with Operator maintains core cooling 1.29E-05 No No No failure of MFW and AFW. by "feed and bleed" using HPI Containment spray actuation to feed and two PORVs to assumed due to PORV bleed. discharge. 11 32 Turbine/reactor trip with Operator maintains core cooling 1.08E-06 No No No failure of MFW and AFW. by "feed and bleed" using HPI Containment spray actuation to feed and two PORV to bleed. assumed due to PORV AFW is recovered 15 minutes discharge. after initiation of "feed and bleed" cooling. Operator closes PORVs when SG level reaches 60 percent. ,_ 12 34 Main steam line break Operator isolates AFW to 1.48E-05 No No No concurrent with a single tube affected SG at 15 minutes after failure in SG-A due to MSLB initiation. Operator trips RCPs vibration, 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 subcooling > 25 F. HPI is throttled to maintain pressurizer level between 40 and 60 %. 13 40 40.64 cm (16 in) hot leg None. Operator does not 3.22E-05 No Yes Yes at 32, break. Containment sump throttle HPI. 60, 200, recirculation included in the 500 EFPY analysis. WCAP- 16168-NP-A June 2008 Revision 2

H-3 Table H-1 PTS Transient Descriptions for Palisades Count TH System Failure Operator Action Mean IE HZP HiK Dominant* Case Frequency 14 42 Turbine/reactor trip with two Operator assumed to throttle 7.67E-07 No No No stuck open pressurizer SRVs. HPI if auxiliary feedwater is Containment spray is running with SG wide range assumed not to actuate. level > -84% and RCS subcooling > 25 F. HPI is throttled to maintain pressurizer level between 40 and 60 %. 15 48 Two stuck-open pressurizer None. Operator does not 7.67E-07 Yes No Yes at 32 SRVs that reclose at 6000 sec throttle HPI. EFPY after initiation. Containment spray is assumed not to actuate. 16 49 Main steam line break with Operator isolates AFW to 1.00e-05 Yes No No the break assumed to be affected SG at 30 minutes after inside containment causing initiation. Operator does not containment spray actuation. throttle HPI. 17 50 Main steam line break with Operator starts second AFW 5.81E-07 Yes No No controller failure resulting in pump. Operator does not the flow from two AFW throttle HPI. pumps into affected steam generator. Break assumed to be inside containment causing containment spray actuation. 18 51 Main steam line break with Operator does not isolate AFW 7.51E-08 Yes No No failure of both MSIVs to on affected SG. Operator does close. Break assumed to be not throttle HPI. inside containment causing containment spray actuation. 19 52 Reactor trip with 1 stuck- Operator does not isolate AFW 6.37E-04 Yes No Yes at 500 open ADV on SG-A. Failure on affected SG. Nonnal AFW EFPY of both MSIVs (SG-A and flow assumed (200 gpm). SG-B) to close. Operator does not throttle HPI. 20 53 Turbine/reactor trip with two None. Operator does not 1.09E-03 No No Yes at 500 stuck-open pressurizer SRVs throttle HPI. EFPY that reclose at 6000 sec after initiation. Containment spray is assumed not to actuate. 21 54 Main steam line break with Operator does not isolate AFW 4.26E-06 No No Yes at 32, failure of both MSIVs to on affected SG. Operator does 60, 200, close. Break assumed to be not throttle HPI. 500 EFPY inside containment causing containment spray actuation. WCAP- 16168-NP-A June 2008 Revision 2

H-4 Table H-1 PTS Transient Descriptions for Palisades Count TH System Failure Operator Action Mean IE HZP HiK Dominant* Case Frequency 22 55 Turbine/reactor trip with 2 Operator starts second AFW 2.74E-03 No No Yes at 32, stuck-open ADVs on SG-A pump. 60, 200, combined with controller 500 EFPY failure resulting in the flow from two AFW pumps into affected steam generator. 23 58 10.16 cm (4 in) cold leg None. Operator does not 2.66E-04 No Yes Yes at 32, break. Winter conditions throttle HPI. 60,200, assumed (HPI and LPI 500 EFPY injection temp = 40 F, Accumulator temp = 60 F) 24 59 10.16 cm (4 in) cold leg None. Operator does not 2.09E-04 No Yes Yes at 500 break. Sunmner conditions throttle HPI. EFPY assumed (HPI and LPI injection temp = 100 F, Accumulator temp = 90 F) 25 60 5.08 cm (2 in) surge line None. Operator does not 2.09E-04 No Yes Yes at 60, break. Winter conditions throttle HPI. 200, 500 assumed (HPI and LPI EFPY injection temp = 40 F, Accumulator temp = 60 F) 26 61 7.18 cm (2.8 in) cold leg None. Operator does not 2.09E-04 No Yes No break. Summer conditions throttle HPI. assumed (HPI and LPI injection temp = 100 F, Accumulator temp = 90 F) 27 62 20.32 cm (8 in) cold leg None. Operator does not 7.07E-06 No Yes Yes at 32, break. Winter conditions throttle HPI. 60, 200, assumed (HPI and LPI 500 EFPY injection temp = 40 F, Accumulator temp = 60 F) 28 63 14.37 cm (5.656 in) cold leg None. Operator does not 6.06E-06 No Yes Yes at 60, break. Winter conditions throttle HPI. 200, 500 assumed (HPI and LPI EFPY injection temp = 40 F, Accumulator temp = 60 F) 29 64 10.16 cm (4 in) surge line None. Operator does not 7.07E-06 No Yes Yes at 32, break. Summer conditions throttle HPI. 60,200, assumed (HPI and LPI 500 EFPY injection temp = 100 F, Accumulator temp = 90 F) 30 65 One stuck-open pressurizer None. Operator does not 1.24E-04 Yes No Yes at 32, SRV that recloses at 6000 sec throttle HPI. 60, 200, after initiation. Containment 500 EFPY spray is assumed not to actuate. WCAP-16168-NP-A June 2008 Revision 2

H-5 Notes:

1. TH ### - Thermal hydraulics run number ###

2. LOCA- Loss-of-coolant accident

3. SBLOCA- Small-break loss-of-coolant accident

4. MBLOCA - Medium-break loss-of-coolant accident

5. LBLOCA - Large-break loss-of-coolant accident

6. HZP - Hot-zero power

7. ADV - Atmospheric dump valve

8. SRV - Safety and relief valve

9. MSLB - Main steam line break

10. AFW - Auxiliary feedwater

11. HPI - High-pressure injection

12. RCP- Reactor coolant pump

13. SG - Steam generator

• The arbitrary definition of a dominant transient is a transient that contributes 1% or more of the total Through-Wall Cracking Failure (TWCF).

June 2008 WCAP- 161 68-NP-A WCAP-16168-NP-A June 2008 Revision 2

1-1 APPENDIX I PALISADES FAVPOST OUTPUT WCAP- 16168-NP-A June 2008 Revision 2

1-2 I-1: 10 Year ISI only

• WELCOME TO FAVOR *
• FRACTURE ANALYSIS OF VESSELS: OAK RIDGE *
• VERSION 06.1 *
• FAVPOST MODULE: POSTPROCESSOR MODULE *
• COMBINES TRANSIENT INITIAITING FREQUENCIES *
• WITH RESULTS OF PFM ANALYSIS *
• PROBLEMS OR QUESTIONS REGARDING FAVOR *
• SHOULD BE DIRECTED TO *
• TERRY DICKSON *
• OAK RIDGE NATIONAL LABORATORY *
• e-mail: dicksontl@ornl.gov *
• This computer program was prepared as an account of *
• work sponsored by the United States Government *
• Neither the United States, nor the United States *
• Department of Energy, nor the United States Nuclear *
• Regulatory Commission, nor any of their employees, *
• nor any of their contractors, subcontractors, or their *
• employees, makes any warranty, expressed or implied, or *
• assumes any legal liability or responsibility for the *
• accuracy, completeness, or usefulness of any *
• information, apparatus, product, or process disclosed, *
• or represents that its use would not infringe *
• privately-owned rights.
• DATE: 27-Jun-2007 TIME: 10:25:13 Begin echo of FAVPost input data deck 10:25:13 27-Jun-2007 End echo of FAVPost input data deck 10:25:13 27-Jun-2007 FAVPOST INPUT FILE NAME = postpl.in FAVPFM OUTPUT FILE CONTAINING PFMI ARRAY = INITIATE.DAT FAVPFM OUTPUT FILE CONTAINING PFMF ARRAY = FAILURE.DAT FAVPOST OUTPUT FILE NAME = plpostlOyronly.out WCAP- 16168-NP-A

1-3 I-1: 10 Year ISI only (cont.)

• NUJMBER OF SIMULATIONS = 7.00,00
• CO]NDITIONAL PROBABILITY CONDITIONAL PROBABILITY I-3 OF INITIATION CPI=P(IIE) OF FAILURE CPF=P(FIE)

TRANSIEN T MEAN 95th % 99th % MEAN 95th % 99th % RATIO CPI CPI CPF CPF CPF CPFmn/CPImn NUMBER CPI 2 0.OOOOE+00 0.OOOOE+00 0 OOOOE+00 0.OOOOE+00 0 OOOOE+00 0.OOOOE+00 0.0000 16 2.0991E-09 0.OOOOE+00 0 OOOOE+00 2.6647E-10 0 .OOOOE+00 0.0000E+00 0.1269 18 4.1893E-11 0.OOOOE+00 0 OOOOE+00 2. 0525E-11 0 .OOOOE+00 0.0000E+00 0.4899 19 5.1457E-07 0.OOOOE+00 0 .OOOOE+00 3.4107E-07 0 OOOOE+00 0.OOOOE+00 0.6628 22 6. 0538E-09 0.OOOOE+00 0. 0000E+00 1.7933E-09 0 OOOOE+00 0 OOOOE+00 0.2962 24 7. 2366E-07 0.OOOOE+00 0 OOOOE+00 8. 5491E-08 0 OOOOE+00 0 OOOOE+00 0.1181 26 7. 2366E-07 0 OOOOE+00 0.OOOOE+00 9.3663E-08 0.OOOOE+00 0 .OOOOE+00 0.1294 27 2. 0421E-05 4 3132E-04 2. 0173E-04 5.9362E-06 1. 6195E-04 6 .0880E-05 0.2907 29 6. 8282E-07 0 OOOOE+00 0.OOOOE+00 4.6244E-07 0.OOOOE+00 0 OOOOE+00 0.6772 31 2 .3780E-05 2. 9665E-04 2.6221E-04 5.4087E-06 9.2535E-05 5. 7966E-05 0.2274 32 4 .3551E-07 0 .OOOOE+00 0.OOOOE+00 3.0097E-07 0 OOOOE+00 0 .OOOOE+00 0.6911 34 3 .4459E-06 0 OOOOE+00 1. 1896E-06 6.4355E-07 0 OOOOE+00 1.6486E-07 0.1868 40 6. 0211E-03 1. 2838E-02 6.9324E-02 3.8794E-04 8. 0216E-04 5. 1013E-03 0.0644 42 0 .0000E+00 0 OOOOE+00 0 .OOOOE+00 0.OOOOE+00 0 OOOOE+00 0.0000E+00 0.0000 48 5. 4079E-04 2 .4464E-03 7 3391E-03 5. 3190E-04 2 .4464E-03 7.2226E-03 0.9835 49 2. 8452E-07 0 OOOOE+00 0 OOOOE+00 4.9008E-08 0 OOOOE+00 0.OOOOE+00 0. 1722 50 4. 4689E-05 6. 9298E-04 6 .6733E-04 1.4807E-05 2. 7591E-04 2.2267E-04 0.3313 51 2. 3653E-04 1.5112E-03 3. 3702E-03 1.1828E-04 8. 5384E-04 1.6646E-03 0.5001 52 6. 0157E-07 0.OOOOE+00 0 OOOOE+00 3.9569E-07 0 .OOOOE+00 0.OOOOE+00 0.6578 53 2.6708E-07 0.OOOOE+00 0 .OOOOE+00 1.8524E-07 0 OOOOE+00 0.0000E+00 0.6936 54 4.7146E-04 2.0579E-03 6. 9604E-03 2.1455E-04 1. 1565E-03 3.1205E-03 0.4551 55 1.3039E-06 0.OOOOE+00 0.OOOOE+00 9. 5266E-07 0.OOOOE+00 0.OOOOE+00 0.7306 58 3.9842E-04 1.9057E-03 5.2952E-03 7.5527E-05 5. 0201E-04 8.2070E-04 0.1896 59 1.6591E-05 2.2922E-04 1.5697E-04 1. 1995E-06 3. 1036E-05 9.8600E-06 0.0723 60 4.2014E-05 7.4348E-04 5.9014E-04 5. 3889E-06 8. 0611E-05 7. 5170E-05 0.1283 61 9.0696E-07 0.OOOOE+00 2 .2070E-10 3. 9814E-08 0.OOOOE+00 1. 5568E-12 0.0439 62 4.8426E-03 1.0080E-02 5.6235E-02 5. 1473E-04 1.0533E-03 6. 6194E-03 0.1063 63 1.7066E-03 3.3225E-03 2.3408E-02 2 .4011E-04 8.3884E-04 3. 0140E-03 0.1407 64 1.9466E-03 4.0456E-03 2.5502E-02 2. 8279E-04 5.7833E-04 3.9028E-03 0.1453 WCAP-16168-NP-A June 2008 Revision 2

1-4 I-1: 10 Year ISI only (cont.) 65 1.7674E-04 1.5370E-03 2.8655E-03 1.7065E-04 1.5297E-03 2.7897E-03 0.9655 NOTES: CPI IS CONDITIONAL PROBABILITY OF CRACK INITIATION, P(IIE) CPF IS CONDITIONAL PROBABILITY OF TWC FAILURE, P(FIE)

• PROBABILITY DISTRIBUTION FUNCTION (HISTOGRAM) *
• FOR THE FREQUENCY OF CRACK INITIATION
• FREQUENCY OF RELATIVE CUMULATIVE CRACK INITIATION DENSITY DISTRIBUTION (PER REACTOR-OPERATING YEAR) (%) (%)

0.OOOOE+00 1.7957 1.7957 1.8255E-06 95.9914 97.7871 5.4764E-06 1.1900 98.9771 9.1273E-06 0.3786 99.3557 1.2778E-05 0.1929 99.5486 1.6429E-05 0.1129 99.6614 2.0080E-05 0.0829 99.7443 2.3731E-05 0.0514 99.7957 2.7382E-05 0.0357 99.8314 3.1033E-05 0.0343 99.8657 3.4684E-05 0.0143 99.8800 3.8335E-05 0.0171 99.8971 4.1986E-05 0.0171 99.9143 4.5636E-05 0.0129 99.9271 4.9287E-05 0.0071 99.9343 5.2938E-05 0.0086 99.9429 5.6589E-05 0.0086 99.9514 6.0240E-05 0.0043 99.9557 6.3891E-05 0.0014 99.9571 6.7542E-05 0.0043 99.9614 WCAP- 16168-NP-A June 2008 Revision 2

                                                                                           '-5 I-1: 10 Year ISI only (cont.)

7. 1193E-05 0.0029 99.9643 7.4844E-05 0.0014 99.9657 7.8495E-05 0.0043 99.9700

8. 2146E-05 0.0029 99.9729 8.9447E-05 0.0029 99.9757 9.3098E-05 0.0014 99. 9771 9.6749E-05 0.0014 99. 9786 1.0040E-04 0.0014 99.9800 1.0405E-04 0.0029 99.9829 1.0770E-04 0. 0014 99.9843

1. 1500E-04 0 .0014 99.9857 1.1865E-04 0.0029 99.9886 1.3326E-04 0 .0014 99.9900

1. 3691E-04 0. 0014 99.9914 1.4056E-04 0. 0014 99.9929 1.4786E-04 0. 0014 99.9943 1.5516E-04 0. 0014 99.9957 2.0263E-04 0.0014 99.9971

3. 0120E-04 0 .0014 99.9986 3.5962E-04 0.0014 100.0000

                              ==       Summary Descriptive Statistics           ==

Minimum = 0.OOOOE+00 Maximum = 3.6144E-04 Range = 3.6144E-04 Number of Simulations = 70000 5th Percentile 1.0576E-11 Median 3. 1534E-08 95.0th Percentile 1.8255E-06 99.0th Percentile 5.6968E-06 99.9th Percentile 3.8943E-05 Mean 4.7962E-07 Standard Deviation 3.4846E-06 Standard Error 1.3171E-08 Variance (unbiased) 1.2142E-11 Variance (biased) 1.2142E-11 Moment Coeff. of Skewness 4. 1589E+01 Pearson's 2nd Coeff. of Skewness 4.1292E-01 Kurtosis 3.0077E+03

• PROBABILITY DISTRIBUTION FUNCTION (HISTOGRAM) *
• FOR THROUGH-WALL CRACKING FREQUENCY (FAILURE)
• WCAP- 16168-NP-A June 2008 Revision 2

1-6 FREQUENCY OF RELATIVE CUMULATIVE TWC FAILURES DENSITY DISTRIBUTION (PER REACTOR-OPERATING YEAR) (%) (%) 0.OOOOE+00 2.7414 2.7414 9.9132E-07 96.6714 99.4129 2.9740E-06 0.3271 99.7400 4.9566E-06 0.0971 99.8371 6.9392E-06 0.0557 99.8929 8.9219E-06 0.0286 99.9214 1.0904E-05 0.0186 99.9400 1.2887E-05 0.0100 99.9500 1.4870E-05 0.0100 99.9600 1.6852E-05 0.0071 99.9671 1.8835E-05 0.0071 99.9743 2.0818E-05 0.0029 99.9771 2.2800E-05 0.0057 99.9829 2.4783E-05 0.0029 99.9857 2.8748E-05 0.0014 99.9871 3.2713E-05 0.0014 99.9886 3.4696E-05 0.0014 99.9900 3.6679E-05 0.0014 99.9914 4.2627E-05 0.0014 99.9929 4.4609E-05 0.0014 99.9943 4.6592E-05 0.0029 99.9971 1.0409E-04 0.0014 99.9986 1.9727E-04 0.0014 100.0000

                         ==       Summary Descriptive     Statistics             ==

Minimum 0.OOOOE+00 Maximum 1.9628E-04 Range 1.9628E-04 Number of Simulations = 70000 5th Percentile 1.2334E-13 Median 2.0809E-09 95.0th Percentile 9. 9132E-07 99.0th Percentile 1. 1299E-06 99.9th Percentile 7.4349E-06 Mean 7.6237E-08 Standard Deviation 1. 0800E-06 Standard Error 4. 0821E-09 Variance (unbiased) 1. 1664E-12 Variance (biased) 1. 1664E-12 Moment Coeff. of Skewness 1. 0725E+02 Pearson's 2nd Coeff. of Skewness 1 .2417E-01 Kurtosis 1. 7088E+04 WCAP- 16168-NP-A June 2008 Revision 2

1-7 1-1: 10 Year ISI only (cont.)

• FRACTIONALIZATION OF FREQUENCY OF CRACK INITIATION *
• AND THROUGH-WALL CRACKING FREQUENCY (FAILURE) *
• WEIGHTED BY TRANSIENT INITIATING FREQUENCIES *

                                                    % of total               % of total frequency of               frequency of crack initiation             of TWC failure 2                  0.00                          0.00 16                   0.00                          0.00 18                    0.00                         0.00 19                    0.28                         1.10 22                    0.00                         0.00 24                    0.00                         0.00 26                    0.10                         0.08 27                    0. 16                        0.30 29                    0.00                         0.00 31                    0.07                         0.10 32                    0.00                         0.00 34                    0.01                         0.01 40                 50.98                         20.66 42                    0.00                         0.00 48                    0.09                         0.58 49                    0.00                         0.00 50                    0.01                         0.01 51                    0.01                         0.02 52                    0.05                         0.21 53                    0. 06                        0.26 54                    0.57                         1.62 55                    1 .02                        4.91 58                24 .02                        27.50 59                   0  .79                       0.36 60                   1.91                         1.55 61                   0.05                         0.01 62                   8.97                         5.93 63                   2.59                         2.33 64                   3.42                         3.01 65                   4.84                       29.44 TOTALS      100.00                       100.00 WCAP- 16168-NP-A                                                                             June 2008 Revision 2

1-8 I-1: 10 Year ISI only (cont.)

• FRACTIONALIZATION OF FREQUENCY OF CRACK INITIATION *
• AND THROUGH-WALL CRACKING FREQUENCY (FAILURE) -
• BY *
• RPV BELTLINE MAJOR REGION *
• BY PARENT SUBREGION *
• WEIGHTED BY % CONTRIBUTION OF EACH TRANSIENT *
• TO FREQUENCY OF CRACK INITIATION AND *
• THROUGH-WALL CRACKING FREQUENCY (FAILURE) *

                                                                                 % of total
                                 % of           % of total                 through-wall crack MAJOR          RTndt       total         frequency of                        frequency REGION         (MAX)       flaws      crack initiation          cleavage ductile total 1        237.44        2 .04               7.75                3.95           2.55            6.51 2        246.96        2 .04              10.50                6.85           4.40         11.24 3        246.96        2 .04              11 34                7.15           4.89         12.04 4        247.43        3 .16              23 .77              19.46         10.46          29.92 5        237.77        3.16               16 .33              10.62           5.75         16.36 6        247.43        3.16               22 .11              14.52           9.40         23 . 92 7        231.49       19.12                7.68                0.01           0.00            0.01 8        195.14        8.55                0.04                0.00           0.00            0.00 9       166.15        8.55                0.00                0.00           0.00             0.00 10        130.00        8.55                0.00                 0.00          0.00             0.00 11        206.88       13.21                0.18                 0.00          0.00             0.00 12        186.36       13.21                 0.03                0.00          0.00             0.00 13        208.62       13.21                 0.26                0.00          0.00             0.00 TOTALS 100.00                   100.00               62.56          37.44       100.00

• FRACTIONALIZATION OF FREQUENCY OF CRACK INITIATION *
• AND THROUGH-WALL CRACKING FREQUENCY (FAILURE) - *
• BY *
• RPV BELTLINE MAJOR REGION *
• BY CHILD SUBREGION *
• WEIGHTED BY % CONTRIBUTION OF EACH TRANSIENT *
• TO FREQUENCY OF CRACK INITIATION AND *
• THROUGH-WALL CRACKING FREQUENCY (FAILURE)
• June 2008 WCAP- 16168-NP-A WCAP-16168-NP-A June 2008 Revision 2

1-9 I-1: 10 Year ISI only (cont.)

                                                                           % of total
                               % of          % of total              through-wall crack MAJOR         RTndt       total       frequency of                     frequency REGION        (MAX)      flaws    crack initiation         cleavage ductile total 1       237.44        2.04              7.75               3.95         2.55        6.51 2       246.96        2.04            10.50                6.84         4.40       11.24 3       246.96        2.04            11.34                7.15         4.88       12.04 4       247.43        3.16            23.77              19.46        10.46        29.92 5       237.77        3.16            16.33              10.62          5.75       16.36 6       247.43        3.16            22.11              14.52          9.40       23.92 7       231.49       19.12              7.32               0.01         0.00        0.01 8       195.14        8.55              0.04               0.00         0.00        0.00 9      166    .15    8 .55             0.00               0.00         0.00        0.00 10       130.00        8.55              0.00               0.00         0.00        0.00 11       206.88       13 .21             0.18               0.00         0.00        0.00 12       186.36       13 .21             0.03               0.00         0.00        0.00 13       208.62       13.21              0.61               0.00         0.00        0.00 TOTALS 100.00                100.00               62.56        37.44     100.00

• FRACTIONALIZATION OF FREQUENCY OF CRACK INITIATION *
• AND THROUGH-WALL CRACKING FREQUENCY (FAILURE) - *
• MATERIAL, FLAW CATEGORY, AND FLAW DEPTH *
• WEIGHTED BY % CONTRIBUTION OF EACH TRANSIENT *
• TO FREQUENCY OF CRACK INITIATION AND *
• THROUGH-WALL CRACKING FREQUENCY (FAILURE) *

                                                     **'**  ****A******

• WELD MATERIAL *

                                 % of total     frequency              of total      through-wall of crack initiation                    crack frequency FLAW DEPTH          CAT I       CAT 2        CAT 3      CAT 1        CAT 2    CAT 3 (in)         flaws       flaws        flaws      flaws         flaws    flaws

0. 088 0.00 3.48 0.00 0.00 1.59 0.00 0 .175 0.00 36.00 0.00 0.00 19.98 0.00 0.263 0.00 13.27 0.00 0.00 10.08 0.00 0.350 0.00 8.79 0.00 0.00 9.08 0.01 0.438 0.00 7.99 0.01 0.00 9.15 0.03

0. 525 0.00 6.37 0. 01 0.00 9.36 0.06 WCAP-16168-NP-A June 2008 Revision 2

1-10 1-1: 10 Year ISI only (cont.) 0.613 0.00 4.47 0.01 0.00 5.81 0.05 0.700 0.00 3.48 0.01 0.00 5.76 0 .08 0.787 0.00 2.88 0.02 0.00 4.99 0.10 0.875 0. 00 2.33 0.01 0.00 3 .81 0.05

0. 963 0.00 1.55 0.03 0.00 2.79 0.18 1.050 0.06 2.27 0.04 0.00 6.36 0.20 1.137 1.00 1.78 0.02 0. 03 3.20 0.13 1.225 0.02 1.27 0 .01 0.00 2.32 0.09 1.313 0.09 0.76 0.00 0 .0~0 1.53 0.02 1.400 0.00 0.35 0.01 0.00 0.63 0.03 1.488 0. 00 0.19 0.00 0.00 0.37 0.02 1.575 0.00 0.29 0.00 0.00 0.49 0.00 1.663 0.00 0.03 0.00 0.00 0.10 0.01 1.750 0.00 0.26 0 .01 0.00 0.72 0.05 1.837 0.03 0.00 0.00 0. 00 0.03 0.00 1.925 0. 00 0.28 0.00 0. 00 0.68 0.02 2.013 0.00 0.00 0.00 0.00 0.00 0.00 2.100 0.00 0.00 0.00 0.00 0.00 0.00 2.188 0.00 0.00 0.00 0.00 0.00 0.00 2 .275 0.00 0.00 0.00 0.00 0.00 0.00 2 .363 0.00 0.00 0.00 0.00 0.00 0.00 2.450 0.00 0.00 0.00 0.00 0. 00 0.00 2.537 0.00 0.00 0.00 0.00 0.00 0.00 2 .625 0.00 0.00 0.00 0.00 0.00 0.00 2 .712 0.00 0.00 0.00 0.00 0.00 0.00 TOTALS 1.20 98.08 0.19 0.03 98.83 1.13

                                             *PLATE       MATERIAL*
                               %of total frequency               of total through-wall of crack initiation                    crack frequency FLAW DEPTH        CAT I    CAT 2         CAT 3      CAT1I       CAT 2  CAT 3 (in)        flaws    flaws         flaws      flaws       flaws  flaws 0.088         0.00     0.00          0.00       0.00        0.00      0.00 0.175         0.00     0.00          0.00       0.00        0.00      0.00 0.263         0.00     0.00          0.00       0:00        0.00      0.00 0.350         0.00     0.00          0.00       0.00        0.00      0.00 0.438         0.00     0.00          0.00       0.00        0.00      0.00 0.525         0.00     0.00          0.00       0.00        0.00      0.00 0.613         0.00     0.00          0.00       0.00        0.00      0.00 0.700         0.00     0. 00         0.00       0.00        0.00      0.00 0.787         0.00     0.00          0.00       0.00        0.00      0.00 0.875         0.00     0.00          0.00       0.00        0.00      0.00 0 .963        0.00     0.00          0.00       0.00        0.00      0.00

1. 050 0.07 0.00 0.00 0.00 0.00 0.00 1.137 0.28 0.00 0.00 0.00 0.00 0.00 WCAP-16168-NP-A June 2008 Revision. 2

1-11 I-1: 10 Year ISI only (cont.) 1.225 0.10 0.00 0.00 0.00 0.00 0.00 1.313 0 .02 0.00 0.00 0.00 0.00 0.00 1.400 0 .04 0.00 0.00 0.00 0.00 0 .00 1.488 0.00 0.00 0.00 0.00 0.00 0.00 1.575 0.00 0.00 0.00 0.00 0.00 0.00 1.663 0.00 0.00 0.00 0.00 0.00 0.00 1.750 0.01 0.00 0.00 0.00 0.00 0.00 1.837 0.00 0.00 0.00 0.00 0.00 0.00

1. 925 0.00 0.00 0.00 0.00 0.00 0.00 2.013 0.00 0.00 0.00 0.00 0.00 0.00 2.100 0.00 0.00 0.00 0.00 0.00 0.00 2.188 0.00 0.00 0.00 0.00 0.00 0.00 2.275 0.00 0.00 0.00 0.00 0.00 0.00 2.363 0.00 0.00 0.00 0.00 0.00 0.00 2.450 0.00 0.00 0.00 0.00 0.00 0.00 2.537 0.00 0.00 0.00 0 .00 0.00 0.00 2.625 0.00 0.00 0.00 0.00 0.00 0.00 2 .712 0.00 0.00 0.00 0.00 0.00 0.00 TOTALS 0.52 0.00 0.00 0.00 0.00 0.00 DATE: 27-Jun-2007 TIME: 10:25:45 WCAP-16168-NP-A June 2008 Revision 2

1-12 1-2: ISI Every 10 Years

                      *****WW*    *******************            **** * *******

• WELCOME TO FAVOR *
• FRACTURE ANALYSIS OF VESSELS: OAK RIDGE *
• VERSION 06.1 *
• FAVPOST MODULE: POSTPROCESSOR MODULE *
• COMBINES TRANSIENT INITIAITING FREQUENCIES *
• WITH RESULTS OF PFM ANALYSIS *
• PROBLEMS OR QUESTIONS REGARDING FAVOR *
• SHOULD BE DIRECTED TO *
• TERRY DICKSON *
• OAK RIDGE NATIONAL LABORATORY *
• e-mail: dicksontl@ornl.gov *
• This computer program was prepared as an account of *
• work sponsored by the United States Government *
• Neither the United States, nor the United States *
• Department of Energy, nor the United States Nuclear *
• Regulatory Commission, nor any of their employees, *
• nor any of their contractors, subcontractors, or their *
• employees, makes any warranty, expressed or implied, or *
• assumes any legal liability or responsibility for the *
• accuracy, completeness, or usefulness of any *
• information, apparatus, product, or process disclosed, *
• or represents that its use would not infringe *
• privately-owned rights.
• DATE: 27-Jun-2007 TIME: 10:23:32 Begin echo of FAVPost input data deck 10:23:32 27-Jun-2007 End echo of FAVPost input data deck 10:23:32 27-Jun-2007 FAVPOST INPUT FILE NAME = postpl.in FAVPFM OUTPUT FILE CONTAINING PFMI ARRAY = INITIATE.DAT FAVPFM OUTPUT FILE CONTAINING PFMF ARRAY = FAILURE.DAT FAVPOST OUTPUT FILE NAME = plpostlOyrint.out WCAP- 16168-NP-A

1-13 1-2: ISI Every 10 Years (cont.)

• NUMBER OF SIMULATIONS = 70000
• CONDITIONAL PROBABILITY CONDITIONAL PROBABILITY OF INITIATION CPI=P(IIE) OF FAILURE CPF=P(FIE)

TRANSIENT MEAN 95th % 99th % MEAN 95th % 99th % RATIO NUMBER CPI CPI CPI CPF CPF CPF CPFmn/CPImn

            ---------                      0. 0000E+00                      0.0000OE+00   0 OOOOE+00    0.0000 2     0.OOOOE+00         0.OOOOE+00  0 OOOOE+00       0.OOOOE+00      0.OOOOE+00 16      1.2905E-10         0.OOOOE+00                   1.9672E-11                    0 OOOOE+00    0. 1524 0 OOOOE+00                       0.OOOOE+00    0 0000OE+00 18      1.9999E-08         0.OOOOE+00                   1.3349E-08                                  0.6675 0 OOOOE+00                       0.OOOOE+00    0 OOOOE+00 19      4.3542E-07         0.OOOOE+00                   2.9306E-07                    0 OOOOE+00    0.6730 0 OOOOE+00                       0.OOOOE+00 22      4.7785E-10         0.0000E+00                   1.6163E-10                    0 OOOOE+00    0.3382 0 OOOOE+00                       0.OOOOE+00 24      4.2145E-07         0.OOOOE+00                   5.5149E-08                                  0.1309 0 .OOOOE+00                      0.OOOOE+00 26      4.2145E-07         0.OOOOE+00                   5.6127E-08                    0 OOOOE+00    0 .1332 27      1.6689E-05         4. 0851E-04 1.8246E-04       5.0727E-06      1.3999E-04    5.5346E-05    0.3040 29      3.2222E-07         0 OOOOE+00  0.OOOOE+00       2 .2848E-07     0.OOOOE+00    0.OOOOE+00    0 .7091 31      1.9968E-05         3. 1503E-04 2.2641E-04       5.2229E-06      8.4354E-05    5.6231E-05    0.2616 32      4.1319E-07         0 .OOOOE+00 0.OOOOE+00       2. 7634E-07     0.OOOOE+00    0.OOOOE+00    0.6688 34      2.0638E-06         0 OOOOE+00  9.6818E-07       3. 9138E-07     0.OOOOE+00    1.4814E-07    0.1896 40      5.8654E-03         1. 3025E-02 6.3577E-02       3. 8441E-04     8.2240E-04    4.9153E-03    0.0655 0.OOOOE+00       0. 0000E+00     0.OOOOE+00    0.OOOOE+00    0.0000 42      0.OOOOE+00         0 .OOOOE+00 48      5.1460E-04         2. 9847E-03 7. 2004E-03      5. 0771E-04     2.9613E-03    7. 1338E-03   0.9866 49      1.2025E-07         0 OOOOE+00  0 OOOOE+00       2.3074E-08      0.OOOOE+00    0 OOOOE+00    0.1919 50      3.8097E-05         6. 1000E-04 6. 0949E-04      1.3058E-05      2.3488E-04    2. 0982E-04   0.3428 51      2.1960E-04         1.5122E-03  3. 0705E-03      1.1283E-04      6.4702E-04    1. 6553E-03   0.5138 52      5.4043E-07         0.OOOOE+00  0 OOOOE+00       3.6486E-07      0.OOOOE+00    0 OOOOE+00    0.6751 53      2.6394E-07         0.OOOOE+00  0 OOOOE+00       1.9008E-07      0.OOOOE+00    0 OOOOE+00    0.7202 54      4.4223E-04         2.0292E-03  6. 0638E-03      2.0682E-04      9.2843E-04    2. 9785E-03   0.4677 55      1.1909E-06         0.OOOOE+00  0 OOOOE+00       8.8283E-07      0.OOOOE+00    0.000OE+00    0.7413 58      3.4658E-04         1.0963E-03  5. 2758E-03      7.1215E-05      3.8156E-04    9 .2302E-04   0.2055 59      1.3685E-05         2. 8125E-04 1. 3312E-04      1.1435E-06      2.6929E-05    8. 8218E-06   0.0836 WCAP- 16168-NP-A                                                                                                  June 2008 Revision 2

1-14 1-2: ISI Every 10 Years (cont.) 60 3 .4196E-05 5. 1697E-04 4.9575E-04 5.1734E-06 1. 0188E-04 6.8749E-05 0.1513 61 6. 3437E-07 0.OOOOE+00 2 .4648E-11 3 . 7831E-08 0.OOOOE+00 2.4484E-14 0.0596 62 4. 6580E-03 1.0090E-02 5 .1261E-02 5.0659E-04 1.0724E-03 6.4912E-03 0.1088 63 1. 5523E-03 3.1172E-03 2. 1441E-02 2.2840E-04 6. 1778E-04 3. 0185E-03 0.1471 64 1. 8747E-03 4.0838E-03 2 .2910E-02 2.8006E-04 6.6854E-04 3 . 7212E-03 0.1494 65 1. 6533E-04 1. 9541E-03 2.8139E-03 1.6078E-04 1.9452E-03 2.7094E-03 0.9725 NOTES: CPI IS CONDITIONAL PROBABILITY OF CRACK INITIATION, P(IIE) CPF IS CONDITIONAL PROBABILITY OF TWC FAILURE, P(FIE) WCAP-16168-NP-A June 2008 Revision 2

1-15 1-2: ISI Every 10 Years (cont.)

• PROBABILITY DISTRIBUTION FUNCTION (HISTOGRAM) *
• FOR THE FREQUENCY OF CRACK INITIATION
• FREQUENCY OF RELATIVE CUMULATIVE CRACK INITIATION DENSITY -DISTRIBUTION (PER REACTOR-OPERATING YEAR) (%) (%)

0.OOOOE+00 1.8657 1.8657 1.3669E-06 95.1129 96.9786 4.1008E-06 1.6643 98.6429 6.8346E-06 0.5400 99.1829 9.5685E-06 0.2614 99.4443 1.2302E-05 0.1471 99.5914 1.5036E-05 0.0986 99.6900 1.7770E-05 0.0771 99.7671 2.0504E-05 0.0329 99.8000 2.3238E-05 0.0300 99.8300 2.5972E-05 0.0186 99.8486 2.8705E-05 0.0229 99.8714 3.1439E-05 0.0157 99.8871 3.4173E-05 0.0186 99.9057 3.6907E-05 .0.0071 99.9129 3.9641E-05 0.0043 99.9171 4.2375E-05 0.0029 99.9200 4.5109E-05 0.0057 99.9257 4.7842E-05 0.0157 99.9414 5.0576E-05 0.0086 99.9500 5.3310E-05 0.0029 99.9529 5.6044E-05 0.0029 99.9557 6.1512E-05 0.0029 99.9586 6.4246E-05 0.0071 99.9657 6.6979E-05 0.0043 99.9700 6.9713E-05 0.0043 99.9743 7.2447E-05 0.0029 99.9771 7.5181E-05 .0.0014 99.9786 8.0649E-05 0.0014 99.9800 8.3383E-05 0.0014 99.9814 8.8850E-05 0.0043 99.9857 9.4318E-05 0.0029 99.9886 9.7052E-05 0.0029 99.9914 9.9786E-05 0.0014 99.9929 1.1072E-04 0.0014 99.9943 1.2166E-04 0.0014 99.9957 1.7907E-04 0.0014 99.9971 2.4195E-04 0.0014 99.9986 2.6928E-04 0.0014 100.0000 WCAP- 16168-NP-A June 2008 Revision 2

1-16 1-2: ISI Every 10 Years (cont.)

                             ==        Summary Descriptive Statistics Minimum                                   =   0.0000E+00 Maximum                                   =   2.7065E-04 Range                                     =   2.7065E-04 Number of Simulations                     =    70000 5th Percentile                            =   9.2117E-12 Median                                    =   3.0552E-08 95.0th Percentile                         =   1.3669E-06 99.0th Percentile                         =   5.9089E-06 99.9th Percentile                         =   3.3332E-05 Mean                                      =   4.4736E-07 Standard Deviation                        =   2.9010E-06 Standard Error                            =   1.0965E-08 Variance (unbiased)                       =   8.4156E-12 Variance (biased)                         =   8.4155E-12 Moment Coeff. of Skewness                 =   3.6027E+01 Pearson's 2nd Coeff. of Skewness          =   4.6263E-01 Kurtosis                                  =   2.3099E+03

• PROBABILITY DISTRIBUTION FUNCTION (HISTOGRAM) *
• FOR THROUGH-WALL CRACKING FREQUENCY (FAILURE)
• FREQUENCY OF RELATIVE CUMULATIVE TWC FAILURES DENSITY DISTRIBUTION (PER REACTOR-OPERATING YEAR) (-%.) (%)

0 OOOOE+00 2.7414 2.7414

7. 0253E-07 96.4900 99.2314 2 1076E-06 0.3986 99.6300

3. 5126E-06 0.1529 99.7829

4. 9177E-06 0.0600 99.8429

6. 3228E-06 0.0414 99.8843 7.7278E-06 0.0286 99.9129 9.1329E-06 0.0257 99.9386 1.0538E-05 0.0043 99.9429 1.1943E-05 0.0086 99.9514 1.3348E-05 0.0114 99.9629 1.4753E-05 0.0014 99.9643 1.6158E-05 0.0043 99.9686 1.7563E-05 0.0071 99.9757 2.4589E-05 0.0029 99.9786 WCAP- 16168-NP-A June 2008 Revision 2

1-17 1-2: ISI Every 10 Years (cont.) 2.7399E-05 0.0029 99.9814 3.0209E-05 0. 0014 99.9829 3 . 1614E-05 0.0029 99.9857 3.4424E-05 0.0014 99.9871

3. 5829E-05 0.0029 99.9900 3 .7234E-05 0. 0014 99.9914

3. 8639E-05 0 .0014 99.9929

4. 0044E-05 0 .0014 99.9943 4 .4259E-05 0. 0014 99.9957 6.9550E-05 0.0014 99.9971 1 .3278E-04 0.0014 99.9986

1. 3980E-04 0.0014 100.0000

                              ==        Summary Descriptive Statistics            ==

Minimum = 0.OOOOE+00 Maximum = 1.3910E-04 Range = 1.3910E-04 Number of Simulations = 70000 5th Percentile = 1.2899E-13 Median = 2.0727E-09 95.0th Percentile = 7.0253E-07 99.0th Percentile = 1.0895E-06 99.9th Percentile = 7.0955E-06 Mean = 7.3880E-08 Standard Deviation = 1.0054E-06 Standard Error = 3.8001E-09 Variance (unbiased) = 1.0108E-12 Variance (biased) = 1.0108E-12 Moment Coeff. of Skewness = 8.5112E+01 Pearson's 2nd Coeff. of Skewness = 1.2928E-01 Kurtosis = 1.0251E+04

• FRACTIONALIZATION OF FREQUENCY OF CRACK INITIATION *
• AND THROUGH-WALL CRACKING FREQUENCY (FAILURE) *
• WEIGHTED BY TRANSIENT INITIATING FREQUENCIES *

                                             % of total               % of total frequency of            frequency of crack initiation         of TWC failure 2             0.00                       0.00 16              0.00                       0.00 18             0.00                       0.01 19             0.16                       0.66 WCAP- 16168-NP-A                                                                    June 2008 Revision 2

1-18 1-2: ISI Every 10 Years (cont.) 22 0.00 0.00 24 0.00 0.00 26 0.05 0.04 27 0.18 0.34 29 0.00 0.00 31 0 .06 0.10 32 0.00 0.00 34 0.01 0.01 40 52.00 21.41 42 0.00 0.00 48 0.10 0.57 49 0.00 0.00 50 0.01 0.01 51 0 .01 0.02 52 0.09 0.36 53 0.08 0.37 54 0 .57 1.63 55 0.66 3'.11 58 22.68 28.39 59 0.69 0.33 60 1.80 1.68 61 0.03 0 .01 62 9.31 6 .13 63 2.87 2.66 64 3.76 3.33 65 4.90 28.85 TOTALS 100.00 100.00

• FRACTIONALIZATION OF FREQUENCY OF CRACK INITIATION *
• AND THROUGH-WALL CRACKING FREQUENCY (FAILURE) -
• BY *
• RPV BELTLINE MAJOR REGION
• BY PARENT SUBREGION *
• WEIGHTED BY % CONTRIBUTION OF EACH TRANSIENT *
• TO FREQUENCY OF CRACK INITIATION AND *
• THROUGH-WALL CRACKING FREQUENCY (FAILURE) *

                                                                      % of total
                            % of         % of total              through-wall crack MAJOR      RTndt      total       frequency of                   frequency REGION      (MAX)     flaws     crack initiation       cleavage ductile total 1     237.44      2.04             8.49               6.27       3.00        9.27 2     246.96      2.04            11.97               8.50       5.03     13.53 3     246.96      2.04            11.45               7.14       4.59     11.73 4     247.43      3.16            21.86              12.94      10.00     22.94 June 2008 WCAP- 16168-NP-A WCAP-16168-NP-A                                                                            June 2008 Revision 2

1-19 1-2: ISI Every 10 Years (cont.) 5 237.77 3.16 15.63 8.90 5.70 14.60 6 247.43 3.16 23.36 17.70 10.23 27.93 7 231.49 19.12 7.26 0.00 0.00 0.00 8 195.14 8 .55 0.00 0.00 0.00 0.00 9 166.15 8.55 0.00 0.00 0.00 0.00 10 130.00 8.55 0.00 0.00 0.00 0.00 11 206.88 13 .21 0.00 0.00 0.00 0.00 12 186.36 13 .21 0.00 0.00 0.00 0.00 13 208.62 13.21 0.00 0.00 0.00 0.00 TOTALS 100.00 100.00 61.45 38.55 100.00

• FRACTIONALIZATION OF FREQUENCY OF CRACK INITIATION *
• AND THROUGH-WALL CRACKING FREQUENCY (FAILURE) - *
• BY
• RPV BELTLINE MAJOR REGION
• BY CHILD SUBREGION
• WEIGHTED BY % CONTRIBUTION OF EACH TRANSIENT *
• TO FREQUENCY OF CRACK INITIATION AND *
• THROUGH-WALL CRACKING FREQUENCY (FAILURE) *

                                                                 % of total
                             % of        %*of total          through-wall crack MAJOR       RTndt      total      frequency of              frequency REGION       (MAX)     flaws    crack initiation   cleavage ductile total 1    237.44       2 .04           8.49           6.27      3.00      9.27 2    246.96       2 .04          11.97           8.50      5.03    13.53 3    246.96       2 .04          11.45           7.14      4.59    11.73 4    247.43       3 .16          21.86         12 .94     10.00    22.94 5    237.77       3 .16          15.63           8.90      5.70    14.60 6    247.43       3.16           23.36         17.70      10.23    27.93 7    231.49     19.12             6.96           0.00       0.00      0.00 8     195.14      8.55            0.00           0.00       0.00      0.00 9     166.15      8.55             0.00          0.00       0.00      0.00 10      130.00      8.55             0.00          0.00       0.00      0.00 11      206.88     13.21             0.00          0.00       0.00      0.00 12      186.36     13.21             0.00          0.00       0.00      0.00 13      208.62     13.21             0.29          0.00       0.00      0.00 TOTALS 100.00             100.00          61.45     38.55   100.00 WCAP- 16168-NP-A                                                                     June 2008 Revision 2.

1-20 1-2: ISI Every 10 Years (cont.)

• FRACTIONALIZATION OF FREQUENCY OF CRACK INITIATION *
• AND THROUGH-WALL CRACKING FREQUENCY (FAILURE) - *
• MATERIAL, FLAW CATEGORY, AND FLAW DEPTH *
• WEIGHTED BY % CONTRIBUTION OF EACH TRANSIENT *
• TO FREQUENCY OF CRACK INITIATION AND *
• THROUGH-WALL CRACKING FREQUENCY (FAILURE) *

                                        ** ************L*****************              *****

• WELD MATERIAL *

                             % of total       frequency          % of total      through-wall of crack initiation                     crack frequency FLAW DEPTH     CAT I         CAT 2       CAT 3       CAT 1        CAT 2    CAT 3 (in)      flaws         flaws       flaws       flaws        flaws    flaws 0.088       0. 00         3.70       0.00,        0.00         1.68       0.00 0.175       0.00        37.45        0.00         0.00       20.91        0.00 0.263       0.00        13.39        0.00         0.00       10.25        0.00 0.350       0.00        10.03        0.00         0.00         9.61       0.01 0.438       0.00          7.97        0.01        0.00         9.27       0.05 0.525       0.00          6.65        0.01        0.00       10.34        0.06 0.613       0.00          5 .13       0.01        0.00         7.94       0.06 0.700       0.00          3 .58       0.03        0.00         5.22       0.14 0.787       0.00          2.47        0.02        0.00         3.86       0.13 0.875       0.00          1.59        0.00        0.00         1.90       0. 02 0.963       0.00          2.29        0.01        0.00         4.94       0.05 1.050       0.00          0. 96       0 .02       0.00         1.60       0.10 1.137       0.00          1.31        0 .01       0.00         2.78       0.03 1.225       0.00          0.78        0.00        0.00         1.67       0.01 1.313        0. 00        0 .34       0. 01       0.00         0.43       0. 04 1.400        0.00         0.54        0.00        0.00         1.31       0. 03 1.488        0.00         0.40        0.00        0.00         1.14       0.01 1.575        0.00         0.05        0.00        0.00         0.13       0.01 1.663        0.00         0.84        0 .01       0.00         3.14       0. 04 1.750        0.00         0.13        0.00        0.00         0.42       0.00 1.837        0. 00        0.17        0.00        0.00         0-.60      0.00 1.925        0.00         0.06        0.00        0.00         0.04       0.02 TOTALS        0.00       99.86         0.14        0.00       99.17        0.83 WCAP- 16168-NP-A                                                                                June 2008 Revision 2

1-21 1-2: 1SI Every 10 Years (cont.)

• PLATE MATERIAL *

                             % of total     frequency          % of total    through-wall of crack initiation                   crack frequency FLAW DEPTH      CAT I        CAT 2      CAT 3       CAT 1      CAT 2   CAT 3 (in)       flaws        flaws      flaws       flaws      flaws   flaws 0.088        0.00        0.00       0.00         0.00       0.00      0.00 0.175        0.00         0.00       0.00        0.00       0.00      0.00 0.263        0.00         0.00       0:.00       0.00       0.00      0.00 0.350        0.00         0.00       0.00        0.00       0.00      0.00 0.438        0.00         0.00       0.00        0.00       0.00      0.00 0.525        0.00         0.00       0.00        0.00       0.00      0.00 0.613        0.00         0.00       0.00        0.00       0.00      0.00 0.700        0.00         0.00       0.00        0.00       0.00      0.00 0.787        0.00         0.00       0.00        0.00       0.00      0.00 0.875        0.00         0.00       0.00        0.00       0.00      0.00 0.963        0.00         0.00       0.00        0.00       0.00      0.00 1.050        0.00         0.00       0.00        0.00       0.00      0.00 1.137        0.00         0.00       0.00        0.00       0.00      0.00 TOTALS         0.00         0.00       0.00        0.00       0.00      0.00 DATE:    27-Jun-2007      TIME:   10:24:06 WCAP- 16168-NP-A                                                                         June 2008 Revision 2

J-1 APPENDIX J INPUTS FOR THE OCONEE UNIT 1 PILOT PLANT EVALUATION WCAP- 16168-NP-A June 2008 Revision 2

J-2 A summary of the NDE inspection history based on Regulatory Guide 1.150 and pertinent input data for OC1 is as follows:

1. Number of inservice inspections perfonrmed (relative to initial pre-service and 10 year interval inspections) for full penetration category B-A and B-D vessel welds assuming all of the candidate welds were inspected: 3 (covering all welds of the specified categories).

2. The inspections performed covered: 62 total examinations. 23 items with 100% coverage, 22 items with < 90% coverage and 17 items with coverage >90% but less than 100%.

3. Number of indications found during most recent inservice inspection: 44 This number includes consideration of the following additional information.

a. Indications found that were reportable: 0

b. Indications found that were within acceptable limits: 44

c. Indications/anomalies currently being monitored: 0

4. Full Penetration Relief requests for the reactor vessel submitted and accepted by the NRC: 2 relief requests for limited coverage for 22 items, as noted in item 2

5. Fluence distribution at inside surface of RV Beltline until end of life is shown in: see Figure J-1 taken from the NRC PTS Risk Study [44], Figure 4.1.

June 2008 WCAP- 16168-NP-A WCAP-16168-NP-A June 2008 Revision 2

J-3 7::= , i= -*:- ' Figure J-1 Rollout Diagram of Beltline Materials and Representative Fluence Maps for OCI

6. Reactor vessel cladding details:

a. Number of layers: 1

b. Thickness: 0.188

c. Material properties are identified in Table J-l:

WCAP- 16168-NP-A June 2008 Revision 2

J-4 Table J-1 Cladding Material Properties Specific Young's Thermal Thermal Heat Modulus of Expansion Conductivity (Btu/LBM- Elasticity Coefficient (OF-I) Density Poisson's Temperature (Btu/hr-ft- 0F) OF) (KSI) (LBM/fl3) Ratio , (OF) "K" "C"C "E" "." .p v 0 - 489 .3 68 22045.7 489 .3 70 8.1 0.1158 - 489 .3 100 8.4 0.1185 - 8.55E-06 489 .3 150 8.6 0.1196 - 8.67E-06 489 .3 200 8.8 0.1208 - 8.79E-06 489 .3 250 9.1 0.1232 - 8.9E-06 489 .3 300 9.4 0.1256 - 9.OE-06 489 .3 302 - - 20160.2 489 .3 350 9.6 0.1258 - 9.1E-06 489 .3 400 9.9 0.1281 - 9.19E-06 489 .3 450 10.1 0.1291 - 9.28E-06 489 .3 482 - - 18419.8 489 .3 500 10.4 0.1305 - 9.37E-06 489 .3 550 10.6 0.1306 - 9.45E-06 489 .3 600 10.9 0.1327 - 9.53E-06 489 .3 650 11.1 0.1335 - 9.61E-06 489 .3 700 11.4 0.1348 - 9.69E-06 489 .3 750 11.6 0.1356 - 9.76E-06 489 .3 800 11.9 0.1367 - 9.82E-06 489 .3

d. Material including copper and nickel content: Material properties assigned to clad flaws are that of the underlying material be it base metal or weld. These properties are identified in Table J-3. This is consistent with the PTS evaluation [8, 9].

e. Material property uncertainties:

1) Bead width: 1 inch - bead widths vary for all plants. Based on the NRC PTS Risk Study [8, 9], a nominal dimension of 1 inch is selected for all analyses because this parameter is not expected to significantly influence the predicted vessel failure probabilities.

2) Truncation Limit: Cladding thickness rounded up to the next 1 / 1 0 0 th of the total vessel thickness to be consistent with the NRC PTS Risk Study [8, 9].

3) Surface flaw depth: 0.03 x 8.626 = 0.259 in

4) All flaws are surface-breaking. Only flaws in cladding that would influence brittle fracture of the reactor vessel are brittle. This is consistent with the NRC PTS Risk Study [8, 9].

f. Additional cladding properties are identified in Table J-4

7. Base metal:

WCAP- 16168-NP-A June 2008 Revision 2

J-5

a. Wall thickness: 8.438 inches

b. Material properties are identified in Tables J-2 and J-3:

Table J-2 Base Metal Material Properties Specific Young's Thermal Thermal Heat Modulus of Expansion Conductivity (Btu/LBM- Elasticity Coefficient (Btu/hr-ft-°F) OF) (KSI) (OF-I) Density Poisson's 3 Temperature (LBM/ft ) Ratio ( OF) " K ..." C"..E ..." ... P 'I , I.. 0 - - 489 .3 70 24.8 0.1052 29200 489 .3 100 25 0.1072 - 7.06E-06 489 .3 150 25.1 0.1101 - 7.16E-06 489 .3 200 25.2 0.1135 28500 7.25E-06 489 .3 250 25.2 0.1166 - 7.34E-06 489 .3 300 25.1 0.1194 28000 7.43E-06 489 .3 350 25 0.1223 - 7.5E-06, 489 .3 400 25.1 0.1267 27400 7.58E-06 489 .3 450 24.6 0.1277 - 7.63E-06 489 -. 3 500 24.3 0.1304 27000 7.7E-06 489 .3 550 24 0.1326 - 7.77E-06 489 .3 600 23.7 0.135 26400 7.83E-06 489 A 650 23.4 0.1375 - 7.9E-06 489 .3 700 23 0.1404 25300 7.94E-06 489 .3 750 22.6 0.1435 - 8.OE-06 489 .3 800 22.2 0.1474 23900 8.05E-06 489 .3 June 2008 WCAP- 16168-NP-A WCAP-16168-NP-A June 2i008 Revision 2

J-6 Table J-3 OCI-Specific Material Values Drawn from the RVID (see Ref. 44 Table 4.1) Major Material Region Description Un-Irradiated Cu Ni P Mn RTNDT,

        #        Type        Heat          Location     [wt%]    lwt%]. Iwt%]      [wt%1        [OF]

1 Axial Weld SA-1430 Lower 0.190 0.570 0.017 1.480 -5 2 Axial Weld SA-1493 intermediate 0.190 0.570 0.017 1.480 -5 3 Axial Weld SA-1073 Upper 0.210 0.640 0.025 1.380 -5 4 Circ Weld SA-1585 Lower 0.220 0.540 0.016 1.436 -5 5 Circ Weld SA-1229 Intermediate 0.230 0.590 0.021 1.488 10 6 Circ Weld SA-1135 Upper 0.230 0.520 0.011 1.404 -5 7 Plate C-2800 Lower 0.110 0.630 0.012 1.400 1 8 Plate C3265-1 Intermediate 0.100 0.500 0.015 1.420 1 9 Plate C3278-1 Intermediate 0.120 0.600 0.010 1.260 1 10 Plate C2197-2 Upper 0.150 0.500 0.008 1.280 1 11 Forging ZV2861 Upper 0.160 0.650 0.006 0.800 3

8. Weld metal details: Details of information used in addressing weld-specific information are taken directly from the NRC PTS Risk Study [44], Table 4.2. Summaries are reproduced as Table J-4.

Values for SAW Weld Volume fraction and Repair Weld Volume fraction in Table J-4 were changed to 96.7% and 2.3% respectively per NUREG- 1874 [9]. June 2008 WCAP- 16168-NP-A WCAP-16168-NP-A June 2008 Revision 2

J-7 Table J-4 Summary of Reactor Vessel-Specific Inputs for Flaw Distribution Inner Radius (td cladding) finj 85.51 78.51 86 8 Vessel specific info Base Metal Thickness [inL 8.438 7.875 8k5 8.675 V Vessel specific info Total Wall Thickness [in] 8.626 8.031 8.75 8.988 Vessel specific info Voiume iracuon Thru-Wall Bead All plants report plant specific [in] 011875 10,1875 0.1875 0.1875 dimensions of 3116-in, Thickness Judgment. Approx. 2X the size of the largest non-repair Tninrction Limit Fin] 1 flaw observed in PVRUF & Shoreham: Buried or Surface All flaws are buried Observation Observation: Virtually all of the weld flaws in.PVRUF &

                  .rientation              Circ flaws in circ welds,.axial flaws.in axial      Shoreham were aligned with Oreto                                       welds,                           the welding direction because they were. lack of'sidewall SAW                                                                                       fusion defects.

Weld- DenSity basis -- Shoreham density Highest of observations Statistically similar distributions trom Shoreham and PVRUF were combined to provide more robust Aspect ratio Shoreham & PVRUF observations estimates, when based on. basis, judgment the amount data were limited and/or insufficient to. identify different trends for aspect ratios for flaws in the two vessels. Statisticallysimilar distributions combined to Depth basis Shoreham & PVRUF observations provide, more, robust

              .1                ~       t estimates
                                                                                            .J___________________________           .3 WCAP- 16168-NP-A                                                                                                           June 2008 Revision 2

J-8I Table J-4 Summary of Reactor Vessel-Specific Inputs for Flaw Distribution (cont.) specific info provided by Volume fraction [1%] 1% Steve Byrne (Westinghouse - Windso0r. ._...... Oconee is generic value based on average of all Thru-Wall Bead plants specific values Thickness (in] 0.21 0.20 0,22 0.25 (including Shoreham & PVRUF data). Other values are plant specific as reported

                                                                                       %ySteveByrne.

Judgment. Approx. 2X the size of the largest non-repair Truncation Limit Vill) I flaw observed in PVRUF &

                                  --   I-----------------.---                     4.

Shoreham, Buried or Surface All flaws are buried Observation Observation: Virtually all of the weldflaws in PVRUF & SMAW Circ flaws in circ welds, axial flaws in axial Shoreham were aligned with Weld Orientation welds. the welding direction because they were lack of.sidewall fusion defects, Density basis Shoreham density Highest of observations Statistically similar distribution[s from Shoreham and PVRUF were combined to provide more robust Aspect ratio estimates, when based. on basis Shoreham & PVRUF observations judgment the: armount data were limited and/or insufficient to identify different trends for aspect ratios for flaws in the two vessels. Statistically similar Depth basis Shoreharn & PVRUF observations distributions-combinedto provide more robust estimates WCAP- 16168-NP-A June 2008 Revision: 2

J-9 Table J-4 Summary of Reactor Vessel-Specific Inputs for Flaw Distribution (cont.) Judgment. A rounoed integral percentage that Repair Volume fraction [%] exceeds the repaired volume Weld observed for Shoreham and for PVRUF, which was 1,5%. Generic value: As observed Thru-Wall Bead Thickness [in] 0,14 in PVRUF and Shoreham by IPNNL. Judgment. Approx. 2X the largest repair flaw found in Truncation Limit [in] PVRUF & Shoreham. Also based ion maximum expected width of repaircavity. Buried or Surf ace All flaws are buded Observation The repair flaws had complex shapes and orientations that were not aligned,with either the axial or circumferential Welds; for consistency with Circ flaws in circ welds, axial flaws in axial Orientation welds. the available treatments of flaws by the FAVOR. code, a common treatment of-orientations was adopted for flaws in SAW/SMAW and repair welds. Density basis -Shoreham density Highest of observations Statistically similar distributions from Shoreham and PVRUF were combined to.provide more robust Aspect ratio, - Shbreham &PVRUF observations estimates, when based on basis judgment .the a mount data were limited andlor insufficient to identify. different trends foraspect. ratios for flaws in the two vessels. Statistically similar distributions combined to Depth basis: Shoreham & PVRUF observations provide more robust estimates WCAP- 16168-NP-A June 2008 Revision 2

J-10 Table J-4 Summary of Reactor Vessel-Specific Inputs for Flaw Distribution (cont.) Cladding Actual Thickness [in] 0.188 0.156 0.25 0.313 Vessel specific info

              # ofLayers            ......        II             21             2.1         2    Vessel specific info Bead widths of I to 5-in.

characteristic of. machine deposited cladding. Bead widths down to %-in. can occur over welds. Nominal dimension of I-in, selected Bead Width 1 for all analyses because this [in] parameter is not expected to influence significantly the predicted vessel failure probabilities. May need to refine this estimate later, particularly for Oconee who reported a 5-in bead width. Truncation Limit [in] Actual clad thickness rounded to the nearest 11100d" of the total vessel wall thickness Judgment & computational Surface flaw [in] 0L 0-263 0.360 convenience depth in FAVOR Judgment. Only flaws in cladding that would influence brittle fracture of the vessel Buried orSurface All flaws are surface breaking are brittle. Material properties assigned to clad flaws are that of the underlying

                                                                                                .,material, be it base or weld.

Observation: All flaws observed in PVRUF & All circumferential. Shoreham were lack of inter-Orientation run fusion defects, and cladding is'always deposited circumferentiatlv No surface flaws observed. Density is 111000 that of the observed buried flaws in Density basis .cladding of vessels examined by PNNL If Judgment there is morethanIoneclad layer then there are no clad flaws. Aspect ratio Observations on buried flaws Judgment basis . I Depth of all surface flaws is the actual clad Depth basis of therounded thickness up.to the nearest 1/100d7 total vessel ,Judgment. wall thickness. June 2008 WCAP- 16168-NP-A WCAP-16168-NP-A June 2008 Revision 2

J-11 Table J-4 Summary of Reactor Vessel-Specific Inputs for Flaw Distribution (cont.) JUUYJ II t:imt., I Ut:IL'U Ll l il I Truncation Limit [in] 0.433 of the largest flaw observed in all PNNL p ate inspections. Buried or Surface All flaws are buried Observation Observation & Physics: No observed orientation Hall of the simulated flaws are Orientation preference, and no reason to circumferential, half-are axial. Plate suspect one (other than laminations which are benign. Density basis - 1/10 of small weld flaw density. 1140 of large Judgment, Supported by we!d flaw density of the PVRUF data limited data-Aspect ratio Same as for PVRUF welds Judgment basis Judgment. Supported by Depth basis Same as for PVRUF welds limited data.

9. TWCF95.TOTAL value calculated at 500 EFPY using correlations from NUREG-1874 (Reference 9): 3.16E-07 Events per year WCAP-16168-NP-A June 2008 Revision 2

K-1 APPENDIX K OCONEE UNIT 1 PROBSBFD OUTPUT WCAP-16168-NP-A June 2008 Revision 2

K-2 K-i: 10 Year ISI only STRUCTURAL RELIABILITY AND RISK ASSESSMENT (SRRA) WESTINGHOUSE MONTE-CARLO SIMULATION PROGRAM PROBSBFD VERSION 1.0 INPUT VARIABLES FOR CASE 3: OCI 10 YEAR ISI ONLY NCYCLE = 80 NFAILS = 1001 NTRIAL = 1000 NOVARS = 19 NUMSET = 2 NUMISI = 5 NUMSSC = 4 NUMTRC = 4 NUMFMD = 4 VARIABLE DISTRIBUTION MEDIAN DEVIATION SHIFT USAGE NO. NAME TYPE LOG VALUE OR FACTOR MV/SD NO. SUB 1 FIFDepth - CONSTANT - 3.OOOOD-02 1 SET 2 IFlawDen - CONSTANT - 3.6589D-03 2 SET 3 ICy-ISI - CONSTANT - 1. OOOOD+01 1 ISI 4 DCy-ISI - CONSTANT - 8 . OOOOD+01 2 ISIý, 5 MV-Depth - CONSTANT - 1.5000D-02 3 ISI 6 SD-Depth - CONSTANT - 1.8500D-01 4 ISI 7 CEff-ISI - CONSTANT - 1.0000D+00 5 ISI 8 Aspectl - CONSTANT - 2.OOOOD+00 1 SSC:! 9 Aspect2 - CONSTANT - 6.OOOOD+00 2 SSC 10 Aspect3 - CONSTANT - 1 . OOOOD+01 3 SSC 11 Aspect4 - CONSTANT - 9. 9000D+01 4 SSC 12 NoTr/Cy - CONSTANT - 1. 2000D+01 1 TRC 13 FCGTh1d - CONSTANT - 1.5000D+00 2 TRC 14 FCGR-UC NORMAL NO 0.OOOOD+00 1.OOOOD+00 .00 3 TRC 15 DKINFile - CONSTANT - 1.OOOOD+00 4 TRC; 16 Percentl - CONSTANT - 6. 7450D+01 1 FMD 17 Percent2 - CONSTANT - 2.0769D+01 2 FMD 18 Percent3 - CONSTANT - 3.9642D+00 3 FMD 19 Percent4 - CONSTANT - 7 . 8166D+00 4 FMD' INFORMATION GENERATED FROM FAVLOADS.DAT FILE AND SAVED IN DKINSAVE.DAT FILE: WALL THICKNESS = 8.6260 INCH FLAW DEPTH MINIMUM K AND MAXIMUM K FOR TYPE 1 WITH AN ASPECT RATIO OF 2. 8.62600D-02 2.26895D+00 1. 06757D+01 1.58718D-01 3.02106D+00 1. 44232D+01 4.3130OD-01 1.30893D+01 2. 08943D+01 6.46950D-01 1.39096D+01 2 .49826D+01 8.62600D-01 1.44263D+01 2. 80058D+01 1.72520D+00 1.30110D+01 3. 31903D+01 2.58780D+00 7.51977D+00 3.23837D+01 4.31300D+00 -2.67288D+00 3.20852D+01 TYPE 2 WITH AN ASPECT RATIO OF 6. WCAP- 16168-NP-A June 2008 Revision 2

K-3 K-i: 10 Year ISI only (cont.) 8.62600D-02 3.40901D+00 1.61172D+01 1.58718D-01 4.63620D+00 2.21942D+01 4.31300D-01 1.99455D+01 3.13897D+01 6.46950D-01 2.33230D+01 3.76625D+01 8.62600D-01 2.45197D+01 4.30412D+01 1.72520D+00 2.46021D+01 5.46183D+01 2.58780D+00 1.95704D+01 5.81373D+01 4.31300D+00 8.31986D+00 6.38027D+01 TYPE 3 WITH AN ASPECT RATIO OF 10. 8.62600D-02 3.73472D+00 1.76698D+01 1.58718D-01 4.95671D+00 2.37364D+01 4.31300D-01 2.11257D+01 3.35265D+01 6.46950D-01 2.53490D+01 4.01563D+01 8.62600D-01 2.66367D+01 4.59818D+01 1.72520D+00 2.73025D+01 5.94651D+01 2.58780D+00 2.36720D+01 6.65485D+01 4.31300D+00 1.21426D+01 7.64376D+01 TYPE 4 WITH AN ASPECT RATIO OF 99. 8.62600D-02 6.74437D+00 1.82354D+01 1.72520D-01 9.55233D+00 2.55450D+01 2.58780D-01 1.62039D+01 2.74271D+01 4.31300D-01 2.37153D+01 3.58624D+01 6.46950D-01 2.70360D+0i 4.44287D+01 8.6260OD-01 2.84566D+01 5.07281D+01 1.72520D+00 3.19293D+01 6.96665D+01 2.58780D+00 2.97815D+01 8.22041D+01 AVERAGE CALCULATED VALUES FOR: Surface Flaw Density with FCG and ISI NUMBER FAILED = 0 NUMBER OF TRIALS = 1000 DEPTH (WALL/400) AND FLAW DENSITY FOR ASPECT RATIOS OF 2, 6, 10 AND 99 12 2.2380D-04 1.0377D-05 1 .4547D-06 1.:1205D-05 13 6.5980D-06 4.0083D-05 7. 1947D-06 1.1813D-05 14 0.OOOOD+00 1.2906D-05 2 .8652D-06 2. 3081D-06 15 0.OOOOD+00 3.4523D-06 1. 0131D-06 4.5211D-07 16 0.OOOOD+00 1 . 1683D-06 2. 9704D-07 2. 7150D-07 17 0.OOOOD+00 5. 0981D-07 1. 5720D-07 1.2084D-07 18 0.OOOOD+00 3. 1177D-07 3. 5675D-08 7.1479D-08 19 0.OOOOD+00 1 .2295D-07 5.8386D-08 0.OOOOD+00 20 0.OOOOD+00 0 OOOOD+00 2. 2976D-08 0.OOOOD+00 22 0.OOOOD+00 5. 7099D-08 0 OOOOD+00 0.OOOOD+00 24 0.OOOOD+00 0 .OOOOD+00 0 OOOOD+00 2.2058D-08 25 0.OOOOD+00 5.4884D-08 1. 0551D-08 0.OOOOD+00 28 0.OOOOD+00 0.OOOOD+00 1.0078D-08 2.1150D-08 WCAP-16168-NP-A June 2008 Revision 2

K-4 K-2: ISI Every 10 Years STRUCTURAL RELIABILITY AND RISK ASSESSMENT (SRRA) WESTINGHOUSE MONTE-CARLO SIMULATION PROGRAM PROBSBFD VERSION 1.0 INPUT VARIABLES FOR CASE 2: OCI 10 YEAR INTERVAL NCYCLE 80 NFAILS = 1001 NTRIAL = 1000 NOVARS 19 NUMSET = 2 NUMISI = 5 NUMSSC 4 NUMTRC = 4 NUMFMD = 4 VARIABLE DISTRIBUTION MEDIAN DEVIATION SHIFT USAGE NO. NAME TYPE LOG VALUE OR FACTOR MV/SD NO. SUB 1 FIFDepth - CONSTANT - 3.OOOOD-02 1 SET 2 IFlawDen - CONSTANT - 3.6589D-03 2 SET 3 ICy-ISI - CONSTANT - 1. OOOOD+01 1 ISI 4 DCy-ISI - CONSTANT - 1. OOOOD+01 2 ISI 5 MV-Depth - CONSTANT - 1.5000D-02 3 ISI 6 SD-Depth - CONSTANT - 1.850OD-01 4 ISI 7 CEff-ISI - CONSTANT - 1. OOOOD+00 5 ISI, 8 Aspectl - CONSTANT - 2.000OD+00 1 SSC 9 Aspect2 - CONSTANT - 6. OOOOD+00 2 SSC 10 Aspect3 - CONSTANT - 1. OOOOD+01 3 SSC 11 Aspect4 - CONSTANT - 9 .9000D+01 4 SSC 12 NoTr/Cy - CONSTANT - 1. 2000D+01 1 TRC 13 FCGThld - CONSTANT - 1 .5000D+00 2 TRC 14 FCGR-UC NORMAL NO 0.OOOOD+00 1.OOOOD+00 .00 3 TRC 15 DKINFile - CONSTANT - 1.OOOOD+00 4 TRC 16 Percentl - CONSTANT - 6. 7450D+01 1 FMD 17 Percent2 - CONSTANT - 2.0769D+01 2 FMD 18 Percent3 - CONSTANT - 3.9642D+00 3 FMD 19 Percent4 - CONSTANT - 7. 8166D+00 4 FMD INFORMATION GENERATED FROM FAVLOADS.DAT FILE AND SAVED IN DKINSAVE.DAT FILE: WALL THICKNESS = 8.6260 INCH FLAW DEPTH MINIMUM K AND MAXIMUM K FOR TYPE 1 WITH AN ASPECT RATIO OF 2. 8.62600D-02 2.26895D+00 1.06757D+01

1. 58718D- 01 3.02106D+00 1.44232D+01 4.31300D-01 1.30893D+01 2.08943D+01 6.46950D-01 1.39096D+01 2.49826D+01 8.62600D-01 1.44263D+01 2.80058D+01 1.72520D+00 1.30110D+01 3.31903D+01 2.58780D+00 7.51977D+00 3.23837D+01 4.31300D+00 -2.67288D+00 3.20852D+01 TYPE 2 WITH AN ASPECT RATIO OF 6.

WCAP- 16168-NP-A June 2008 Revision 2

K-5 K-2: ISI Every 10 Years (cont.)

8. 62600D-02 3.40901D+00 1. 61172D+01

1. 58718D-01 4.63620D+00 2. 21942D+01 4.31300D-01 1.99455D+01 3. 13897D+01 6.46950D-01 2.33230D+01 3 .76625D+01

8. 62600D-01 2.45197D+01 4. 30412D+01

1. 72520D+00 2.46021D+01 5 .46183D+01 2 .58780D+00 1.95704D+01 5.81373D+01 4.31300D+00 8.31986D+00 6.38027D+01 TYPE 3 WITH AN ASPECT RATIO OF 10.

8.62600D-02 3.73472D+00 1.76698D+01 1.58718D-01 4.95671D+00 2.37364D+01 4.31300D-01 2.11257D+01 3.35265D+01 6.46950D-01 2.53490D+01 4.01563D+01 8.62600D-01 2.66367D+01 4.59818D+01 1.72520D+00 2.73025D+01 5.94651D+01 2.58780D+00 2.36720D+01 6.65485D+01 4.31300D+00 1.21426D+01 7.64376D+01 TYPE 4 WITH AN ASPECT RATIO OF 99. 8.62600D-02 6. 74437D+00 1. 82354D+01 1.72520D-01 9 .55233D+00 2 .55450D+01 2.58780D-01 1 .62039D+01 2.74271D+01 4.31300D-01 2. 37153D+01 3.58624D+01 6.46950D-01 2.70360D+01 4. 44287D+01 8.62600D-01 2 .84566D+01 5.07281D+01 1.72520D+00 3.19293D+01 6 .96665D+01 2.58780D+00 2. 97815D+01 8.22041D+01 AVERAGE CALCULATED VALUES FOR: Surface Flaw Density with FCG and ISI NUMBER FAILED = 0 NUMBER OF TRIALS = 1000 DEPTH (WALL/400) AND FLAW DENSITY FOR ASPECT RATIOS OF 2, 6, 10 AND 99 12 1 .3580D-10 5.4482D-12 7.5613D-13 6.1767D-12 13 2.8117D-12 1.4377D-11 2.5387D-12 4.4630D-12 14 0. 0000D+00 2.2869D-12 5.0820D-13 4.3208D-13 15 0.0000D+00 2 9908D-13 8.6948D-14 4.2493D-14 16 0. 0000D+00 4 .7816D-14 1.1866D-14 1.3716D-14 17 0. 0000D+00 1. 0793D-14 2.7598D-15 2.7273D-15 18 0. 0000D+00 2.8658D-15" 3.3064D-16 8.9749D-16 19 0.0000D+00 6.3484D-16 2.5927D-16 0.0000D+00 20 0.0000D+00 0.0000D+00 5.0956D-17 0.0000D+00 22 0.0000D+00 1.1431D-17 0.0000D+00 0.0000D+00 24 0.0000D+00 0.0000D+00 0.0000D+00 5.0464D-18 25 0.00000D00 1.4911D-18 3.6983D-19 0.0000D+00 28 0.0000D+00 0.0000D+00 2.2911D-20 2.7483D-19 WCAP-16168-NP-A June 2008 Revision 2

L- 1 APPENDIX L OCONEE UNIT 1 PTS TRANSIENTS Table L-I PTS Transient Descriptions for OCI TH Case Mean 1E Count # System Failure Operator Action Frequency HZP Hi K Dominant* 1 8 2.54 cm [1 in] surge line None 9.68E-08 No No No break with 1 stuck open safety valve in SG-A. 2 12 2.54 cm [1 in] surge line HPI throttled to maintain 27.8 9.24E-07 No No No break with I stuck open K [500 F] subcooling margin safety valve in SG-A. 3 15 2.54 cm [1 in] surge line At 15 minutes after transient 3.39E-08 No No No break with HPI Failure initiation, operator opens all TBVs to lower primary system pressure and allow CFT and LPI injection. 4 27 MSLB without trip of turbine Operator throttles HPI to 2.13E-06 No No No driven emergency feedwater. maintain 27.8 K [50' F] subcooling margin. 5 28 Reactor/turbine trip with 1 None 7.53E-08 No No No stuck open safety valve in SG-A 6 29 Reactor/turbine trip with 1 None 3.09E-07 No No No stuck open safety valve in SG-A and a second stuck open safety valve in SG-B 7 30 Reactor/turtine trip with 1 None 1.46E-07 Yes No No stuck open safety valve in SG-A 8 31 Reactor/turbine trip with 1 None 8.36E-09 Yes No No stuck open safety valve in SG-A and a second stuck open safety valve in SG-B 9 36 Reactor/turbine trip with 1 Operator throttles HPI to 1.40E-05 No No No stuck open safety valve in maintain 27.8 K [50' F] SG-A and a second stuck subcooling and 304.8 cm [120 open safety valve in SG-B in] pressurizer level. 10 37 Reactor/turbine trip with 1 Operator throttles HPI to 1.41E-06 Yes No No stuck open safety valve in maintain 27.8 K [500 F] SG-A subcooling and 304.8 cm [120 in] pressurizer level. 11 38 Reactor/turbine trip with 1 Operator throttles HPI to 2.65E-06 Yes No No stuck open safety valve in maintain 27.8 K [50' F] SG-A and a second stuck subcooling and 304.8 cm [120 open safety valve in SG-B in] pressurizer level. WCAP- 16168-NP-A June 2008 Revision 2

L-2 Table L-I PTS Transient Descriptions for OCI TH Case Mean IE Count # System Failure Operator Action Frequency HZP HiK Dominant* 12 44 2.54 cm [1 in] surge line At 15 minutes after initiation, 2.69E-07 No No No break with HPI Failure operators open all TBVs to depressurize the system to the CFT setpoint. When the CFTs are 50 percent discharged, HPI is assumed to be recovered. The TBVs are assumed remain open for the duration of the transient. 13 89 Reactor/turbine trip with Operator opens all TBVs to 5.388E-07 No No No,, Loss of MFW and EFW. 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 (609 cm [240 in] startup level). Operator controls booster pump flow to maintain SG level at 76 cm [30 in] due to continued RCP operation. Operator also throttles HPI to maintain 55 K [100EF] subcooling and a pressurizer level of 254 cm [100 in]. The TBVs are kept fully opened due to operator error. 14 90 Reactor/turbine trip with 2 Operator throttles HPI 20 6.29E-07 No No No stuck open safety valves in minutes after 2.7 K [50F] SG-A subcooling and 254 cm [100"] pressurizer level is reached [throttling criteria is 27.8 K [50'F] subcooling]. WCAP-16168-NP-A June 2008 Revision 2

L-3 Table L-1 PTS Transient Descriptions for OCI TH Case Mean IE Count # System Failure Operator Action Frequency HZP Hi K Dominant* 15 98 Reactor/turbine trip with loss Operator opens all TBVs to 9.96E-08 Yes No No of MFW and EFW 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 (610 cm [240 in] startup level). Operator controls booster pump flow to maintain SG level at 76 cm [30 in] due to continued RCP operation. Operator also throttles HPI to maintain 55 K [100EF] subcooling and a pressurizer level of 254 cm [100 in]. The TBVs are kept fully opened due to operator error. 16 99 MSLB with trip of turbine HPI is throttled 20 minutes 2.44E-07 No No No driven EFW by MSLB after 2.7 K [57F] subcooling Circuitry and 254 cm [100"] pressurizer level is reached (throttling criteria is 27.8 K [50OF] subcooling). 17 100 MSLB with trip of turbine Operator throttles HPI 20 5.11E-08 Yes No No driven EFW by MSLB minutes after 2.7 K [57F] Circuitry subcooling and 254 cm [100"] pressurizer level is reached (throttling criteria is 27.8 K [50'F] subcooling). 18 101 MSLB without trip of turbine Operator throttles HPI to 3.86E-07 Yes No No driven EFW by MSLB maintain 27.8 K [500 F] Circuitry subcooling margin (throttling criteria is 27.8 K [50'F] subcooling). 19 102 Reactor/turbine trip with 2 Operator throttles HPI 20 2.03E-07 Yes No No stuck open safety valves in minutes after 2.77 K [57F] SQ-A subcooling and 254 cm [100 in] pressurizer level is reached (throttling criteria is 27 K [50'F] subcooling). WCAP- 16168-NP-A June 2008 Revision 2

L-4 Table L-I PTS Transient Descriptions for OCI TH Case Mean IE Count # System Failure Operator Action Frequency HZP Hi K Dominant* 20 109 Stuck open pressurizer safety None 9.58E-06 No Yes No valve. Valve recloses at 6000 secs [RCS low pressure point]. 21 110 5.08 cm [2 inch] surge line At 15 minutes after transient 3.42E-06 No Yes Yes at 1000 break with HPI failure initiation, operator opens both EFPY TBV to lower primary system pressure and allow CFT and LPI injection. 22 111 2.54 cm [1 in] surge line At 15 minutes after initiation, 4.16E-07 No Yes No break with HPI failure 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 [100lF] subcooling and 254 cm [100"] pressurizer level. 23 112 Stuck open pressurizer safety After valve recloses, operator 1.25E-04 No Yes No valve. Valve recloses at 6000 throttles HPI 1 minute after 2.7 secs. K [5°F] subcooling and 254 cm [100"] pressurizer level is reached (throttling criteria is 27 K [50'F] subcooling) 24 113 Stuck open pressurizer safety After valve recloses, operator 5.07E-05 No Yes No valve. Valve recloses at 6000 throttles HPI 10 minutes after secs. 2.7 K [5°F] subcooling and 254 cm [100"] pressurizer level is reached (throttling criteria is 27.8 K [50'F] subcooling) 25 114 Stuck open pressurizer safety After valve recloses, operator 1.25E-04 No Yes No valve. Valve recloses at 3000 throttles HPI 1 minute after 2.7 secs. K [5°F] subcooling and 254 cm [100"] pressurizer level is reached (throttling criteria is 50'F subcooling) 26 115 Stuck open pressurizer Safety After valve recloses, operator 5.07E-05 No Yes No Valve. Valve recloses at 3000 throttles HPI 10 minutes after secs. 2.7 K [5°F] subcooling and 254 cm [100"] pressurizer level is reached (throttling _ _criteria is 50'F subcooling) WCAP- 16168-NP-A June 2008 Revision 2

L-5 Table L- I PTS Transient Descriptions for OCI TH Case Mean IE Count # System Failure Operator Action Frequency HZP Hi K Dominant* 27 116 Stuck open pressurizer safety At 15 minutes after initiation, 2.60E-07 No Yes No valve and HPI failure 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 [50 F] subcooling and 254 cm [100"] pressurizer level is reached (throttling criteria is 50'F subcooling). 28 117 Stuck open pressurizer safety At 15 minutes after initiation, 5.38E-07 No Yes No valve and HPI failure 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 2.7 K [57F] subcooling and 254 cm [100"] pressurizer level is reached (throttling criteria is 27.8 K [50'F] siibcooling). 29 119 2.54 cm [1 in] surge line At 15 minutes after transient 4.41E-07 Yes Yes No break with HPI Failure initiation, the operator opens all turbine bypass valves to lower primary system pressure and allow core flood tank and LPI injection. 30 120 2.54 cm [1 in] surge line At 15 minutes after sequence 4.22E-08 Yes Yes No break with HPI Failure initiation, operators open all TBVs to depressurize the system to the CFT setpoint. When the CFTs are 50 percent discharged, HPI is assumed to be recovered. The TBVs are assumed remain opened for the duration of the transient. 31 121 Stuck open pressurizer safety Operator throttles HPI at 1 2.28E-05 Yes Yes No valve. Valve recloses at 6000 minute after 2.7 K [5°F] secs. subcooling and 254 cm [100"] pressurizer level is reached [throttling criteria is 27.8 K _[50'F] subcooling]. WCAP- 16168-NP-A June 2008 Revision 2

L-6 Table L-1 PTS Transient Descriptions for OC1 TH Case Mean IE Count # System Failure Operator Action Frequency HZP Hi K Dominant* 32 122 Stuck open pressurizer safety Operator throttles HPI at 10 7.57E-06 Yes Yes Yes at 32, valve. Valve recloses at 6000 minutes after 2.7 K [5°F] 60, 500, secs. subcooling and 254 cm [100"] 1000 EFPY pressurizer level is reached (throttling criteria is 27.8 K [50[F] subcooling). 33 123 Stuck open pressurizer safety Operator throttles HPI at 1 2.28E-05 Yes Yes No valve. Valve recloses at 3000 minute after 2.7 K [5°F] secs. subcooling and 254 cm [100"] pressurizer level is reached (throttling criteria is 27.8 K [50°] subcooling). 34 124 Stuck open pressurizer safety Operator throttles HPI at 10 7.57E-06 Yes Yes Yes at 60, valve. Valve recloses at 3000 minutes after 2.7 K [5°F] 500, 1000 secs. subcooling and 254 cm [100"] EFPY pressurizer level is reached (throttling criteria is 27.8 K [50F] subcooling). 35 125 Stuck open pressurizer safety At 15 minutes after initiation, 4.61E-08 Yes Yes No valve and HPI Failure 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 [57] subcooling and 254 cm [100"] pressurizer level is reached (throttling criteria is 27.8 K [50°F] subcooling). 36 126 Stuck open pressurizer safety At 15 minutes after initiation, 8.41E-08 Yes Yes No valve and HPI Failure 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 2.7 K [5°F] subcooling and 254 cm [100"] pressurizer level is reached (throttling criteria is 27.8 K [507F] subcooling). WCAP- 16168-NP-A June 2008 Revision 2

L-7 Table L-1 PTS Transient Descriptions for OCI TH Case Mean IE Count # System Failure Operator Action Frequency HZP Hi K Dominant* 37 127 SGTR with a stuck open SRV Operator trips RCP's 1 minute 1.25E-07 Yes Yes No in SG-B. A reactor trip is after initiation. Operator also assumed to occur at the time throttles HPI 10 minutes after of the tube rupture. Stuck 2.77 K [50 F] subcooling and safety relief valve is assumed 254 cm [100 in] pressurizer to reclose 10 minutes after level is reached (assumed initiation, throttling criteria is 27 K [50°F] subcooling). 38 141 8.19 cm [3.22 in] surge line None 1.06E-04 No Yes Yes at 500, break [Break flow area 1000 EFPY increased by 30% from 7.18 cm [2.828 in] break]. 39 142 6.01 cm [2.37 in] surge line None 1.06E-04 No Yes No break [Break flow area decreased by 30% from 7.18 cm [2.828 in] break]. 40 145 4.34 cm [1.71 in] surge'line None 1.34E-04 No Yes No break [Break flow area increased by 30% from 3.81 cm [1.5 in] break]. Winter conditions assumed [HPI, LPI temp = 277 K [400 F] and CFT temp = 294 K [70' F)). 41 146 TT/RT with stuck open pzr None 4.23E-05 No Yes No SRV [valve flow area reduced by 30 percent]. Summer conditions assumed [HPI, LPI temp = 302 K [85' F] and CFT temp = 310 K [1000 F)). Vent valves do not function. 42 147 TT/RT with stuck open pzr None 3.63E-05 No Yes No SRV. Summer conditions assumed [HPI, LPI temp = . 302 K [85' F] and CFT temp

                 = 310 K [1000 F)).

43 148 TT/RT with partially stuck None 4.23E-05 No Yes No open pzr SRV [flow area equivalent to 1.5 in diameter opening]. HTC coefficients increased by 1.3. 44 149 TT/RT with stuck open pzr None 9.58E-06 No Yes No SRV. SRV assumed to reclose at 3000 secs. Operator does not throttle HPI. WCAP- 16168-NP-A June 2008 Revision 2

L-8 Table L-I PTS Transient Descriptions for OC1 TH Case Mean IE Count # System Failure Operator Action Frequency HZP Hi K Dominant* 45 154 8.53 cm [3.36 in] surge line None 1.34E-04 No Yes No 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. 46 156 40.64 cm [16 in] hot leg None 7.03E-06 No Yes Yes at 500, break. ECC suction switch 1000 EFPY to the containment sump included in the analysis. 47 160 14.37 cm [5.656 in] surge line None 1.82E-05 No Yes Yes at 500, break. ECC suction switch to 1000 EFPY the containment sump included in the analysis. ,_ 48 164 20.32 cm [8 inch] surge line None 2.12E-05 No Yes Yes at 60, break. ECC suction switch to 500, 1000 the containment sump EFPY included in the analysis. 49 165 Stuck open pressurizer safety None 1.76E-06 Yes Yes Yes at 32, valve. Valve recloses at 6000 60, 500, secs [RCS low pressure 1000 EFPY point]. 50 168 TT/RT with stuck open pzr None 1.76E-06 Yes Yes Yes at 500, SRV. SRV assumed to 1000 EFPY reclose at 3000 secs. Operator does not throttle HPI. 51 169 TT/RT with stuck open pzr None 7.33E-06 Yes Yes No SRV [valve flow area reduced by 30 percent]. Summer conditions assumed [HPI, LPI temp = 302 K [850 F] and CFT temp = 310 K [1000 F)). Vent valves do not function. 52 170 TT/RT with stuck open pzr None 6.28E-06 Yes Yes No SRV. Summer conditions assumed [HPI, LPI temp = 302 K [85' F] and CFT temp

                 = 310 K [100- FI].

53 171 TT/RT with partially stuck None 7.33E-06 Yes Yes No open pzr SRV [flow area equivalent to 1.5 in diameter opening]. HTC coefficients increased by 1.3. WCAP- 16168-NP-A June 2008 Revision 2

L-9 Table L-1 PTS Transient Descriptions for OC1 TH Case Mean IE Count # System Failure Operator Action Frequency HZP Hi K Dominant* 54 172 10.16 cm [4 in] cold leg None 1.06E-04 No Yes Yes at 1000 break. ECC suction switch EFPY to the containment sump included in the analysis. 55 178 8.53 cm [3.36 in] surge line None 2.12E-05 No Yes No 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._ Notes:

1. TH - Thermal hydraulics

2. LOCA - Loss-of-coolant accident

3. SBLOCA- Small-break loss-of-coolant accident

4. MBLOCA - Medium-break loss-of-coolant accident

5. LBLOCA - Large-break loss-of-coolant accident

6. HZP - Hot-zero power

7. SRV - Safety and relief valve

8. MSLB - Main steam line break

9. AFW- Auxiliary feedwater

10. HPI - High-pressure injection

11. RCPs - Reactor coolant pumps

• The arbitrary definition of a dominant transient is a transient that contributes 1% or more of the total Through-Wall Cracking Failure (TWCF).

June 2008 WCAP- 16168-NP-A WCAP-16168-NP-A June 2008 Revision 2

M-1 APPENDIX M OCONEE UNIT 1 FAVPOST OUTPUT WCAP- 16168-NP-A June 2008 Revision 2

M-2 M-1: 10 Year ISI only

                 *¢ WELCOME TO FAVOR FRACTURE ANALYSIS OF VESSELS:     OAK RIDGE       *

• VERSION 06.1 FAVPOST MODULE: POSTPROCESSOR MODULE *

                  *k COMBINES TRANSIENT INITIAITING FREQUENCIES         *
                  *k WITH RESULTS OF PFM ANALYSIS                *
                  *k PROBLEMS OR QUESTIONS REGARDING FAVOR
                  *k SHOULD BE DIRECTED TO TERRY DICKSON OAK RIDGE NATIONAL LABORATORY e-mail:   dicksontl@ornl.gov              *

• This computer program was prepared as an account of *
• work sponsored by the United States Government *
• Neither the United States, nor the United States *
• Department of Energy, nor the United States Nuclear *
• Regulatory Commission, nor any of their employees, *
• nor any of their contractors, subcontractors, or their *
• employees, makes any warranty, expressed or implied, or *
• assumes any legal liability or responsibility for the *
• accuracy, completeness, or usefulness of any *
• information, apparatus, product, or process disclosed, *
• or represents that its use would not infringe *
• privately-owned rights.
• DATE: 27-Jun-2007 TIME: 10:37:21 Begin echo of FAVPost input data deck 10:37:21 27-Jun-2007 End echo of FAVPost input data deck 10:37:21 27-Jun-2007 FAVPOST INPUT FILE NAME = postoc55.in FAVPFM OUTPUT FILE CONTAINING PFMI ARRAY = INITIATE.DAT FAVPFM OUTPUT FILE CONTAINING PFMF ARRAY = FAILURE.DAT FAVPOST OUTPUT FILE NAME = ocpostlOyronly.out WCAP- 16168-NP-A June 2008 Revision 2

M-3 M-1: 10 Year ISI only (cont.)

• NUMBER OF SIMULATIONS = 42000
• CONDITIONAL PROBABILITY C(ONDITIONAL PROBA BILITY OF INITIATION CPI=P(IjE) OF FAILURE CPF=P (FIE)

TRANSIENT MEAN 95th % 99th % MEAN 95th % 99th % RATIO NUMBER CPI CPI CPI CPF CPF CPF CPFmn/CPImn 0.OOOE+0 0000 8 0.OOOOE+00 0.OOOOE+00 0 OOOOE+00 0. 0000E+00 0.OOOOE+00 0.0000E+00 0.0000 12 0.OOOOE+00 0.OOOOE+00 0 OOOOE+00 0.OOOOE+00 0.0000E+00 0.0000E+00 0.0000 15 1.3573E-07 0.OOOOE+00 0 OOOOE+00 1. 8219E-12 0.OOOOE+00 0.0000E+00 0.0000 27 3. 5321E-06 0.000OE+00 9.7209E-06 1.0415E-08 0.0000E+00 0.0000E+00 0.0029 28 0.OOOOE+00 0 OOOOE+00 0 OOOOE+00 0. OOOOE+00 0.0000E+00 0.0000E+00 0.0000 29 0.OOOOE+00 0 .OOOOE+00 0 OOOOE+00 0. OOOOE+00 0.00OOE+00 0.0000E+00 0.0000 30 0.OOOOE+00 0 .OOOOE+00 0 OOOOE+00 0. OOOOE+00 0.0000E+00 0.OOOOE+00 0.0000 31 0.OOOOE+00 0 OOOOE+00 0 OOOOE+00 0. OOOOE+00 0.0000E+00 0.OOOOE+00 0.0000 36 0.OOOOE+00 0 OOOOE+00 0 OOOOE+00 0. OOOOE+00 0.0000E+00 0.OOOOE+00 0.0000 37 0.OOOOE+00 0 OOOOE+00 0. OOOE+00 0. OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.0000 0 OOOOE+00 0. 0000E+00 0.OOOOE+00, 0 0000E+00 0.0000 38 0.OOOOE+00 0 OOOOE+00 44 4. 1737E-06 0 OOOOE+00 4 .4260E-06 2.8306E-06 0.OOOOE+00 1.7178E-06 0.6782 89 0.OOOOE+00 0 OOOOE+00 0.OOOOE+00 0.0000E+00 0.0000E+00 0.0000E+00 0.0000 90 0.OOOOE+00 0 OOOOE+00 0 OOOOE+00 0. OOOOE+00 0.0000E+00 0.0000E+00 0.0000 98 0.OOOOE+00 0 .OOOOE+00 0 OOOOE+00 0. OOOOE+00 0.0000E+00 0.0000E+00 0.0000 99 5.9408E-07 0 OOOOE+00 0 OOOOE+00 1.4020E-08 0 OOOOE+00 0.OOOOE+00 0.0236 100 1.7278E-06 0 OOOOE+00 3.6037E-07 4. 0017E-07 0.0000E+00 0.0000E+00 0.2316 101 1 .4057E-05 0 .OOOOE+00 5.7634E-05 1.4131E-08 0. 0000E+00 0.0000E+00 0.0010 102 0 OOOOE+00 0 OOOOE+00 0.OOOOE+00 0. OOOOE+00 0. 0000E+00 0.OOOOE+00 0.0000 109 1 .2464E-07 0 OOOOE+00 0.OOOOE+00 4. 0517E-08 0. 0000E+00 0.0000E+00 0.3251 110 2. 5705E-03 5.9843E-03 3.2961E-02 1.7099E-05 1. 1551E-04 2.3858E-04 0.0067 111 4 .5456E-08 0.OOOOE+00 0.0000E+00 1.4016E-11 0 OOOOE+00 0.0000E+00 0.0003 112 8. 3691E-10 0.OOOOE+00 0.0000E+00 0. OOOOE+00 0 OOOOE+00 0.OOOOE+00 0.0000 113 3 . 1674E-08 0.OOOOE+00 0.OOOOE+00 2.9159E-08 0. 0000E+00 0.0000E+00 0.9206 114 0.000OE+00 0.OOOOE+00 0.000OE+00 0. OOOOE+00 0. 0000E+00 0.OOOOE+00 0.0000 WCAP- 16168-NP-A June 2008 Revision 2

M-4 M-1: 10 Year ISI only (cont.) 115 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.000OE+00 0.OOOOE+00 0 .OOOOE+00 0.0000 116 2. 6284E-06 0.OOOOE+00 3.2436E-07 1. 1695E-09 0 OOOOE+00 0 OOOOE+00 0.0004 117 5. 3511E-05 5.7899E-04 1.0032E-03 6 .4161E-08 0 OOOOE+00 1 .1859E-13 0.0012 119 9 .2288E-05 5. 7141E-04 1.4015E-03 6.2724E-07 1.2359E-05 3. 9221E-06 0.0068 120 3. 3053E-05 4.2682E-04 4.6996E-04 2.4353E-05 4. 0226E-04 3. 3238E-04 0.7368 121 7. 5802E-08 0.OOOOE+00 0.OOOOE+00 2. 9336E-12 0 OOOOE+00 0 .OOOOE+00 0.0000 122 4 .2840E-04 2.6185E-03 4.5275E-03 4. 2818E-04 2. 6185E-03 4 .5275E-03 0.9995 123 7.5802E-08 0.OOOOE+00 0.OOOOE+00 2. 9336E-12 0 .OOOOE+00 0.000OE+00 0.0000 124 1.6752E-04 1.1252E-03 2.2503E-03 1. 6517E-04 1. 1140E-03 2. 2106E-03 0.9860 125 7.0273E-05 5.4000E-04 9.7650E-04 2. 7388E-07 0 OOOOE+00 7: 8181E-07 0.0039 126 1.1462E-06 0.OOOOE+00 6.9098E-08 8. 8915E-10 0 .OOOOE+00 0.OOOOE+00 0.0008 127 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0 OOOOE+00 0 .OOOOE+00 0.OOOOE+00 0.0000 141 1.3088E-04 6.5253E-04 2.1093E-03 2 .4722E-06 5.4576E-05 2. 1956E-05 0.0189 142 4. 5999E-10 0.OOOOE+00 0.OOOOE+00 0 .OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.0000 145 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0 .OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.0000 146 2.7129E-06 0 OOOOE+00 6. 0255E-07 9.5223E-08 0.OOOOE+00 0.OOOOE+00 0.0351 147 0.OOOOE+00 0 OOOOE+00 0 OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.0000 148 0.OOOOE+00 0 .OOOOE+00 0 OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.0000 149 0.OOOOE+00 0 .OOOOE+00 0 OOOOE+00 0.0000E+00 0.OOOOE+00 0.OOOOE+00 0. 0000 154 1.4610E-04 9 .2152E-04 2 .2196E-03 5.0292E-07 0.OOOOE+00 0.OOOOE+00 0. 0034 156 2 .2010E-02 5 .2143E-02 1. 7530E-01 8. 9468E-05 3.2180E-04 1.4740E-03 0.0041 160 1.3270E-02 2. 8340E-02 1. 0630E-01 1. 7457E-04 4.0235E-04 2.9084E-03 0.0132 164 1.3995E-02 3. 2240E-02 1. 2584E-01 8. 8292E-05 2.5946E-04 1.5275E-03 0.0063 165 3.9220E-04 2. 6738E-03 3. 0921E-03 3 9180E-04 2.6738E-03 3. 0921E-03 0.9990 168 2. 0391E-04 1.2392E-03 2. 8916E-03 2 .0139E-04 1.2392E-03 2.8344E-03 0.9876 169 2.1163E-04 1.0072E-03 3.3572E-03 6. 7665E-06 1.0975E-04 8.2906E-05 0.0320 170 4 .6761E-08 0.OOOOE+00 0.OOOOE+00 4.6483E-13 0.OOOOE+00 0.0000 171 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.0000 172 8.0788E-05 5.5702E-04, 1.1954E-03 3.2570E-07 0.OOOOE+00 1. 2316E-06 0.0040 178 1.4610E-04 9.2152E-04 2.2196E-03 5.0292E-07 0.OOOOE+00 0.OOOOE+00 0.0034 NOTES: CPI IS CONDITIONAL PROBABILITY OF CRACK INITIATION, P(IJE) CPF IS CONDITIONAL PROBABILITY OF TWC FAILURE, P(FIE) WCAP- 16168-NP-A June 2008 Revision 2

M-5 M-1: 10 Year ISI only (cont.)

• PROBABILITY DISTRIBUTION FUNCTION (HISTOGRAM) *
• FOR THE FREQUENCY OF CRACK INITIATION
• FREQUENCY OF RELATIVE CUMULATIVE CRACK INITIATION DENSITY DISTRIBUTION (PER REACTOR-OPERATING YEAR) (%) (%)

0.OOOOE+00 2.9119 2.9119 2.1448E-06 92.0857 94.9976 6.4345E-06 2.9429 97.9405 1.0724E-05 0.8714 98.8119 1.5014E-05 0.4571 99.2690 1.9304E-05 0.2405 99.5095 2.3593E-05 0.1524 99.6619 2.7883E-05 0.0905 99.7524 3.2173E-05 0.0619 99.8143 3.6462E-05 0.0357 99.8500 4.0752E-05 0.0214 99.8714 4.5042E-05 0.0214 99.8929 4.9331E-05 0.0214 99.9143 5.3621E-05 0.0167 99.9310 5.7911E-05 0.0095 99.9405 6.2200E-05 0.0048 99'.9452 6.6490E-05 0.0071 99.9524 7.0780E-05 0.0095 99.9619 7.5069E-05 0.0071 99.9690 7.9359E-05 0.0048 99.9738 8.7938E-05 0.0024 99.9762 9.6518E-05 0.0048 99.9810 1.0081E-04 0.0024 99.9833 1.0510E-04 0.0048 99.9881 1.1368E-04 0.0024 99.9905 1.3512E-04 0.0024 99.9929 1.3941E-04 0.0024 99.9952 1.6944E-04 0.0024 99.9976 2.0376E-04 0.0024 100.0000

                              ==      Summary Descriptive    Statistics             ==

Minimum = 0.OOOOE+00 Maximum = 2.0449E-04 Range = 2.0449E-04 Number of Simulations = 42000 5th Percentile = 3.8158E-12 Median = 1.1260E-07 95.0th Percentile = 2.1448E-06 WCAP- 16168-NP-A June 2008 Revision 2

M-6 M-1: 10 Year ISI only (cont.) 99.0th Percentile = 1.2489E-05 99.9th Percentile = 4.6472E-05 Mean = 9.9210E-07 Standard Deviation = 3.7931E-06 Standard Error = 1.8508E-08 Variance (unbiased) = 1.4387E-11 Variance (biased) = 1.4387E-11 Moment Coeff. of Skewness = 1.6997E+01 Pearson's 2nd Coeff. of Skewness = 7.8467E-01 Kurtosis = 5.3092E+02

• PROBABILITY DISTRIBUTION FUNCTION (HISTOGRAM) *
• FOR THROUGH-WALL CRACKING FREQUENCY (FAILURE)
• FREQUENCY OF RELATIVE CUMULATIVE TWC FAILURES DENSITY DISTRIBUTION (PER REACTOR-OPERATING YEAR) (%) (%)

0.OOOOE+00 23.0000 23.0000 3.5468E-07 76.3452 99.3452 1.0641E-06 0.3333 99.6786 1.7734E-06 0.1357 99.8143 2.4828E-06 0.0429 99.8571 3.1922E-06 0.0405 99.8976 3.9015E-06 0.0119 99.9095 4.6109E-06 0.0143 99.9238 5.3203E-06 0.0119 99.9357 6.0296E-06 0.0095 99.9452 6.7390E-06 0.0143 99.9595 7.4484E-06 0.0071 99.9667 8.1577E-06 0.0048 99.9714 8.8671E-06 0.0048 99.9762 1.0286E-05 0.0048 99.9810 1.0995E-05 0.0024 99.9833 1.1705E-05 0.0048 99.9881 1.3833E-05 0.0024 99.9905 1.4542E-05 0.0024 99.9929 2.6601E-05 0.0024 99.9952 5.4267E-05 0.0024 99.9976 7.0582E-05 0.0024 100.0000

                              ==       Summary Descriptive     Statistics              ==

Minimum = 0.OOOOE+00 WCAP- 16168-NP-A June 2008 Revision 2

M-7 M-1: 10 Year ISI only (cont.) Maximum = 7.0227E-05 Range = 7.0227E-05 Number of Simulations = 42000 5th Percentile = 0.0000E+00 Median = 9.3729E-11 95.0th Percentile = 3.5468E-07 99.0th Percentile = 4.7205E-07 99.9th Percentile = 3.3340E-06 Mean = 3.1118E-08 Standard Deviation 5.2306E-07 Standard Error = 2.5523E-09 Variance (unbiased) = 2.7359E-13 Variance (biased) = 2.7358E-13 Moment Coeff. of Skewness = 9.0777E+01 Pearson's 2nd Coeff. of Skewness =-4.6734E-01 Kurtosis = 1.0697E+04

• FRACTIONALIZATION OF FREQUENCY OF CRACK INITIATION *
• AND THROUGH-WALL CRACKING FREQUENCY (FAILURE) - *
• WEIGHTED BY TRANSIENT INITIATING FREQUENCIES *

                                           % of total             % of total frequency of          frequency of crack initiation        of TWC failure 8            0.00                    0.00 12             0.00                    0.00 15             0.00                     0.00 27             0.00                     0.00 28             0.00                     0.00 29             0.00                     0.00 30             0.00                     0.00 31             0.00                     0.00 36             0.00                     0.00 37             0.00                     0.00 38             0.00                     0.00 44             0.00                     0.00 89             0.00                     0.00 90             0.00                     0.00 98            0.00                     0.00 99            0.00                     0.00 100             0.00                     0.00 101             0.00                     0.00 102             0.00                     0.00 109             0.00                     0.00 110                                     -0.22 i1              0.00                     0.00 112             0.00                     0.00 June  2008 WCAP- 161 68-NP-A WCAP-16168-NP-A                                                                  June 2008 Revision 2

M-8 M-1: 10 Year 1ST only (cont.) 112 0.00 0.02 114 0. 00 0.00 115 0.00 0. 00 116 0.00 0.00 117 0.00 0.00 119 0.00 0.00 120 0.00 0.00 121 0.00 0. 00 122 1.38 44.11 123 0.00 0.00 124 0.52 16.27 125 0.00 0. 00 126 0. 00 0.00 127 0.00 0.00 141 1.58 0.91 142 0.00 0. 00 145 0.00 0.00 146 0 .04 0 .04 147 0.00 0.00 148 0.00 - 0. 00 149 0.00 0.00 154 2 .36 0.34 156 21.78 2 .62 160 31.29 12.37 164 37.57 7.32 165 0.30 9.70 168 0.17 5.29 169 0.58 0.60 170 0.00 0.00 171 0.00 0.00 172 0.93 0.12 178 0.37 0.05 TOTALS 100.00 100.00

• FRACTIONALIZATION OF FREQUENCY OF CRACK INITIATION *
• AND THROUGH-WALL CRACKING FREQUENCY (FAILURE) -
• BY
• RPV BELTLINE MAJOR REGION *
• BY PARENT SUBREGION *

                    *WEIGHTED         BY % CONTRIBUTION OF EACH TRANSIENT*
                *             ~TO FREQUENCY OF CRACK INITIATION AND*
             *              ~THROUGH-WALL CRACKING FREQUENCY (FAILURE)*

June 2008 WCAP- 161 68-NP-A WCAP-16168-NP-A Revision 2

M-9 M-1: 10 Year ISI only (cont.)

                                                                      % of total
                           % of           % of total            through-wall crack MAJOR       RTndt      total        frequency of                   frequency REGION       (MAX)     flaws     crack initiation       cleavage ductile total 1      223.80      3.03              2 .10            18.32        2.77     21.09 2      220.74      3.54              4 .01            36.95        5.25     42.21 3      253.04      1.43              3 .54            26.25       10.23     36.48 4      236.56     13.81             14 .70             0.06        0.00      0.06 5      277.07     13.81             75.50              0.14        0. 00     0.14 6      168.76     13.81              0.09              0.00        0.00      0.00 7      158.07     18.93              0 .02             0.01        0.00      0.01 8      152.82     11.05              0 .01             0.00        0.00      0.00 9     154.58     11.05              0 .02             0.01        0.00      0.02 10      155.89      8.92              0.01              0.00        0.00      0.00 11      101.31      0.62              0.00              0.00        0.00      0.00 TOTALS 100.00               100.00             81.74       18.26    100.00

• FRACTIONALIZATION OF FREQUENCY OF CRACK INITIATION *
• AND THROUGH-WALL CRACKING FREQUENCY (FAILURE) -
• BY
• RPV BELTLINE MAJOR REGION *
• BY CHILD SUBREGION *
• WEIGHTED BY % CONTRIBUTION OF EACH TRANSIENT *
• TO FREQUENCY OF CRACK INITIATION AND *
• THROUGH-WALL CRACKING FREQUENCY (FAILURE) *

                                                                      % of total
                            % of          % of total             through-wall crack MAJOR       RTndt      total        frequency of                   frequency REGION       (MAX)     flaws     crack initiation       cleavage ductile total 1      223.80      3 .03             2.10            18.32         2.77    21.09 2      220.74      3 .54             4.01            36.95         5.25    42.21 3      253.04      1.43              3.54            26.25        10.23    36.48 4      236.56    13.81              14.70              0.06        0.00      0.06 5      277.07    13 .81             75.50              0.14        0.00       0.14 6      168.76    13 .81              0.09              0.00        0.00       0.00 7      158.07    18. 93              0.02              0.01        0.00       0.01 8      152.82    11.05               0.01              0.00        0.00       0.00 9     154.58    11.05               0.02              0.01        0.00       0.02 10      155.89       8.92             0.01              0.00        0.00       0.00 11      101.31       0.62             0.00              0.00        0.00       0.00 TOTALS 100.00               100.00             81.74       18.26   100.00 WCAP- 16168-NP-A                                                                           June 2008 Revision 2

M-10 M-1: 10 Year ISI only (cont.)

• FRACTIONALIZATION OF FREQUENCY OF CRACK INITIATION *
• AND THROUGH-WALL CRACKING FREQUENCY (FAILURE) *
• MATERIAL, FLAW CATEGORY, AND FLAW DEPTH *
• WEIGHTED BY % CONTRIBUTION OF EACH TRANSIENT *
• TO FREQUENCY OF CRACK INITIATION AND *
• THROUGH-WALL CRACKING FREQUENCY (FAILURE) *
• WELD MATERIAL *

                             % of total      frequency          % of total      through-wall of crack initiation                     crack frequency FLAW DEPTH      CAT I        CAT 2       CAT 3       CAT 1        CAT 2   CAT 3 (in)      flaws        flaws       flaws       flaws        flaws    flaws 0.086        0.00         3.46       0.00         0.00        0.89       0.00 0 .173       0.00       43.13        0.00         0.00       12.86       0.00 0.259        0.00       15.98        0.00         0.00        6.44       0.00 0.345        0.00         9.47       0.00         0.00        5.39       0. 05 0.431        0.00         7.02       0.01         0.00         6.88       0.17 0.00         4.91       0.01                      7.10 0.518                                             0.00                    0.31

0. 604 0.00 3.69 0. 02 0.00 4.99 0.50 0.690 0.00 2 .24 0.02 0.00 5.04 0.48 0.776 0.00 1.85 0.01 0.00 4.00 0.45 0.863 0.00 1.97 0.02 0.00 5.08 0.52 0.949 0.00 0.88 0.02 0.00 4.28 0.65 1.035 0.03 1.03 0.01 0.00 4.67 0.20 1.121 0.04 0.70 0.00 0.00 2.85 0.10 1.208 0.01 0.61 0.01 0.00 2.08 0.21 1.294 0.00 0.66 0. 01 0.00 3.93 0.23 1.380 0.00 0.86 0.01 0.00 3.34 0.30 1.466 0.00 0.20 0.01 0.00 3.29 0.19 1.553 0.00 0.25 0.00 0.00 2.25 0.13 1.639 0.00 0.24 0. 01 0.00 0.81 0.20 1.725 0. 00 0.12 0.00 0.00 1.45 0.03

1. 811 0.00 0.38 0.00 0.00 5.61 0.06 1.898 0.00 0.05 0.00 0.00 1 .62 0.02

1. 984 0.00 0.04 0.00 0.00 0.25 0.02 2 .070 0.00 0.00 0.00 0.00 0.00 0.00 2.157 0.00 0.00 0.00 0.00 0.00 0.00 2 .243 0.00 0.00 0.00 0.00 0.00 0.00 2.329 0.00 0.00 0.00 0.00 0.00 0.00 2.415 0.00 0.00 0.00 0.00 0.00 0.00 TOTALS 0.08 99.72 0.15 0.00 95.13 4.85 June 2008 WCAP- 161 68-NP-A WCAP-16168-NP-A June 2008 Revision 2

M-11 M-1: 10 Year ISI only (cont.)

• PLATE MATERIAL *

                             % of total     frequency          % of total    through-wall of crack initiation                    crack frequency FLAW DEPTH       CAT I        CAT 2       CAT 3      CAT 1      CAT 2    CAT 3 (in)       flaws        flaws       flaws      flaws      flaws    flaws 0.086         0.00        0.00         0.00       0.00      0.00        0.00 0.173         0.00        0.00         0.00       0.00      0.01        0.00 0.259         0.00        0.00         0.00       0.00      0.01        0.00 0.345         0.00        0. 00        0.00       0.00      0.00        0.00 0.431         0.00         0.00        0.00       0.00       0.01       0.00 0.518         0.00         0. 00       0.00       0.00       0. 00      0.00 0.604         0.00         0. 00       0.00       0.00       0.00       0.00 0.690         0.00         0.00        0.00       0.00       0.00       0.00 0.776         0.00         0.00        0.00       0.00       0.00       0.00 0.863         0.00         0.00        0.00       0.00       0.00       0.00 0.949         0.00         0.00        0.00       0. 00      0.00       0.00 1.035         0.02         0.00        0.00       0.00       0.00       0 .00 1.121         0.02         0.00        0.00       0.00       0.00       0.00 1.208         0.00         0.00        0.00       0.00       0.00       0.00 1.294         0. 00        0.00        0.00       0.00       0.00       0.00 1.380         0.00         0. 00       0.00       0.00       0.00       0.00 1.466         0.00         0.00        0.00       0.00       0.00       0.00 1.553         0.00         0.00        0.00       0.00       0.00       0.00 1.639         0.00         0.00        0.00       0.00       0.00       0.00 1.725         0.00         0.00        0.00       0 .00      0.00       0.00 1.811         0.00         0.00        0.00       0.00       0.00       0.00 1.898         0.00         0.00        0.00       0.00       0.00       0.00

1. 984 0.00 0.00 0.00 0.00 0.00 0.00 2 .070 0.00 0.00 0.00 0.00 0.00 0.00 2.157 0.00 0.00 0.00 0.00 0.00 0.00 2.243 0.00 0.00 0.00 0.00 0.00 0.00 2.329 0.00 0.00 0.00 0.00 0.00 0.00 2.415 0.00 0.00 0.00 0.00 0.00 0.00 TOTALS 0.05 0.00 0.00 0.00 0.03 0.00 DATE: 27-Jun-2007 TIME: 10:38:00 WCAP- 16168-NP-A June 2008 Revision 2

M- 12 M-2: ISI Every 10 Years

• WELCOME TO FAVOR *
• FRACTURE ANALYSIS OF VESSELS: OAK RIDGE *
• VERSION 06.1 *
• FAVPOST MODULE: POSTPROCESSOR MODULE *
• COMBINES TRANSIENT INITIAITING FREQUENCIES *
• WITH RESULTS OF PFM ANALYSIS *
• PROBLEMS OR QUESTIONS REGARDING FAVOR *
• SHOULD BE DIRECTED TO *
• TERRY DICKSON
• OAK RIDGE NATIONAL LABORATORY *
• e-mail: dicksontl@ornl.gov *
• This computer program was prepared as an account of *
• work sponsored by the United States Government *
• Neither the United States, nor the United States *
• Department of Energy, nor the United States Nuclear *
• Regulatory Commission, nor any of their employees, *
• nor any of their contractors, subcontractors, or their *
• employees, makes any warranty, expressed or implied, or *
• assumes any legal liability or responsibility for the *
• accuracy, completeness, or usefulness of any *
• information, apparatus, product, or process disclosed, *
• or represents that its use would not infringe *
• privately-owned rights.
• DATE: 27-Jun-2007 TIME: 10:36:09 Begin echo of FAVPost input data deck 10:36:09 27-Jun-2007 End echo of FAVPost input data deck 10:36:09 27-Jun-2007 FAVPOST INPUT FILE NAME = postoc55.in FAVPFM OUTPUT FILE CONTAINING PFMI ARRAY = INITIATE.DAT FAVPFM OUTPUT FILE CONTAINING PFMF ARRAY = FAILURE.DAT FAVPOST OUTPUT FILE NAME = ocpostlOyrint.out WCAP- 16168-NP-A June 2008 Revision 2

M-13 M-2: ISI Every 10 Years (cont.)

• NUMBER OF SIMULATIONS = 42000
• CONDITIONAL PROBABILITY CONDITIONAL PROBABILITY OF INITIATION CPI=P(IIE) OF FAILURE CPF=P(FIE)

MEAN 95th % 99th % MEAN 95th %9 9th % RATIO TRANSIENT NUMBER CPI CPI CPI CPF CPF CPF CPFmn/CPImn I. ------------------------------------- I ----- I 8 C).OOOOE+/-00 0.0000E+00 0.OOOOE+00 0.0000E+00 0.OOOOE+00 0.0000E+00 0.0000 12 C).0000E+00 0.OOOOE+00 0.OOOOE+00 0.0000E+00 0.OOOOE+00 0.0000E+00 0.0000 0.OOOOE+00 3.6598E-12 0.OOOOE+00 0.0000E+00 0.0001 15 . 9014E-08 0.OOOOE+00 0.OOOOE+00 9.3807E-06 2.6045E-08 0.OOOOE+00 0.OOOOE+00 0.0084 27 3 .0898E-06 C *OOOOE+00 0.OOOOE+00 0.0000E+00 0.0000E+00 0.0000E+00 0.OOOOE+00 0.0000 28 29 C).OOOOE+00 0.0000E+00 0.OOOOE+00 0.0000E+00 0.0000E+00 0 .OOOOE+00 0.0000 C).OOOOE+00 0.0000E+00 0. 0000E+00 0.0000E+00 0.0000E+00 0 .0000E+00 0.0000 30 31 C). 000E+00 0.0000E+00 0. 0000E+00 0.0000E+00 0.O0005E+00 0 .OOOOE+00 0.0000 C).OOOOE+/-00 0.0000E+00 0. 0000E+00 0.OOOOE+00 0.0000E+00 0 .OOOOE+00 0.0000 36 0.0000E+00 0. 0000E+00 0.0000E+00 0.0000E+00 0 OOOOE+00 0.0000 37 C *0000E+00 0.OOOOE+00 0 OOOOE+00 0.0000E+00 0.OOOOE+00 0 .OOOOE+00 0.0000 38 C *OOOOE+00 44 4 .4385E-06 0 .0000E+00 3. 1287E-06 3.3673E-06 0.0000E+00 1.6329E-06 0.7586 89 C *OOOOE+00 0. 0000E+00 0 OOOOE+00 0.OOOOE+00 0.OOOOE+00 0. 0000E+00 0.0000 90 C *0000E+00 0 OOOOE+00 0 .OOOOE+00 0.OOOOE+00 0.0000E+00 0. 0000E+00 0.0000 98 C *OOOOE+00 0 OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.0000E+00 0. 0000E+00 0.0000 4 .3209E-07 0. 0000E+00 0.OOOOE+00 4.2164E-08 0.OOOOE+00 0 OOOOE+00 0.0976 99 100 1 .6582E-06 0. 0000E+00 4.6005E-07 6.5525E-07 0.0000E+00 0.OOOOE+00 0.3952 June 2008 WCAP-16168-NP-A June 2008 Revision 2

M-14 M-2: ISI Every 10 Years (cont.) 101 9.7863E-06 0.OOOOE+00 5.0383E-05 1.9741E-08 0.0000E+00 0.OOOOE+00 0.0020 102 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.0000 109 1. 1900E-07 0.OOOOE+00 0.OOOOE+00 7.7323E-08 0.OOOOE+00 0.OOOOE+00 0.6498 110 2.6087E-03 6.2367E-03 3.2743E-02 1.7303E-05 2 .1472E-04 3. 0001E-04 0.0066 ill 1.4966E-08 0.OOOOE+00 0.OOOOE+00 5.2442E-11 0 .OOOOE+/-00 0.OOOOE+00 0.0035 112 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 o.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.0000 113 6.2055E-08 0.OOOOE+00 0.OOOOE+00 5.9294E-08 0 .OOOOE+00 0.OOOOE+00 0. 9555 114 0.0000E+00 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0 OOOOE+00 0.OOOOE+00 0.0000 115 0.OOOOE+00 0 OOOOE+00 0.OOOOE+00 0.OOOOE+00 0 OOOOE+00 0.OOOOE+00 0.0000 116 2.2291E-06 0 .OOOOE+00 7.0608E-07 5.1321E-09 0.000OE+00 0.OOOOE+00 0.0023 117 5. 5617E-05 3 .4579E-04 7.7732E-04 9.5098E-08 0 OOOOE+00 0.OOOOE+00 0.0017 119 9. 1956E-05 3. 0210E-04 1.5073E-03 7.8504E-07 2. 9569E-05 3 . 7212E-06 0.0085 120 3.1952E-05 2. 8404E-04 4.5357E-04 2. 5547E-05 2. 6061E-04 3.1323E-04 0.7995 121 2.7485E-08 0 OOOOE+00 0.OOOOE+00 7. 2283E-10 0 .OOOOE+00 0.OOOOE+00 0.0263 122 3.8697E-04 2. 6683E-03 3.8068E-03 3. 8690E-04 2. 6683E-03 3.8068E-03 0.9998 123 2.7485E-08 0.OOOOE+00 0.OOOOE+00 7 .2283E-10 0.OOOOE+00 0.OOOOE+00 0.0263 124 1. 7117E-04 8.5298E-04 2. 3535E-03 1. 6870E-04 8.5298E-04 2.3226E-03 0.9856 125 7.0998E-05 3. 0103E-04 1. 153 9E-03 3. 7122E-07 0.OOOOE+00 7.3397E-07 0. 0052 126 8. 9419E-07 0 OOOOE+00 1. 1998E-07 3.0547E-09 0.OOOOE+00 0.OOOOE+00 0 .0034 127 0 OOOOE+00 0 OOOOE+00 0 OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.0000 141 1. 2407E-04 3. 1186E-04 2 .2591E-03 2 . 6331E-06 4.3366E-05 2.3103E-05 0 .0212 142 4. 5517E-11 0 OOOOE+00 0 .OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0. 0000 145 0 OOOOE+00 0 OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0 OOOOE+00 0.0000 146 2. 2255E-06 0 OOOOE+00 7 .2719E-07 1.8547E-07 0.OOOOE+00 0 .OOOOE+00 0.0833 147 0 OOOOE+00 0 OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0 OOOOE+00 0.0000 148 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.000OE+00 0 OOOOE+00 0.0000 149 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0 OOOOE+00 0. 0000E+00 0.0000 154 1.4390E-04 4.9507E-04 2.3339E-03 3.8809E-07 0 OOOOE+00 0 OOOOE+00 0.0027 156 2.2374E-02 5.3757E-02 1.8079E-01 8.5646E-05 4. 1552E-04 1. 2331E-03 0.0038 160 1.3434E-02 2.9083E-02 1.0573E-01 1.7218E-04 5. 9581E-04 2. 6851E-03 0.0128 164 1.4231E-02 3.3616E-02 1.2566E-01 8.7527E-05 5 .4722E-04 1 .2941E-03 0 .0062 165 3.5089E-04 2.6865E-03 2.7402E-03 3.5077E-04 2. 6865E-03 2. 7402E-03 0.9997 168 2.0794E-04 9.4460E-04 2.8225E-03 2.0523E-04 9. 4460E-04 2. 8108E-03 0.9870 169 2. 0891E-04 5.2788E-04 3.5882E-03 7. 6133E-06 1 .4604E-04 8.9295E-05 0.0364 170 1.7072E-08 0.OOOOE+00 0.OOOOE+00 3.3194E-10 0 .OOOOE+00 0.OOOOE+00 0.0194 171 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0.OOOOE+00 0 .OOOOE+00 0.OOOOE+00 0.0000 172 8.2142E-05 3.0025E-04 1. 3571E-03 4.2000E-07 0.OOOOE+00 9.8847E-07 0.0051 WCAP- 16168-NP-A JuRe 2008 Revision 2

M- 15 M-2: ISI Every 10 Years (cont.) 178 1.4390E-04 4.9507E-04 2.3339E-03 3.8804E-07 0.OOOOE+00 0.0000E+00 0.0027 NOTES: CPI IS CONDITIONAL PROBABILITY OF CRACK INITIATION, P(ItE) CPF IS CONDITIONAL PROBABILITY OF TWC FAILURE, P(FIE) WCAP-16168-NP-A June 2008 Revision 2

M-16 M-2: ISI Every 10 Years (cont.)

• PROBABILITY DISTRIBUTION FUNCTION (HISTOGRAM) *
• FOR THE FREQUENCY OF CRACK INITIATION
• FREQUENCY OF RELATIVE CUMULATIVE CRACK INITIATION DENSITY DISTRIBUTION (PER REACTOR-OPERATING YEAR) (%) (%)

0.OOOOE+00 2.8976 2.8976 2.1486E-06 92.1000 94.9976 6.4459E-06 2.9714 97.9690 1.0743E-05 0.8738 98.8429 1.5040E-05 0.3714 99.2143 1.9338E-05 0.2548 99.4690 2.3635E-05 0.1476 99.6167 2.7932E-05 0.1000 99.7167 3.2230E-05 0.0643 99.7810 3.6527E-05 0.0381 99.8190 4.0824E-05 0.0262 99.8452 4.5121E-05 0.0262 99.8714 4.9419E-05 0.0143 99.8857 5.3716E-05 0.0262 99.9119 5.8013E-05 0.0190 99.9310 6.2311E-05 0.0119 99.9429 6.6608E-05 0.0095 99.9524 7.0905E-05 0.0071 99.9595 7.5202E-05 0.0048 99.9643 7.9500E-05 0.0024 99.9667 8.3797E-05 0.0048 99.9714 9.2392E-05 0.0048 99.9762 1.0099E-04 0.0024 99.9786 1.0528E-04 0.0048 99.9833 1.0958E-04 0.0071 99.9905 1.1818E-04 0.0024 99.9929 1.4826E-04 0.0048 99.9976 1.6115E-04 0.0024 100.0000

                            ==       Summary Descriptive    Statistics            ==

Minimum = 0.OOOOE+00 Maximum = 1.6165E-04 Range = 1.6165E-04 Number of Simulations = 42000 5th Percentile = 3.8070E-12 Median = 1.0893E-07 95.0th Percentile = 2.1486E-06 WCAP- 16168-NP-A June 2008 Revision 2

M- 17 M-2: ISI Every 10 Years (cont.) 99.0th Percentile = 1.2561E-05 99.9th Percentile = 5.1763E-05 Mean = 1.0085E-06 Standard Deviation = 3.8747E-06 Standard Error = 1.8907E-08 Variance (unbiased) 1.5013E-I1 Variance (biased) 1.5013E-11 Moment Coeff. of Skewness = 1.4853E+01 Pearson's 2nd Coeff. of Skewness = 7.8082E-01 Kurtosis = 3.6493E+02

• PROBABILITY DISTRIBUTION FUNCTION (HISTOGRAM) *
• FOR THROUGH-WALL CRACKING FREQUENCY (FAILURE)
• FREQUENCY OF RELATIVE CUMULATIVE TWC FAILURES DENSITY DISTRIBUTION (PER REACTOR-OPERATING YEAR) (%) (%)

0.OOOOE+00 22.6333 22.6333 9.0774E-08 74%.9690 97.6024 2.7232E-07 1.1857 98.7881 4.5387E-07 0.4024 99.1905 6.3542E-07 0.1905 99.3810 8.1696E-07 0.1286 99.5095

9. 9851E-07 0.0810 99.5905 1.1801E-06 0.0952 99.6857 1.3616E-06 0.0405 99.7262 1.5432E-06 0.0143 99.7405 1.7247E-06 0.0167 99.7571 1.9062E-06 0.0167 99.7738

2. 0878E-06 0.0238 99.7976 2 .2693E-06 0.0167 99.8143 2 .4509E-06 0.0143 99.8286

2. 6324E-06 0.0310 99.8595

2. 8140E-06 0.0071 99.8667

2. 9955E-06 0.0048 99.8714

3. 1771E-06 0.0071 99.8786

3. 3586E-06 0.0214 99.9000 3.5402E-06 0. 0167 99.9167

3. 7217E-06 0. 0095 99.9262 4.0848E-06 0.0119 99.9381 4.2664E-06 0. 0048 99.9429 4.4479E-06 0. 0024 99.9452 4.6295E-06 0.0095 99.9548 4.8110E-06 0.0095 99.9643 6.2634E-06 0. 0024 99.9667 6.4449E-06 0.0024 99.9690 WCAP- 16168-NP-A June 2008 Revision 2

M-18 M-2: ISI Every 10 Years (cont.)

6. 8080E-06 0.0048 99.9738

6. 9896E-06 0.0024 99.9762

7. 1711E-06 0.0024 99.9786

7. 3527E-06 0. 0024 99. 9810

7. 8973E-06 0 .0024 99.9833 8 .4419E-06 0 .0024 99.9857

1. 0076E-05 0 .0024 99.9881

1. 0802E-05 0. 0048 99.9929

1. 7156E-05 0. 0024 99.9952

1. 7882E-05 0.0024 99.9976

1. 8064E-05 0 .0024 100.0000

                           ==          Summary Descriptive  Statistics            ==

Minimum = 0.OOOOE+00 Maximum = 1.7973E-05 Range = 1.7973E-05 Number of Simulations = 42000 5th Percentile = 0.OOOOE+00 Median = 8.9093E-11 95.0th Percentile = 9.0774E-08 99.0th Percentile = 3.6793E-07 99.9th Percentile = 3.3586E-06 Mean = 2.6200E-08 Standard Deviation = 2.6170E-07 Standard Error = 1.2770E-09 Variance (unbiased) = 6.8486E-14 Variance (biased) = 6.8485E-14 Moment Coeff. of Skewness = 3.5238E+01 Pearson's 2nd Coeff. of Skewness =-9.4834E-01 Kurtosis = 1.8441E+03

• FRACTIONALIZATION OF FREQUENCY OF CRACK INITIATION *
• AND THROUGH-WALL CRACKING FREQUENCY (FAILURE) - *
• WEIGHTED BY TRANSIENT INITIATING FREQUENCIES *

                                            % of total              % of total frequency of           frequency of crack initiation       of TWC failure 8              0.00                       0.00 12               0.00                       0.00 15               0.00                       0.00 27               0.00                       0.00 28               0.00                        0.00 June 2008 WCAP- 16168-NP-A WCAP-16168-NP-A                                                                       June 2008 Revision,2

M-19 M-2: ISI Every 10 Years (cont.) 29 0.00 0.00 30 0.00 0.00 31 0.00 0.00 36 0.00 0. 00 37 0.00 0.00 38 0.00 0.00 44 0.00 0.01 89 0.00 0.00 90 0.00 0.00 98 0.00 0.00 99 0.00 0.00 100 0.00 0.00 I01 0.00 0.00 102 0.00 0.00 109 0.00 0.02 110 1.06 0.30 1il 0.00 0.00 112 0.00 0. 00 113 0.00 0.01 114 0.00 0.00 115 0.00 0.00 116 0.00 0.00 117 0.01 0.00 119 0.01 0.00 120 0.00 0 .01 121 0.00 0.00 122 0.96 36.96 123 0.00 0.00 124 0 .44 16.78 125 0.00 0.00 126 0.00 0.00 127 0.00 0.00 141 1.43 1 .04 142 0.00 0.00 145 0.00 0. 00 146 0.03 0.12 147 0.00 0.00 148 0.00 0.00 149 0.00 0.00 154 2.13 0.22 156 21.66 3.30 160 30.94 14.65 164 38.99 9.54 165 0.24 9.06 168 0.17 6.32 169 0.60 1.45 170 0.00 0.00 171 0.00 0.00 172 0.99 0.20 178 0.36 0.02 TOTALS 100.00 100.00 DATE: 27-Jun-2007 TIME: 10:36:45 WCAP- 16168-NP-A June 2008 Revision 2

N-1 APPENDIX N RESPONSES TO THE NRC REQUEST FOR ADDITIONAL INFORMATION (RAI) REGARDING THE REVIEW OF WCAP-16168-NP, REVISION 1 WCAP- 16168-NP-A June 2008 Revision 2

N-2 Program Management Office 20 International Drive e t6G Windsor, Connecticut 06095 October 16, 2007 WCAP-16168-NP, Rev. 1 Project Number 694 OG-07-455 U.S. Nuclear Regulatory Commission Document Control Desk Washington DC 20555-0001 Subject. Pressurized Water Reactor Owners Group Responses to the NRC Request for Additional Information (RAI) on PWR Owners Group (PWROG) WCAP-16168-NP. Rev. I "Risk-Informed Extension of Reactor Vessel In-Service Inspection interval" (TAC NO. MC9768) MUHP 50971509815099 Task 200812059

References:

1. WOG Letter from Ted Schiffley to Document Control Desk, Request for Review and Approval of WCAP-16168-NP Rev. 1, entitled "Risk-Informed Extension of Reactor Vessel In-Service Inspection Interval," dated January 2006, WOG-05-25, January 26, 2006.

2. Acceptance for Review of Westinghouse Owners Group (WOG) Topical Report WCAP-16168-NP, Rev. I "Risk-Informed Extension of Reactor Vessel In-Service Inspection Interval" (TAC NO. MC9768) MUHP 5097/5098/5099 Task 2008/2059, OG-06-3 1i, September 22, 2006.

3. NRC emails from Sean E. Peters of NRR to Tom Laubbam of PWROG dated March 9 and 12, 2007 "RAIs for WCAP-16168".

In January 2006, the WOG, now known as the Pressurized Water Reactor Owners Group (PWROG), submitted WCAP-16168-NP Rev. .1, entitled "Risk-Informed Extension of Reactor Vessel In-Service Inspection Interval," for review and approval (Reference 1). In September 2006, the NRC accepted the topical report (Reference 2) and provided an informal Request for Additional Information (RAI) (Reference 3) on March 9 and 12, 2007. Enclosure I to this letter provides the RAI responses to the questions received in Reference 3. Enclosure 2 is the marked-up WCAP. WCAP- 16168-NP-A June 2008 Revision 2

N-3 U.S. Nuclear Regulatory Commission October 16, 2007 Document Control Desk Page 2 of 2 Washington DC 20555-0001 OG-.07-455 1f you have any questions, please do not hesitate to contact me at (630) 657-3897, or if you require further information, please contact Mr. Jim Molkenthin of the PWR Owners Group Pni*ect Management Office at (860) 73 t-6727. Regards. H _7.-' ........... / Frederick P. "Ted" Sehiffley. It, Chairman PWR Owners Group FPS:JPM:Ias

Enclosures:

(2) cc: M. Mitchell, USNRC B. Bishop, W S. Peters, USNRC J. Andrachek, W T. Mensah, USNRC J. Carlson, W S. Rosenberg. USNRC PWROG MSC Participants in the RV ISI Program-C. Brinkman,. W PWROG Management Participants in the RV ISI Program C. Boggess, W PWROG PMO N. Palm, W WCAP-16168-NP-A June 2008 Revision 2

N-4 RtLOUST FOR ADDIITONIAL INFORMNIAION T OA1Y 1 "FR01\',1(NUC AR RI'EACT(IORRI AJ I,.ATIION AND PWR.OG RESPONSES RE: PRESSURIZED A<\XItR RFA( 1kR OWNE.RS GROUP TOPICAL REMORt (TR) WCAP t61.68-N. RIEVISION 1. "RISK-INIFORMI.D IPTENSION OF THE REACTiOR :VElSSEL IN-SERVIC INSPICTION INTERVA"U" PRESSIURIZED WATER REACTOR OWNERS GROLP PRO,:JECT NO. 694 Materials 1 Section 3.2 ofWCAP-6t668-NP Revision 1. indicates that tlihepilot-plant studies included a probabilistic representation of the fatigue crack growth correlation for Ic-rritii materials in water consistent with die current models 'containedin .Appwdidi-A of Ariaerican Societ*, of mroviousand

        .Mutianical Enineers Boiler and Pressure Vessel Code (ASMIE (Code) Section:X. "llie probabilisic repres.,ntation was consistent with those used in the pelIiSi           6iode.:and NRC-approted SRRA todl for piping risk infonned inserviec inspection, In Appendix A of di.tqNR(:siaff sally evaltuition (SE) on WCAP-14572. Rewision 1, "Westinghouse Owners Group.Application of Risk-Informed Methods to Pipinig nservic Inspection ~opical Report;. the stafifconeuiied thatthe SRRA code addresses fatigue.

crackgrowth in an acceptable manner since it is consistent with the technical aipproacih usedb:ý other statcof-&the-art clodes for probabilisticfitaetuire'mechanies. Thliestalf noted that realistiepredictions.of failIure probabilities require that the user define input parametersI wIhich accurately representall sources of fatigue.stress and theprobability, forpreexisting fabrication erail*sin '?seld. a) The staff requests th~at the ,\estinlgioute Ownert GroUp (\VOG) pro r6ide the transients and nuisber of transients that wevere assume d in the analysis: of tie pilot-plant studi.s a.nd explain why the proposcd transients~represent all the sources of fatigue stfess. Response: For the Westin houseand CENuelearSteam Supply System (NSSS) plant designs, all of (fhe Reactor Coolant System (RCS) design basis transients were considered in the .analysis. These RCS.design basis transients (lhrein referred to as,.transients unless specifically noted) are identified:in tdie plant final safety analysis reports. The PROBSBFMD program, which was used to, modify the FAVOR surface breaking flaw density input file (S.dat): rtequires d. treiqueney (cycles per year) of the transient that prodiiucesthe most craek growth and could represent the fatigue crack-growth of ýall the transients in the reactor vessel design iasis. lypic.ially, fast transients with high temperaturc spikes produce. high skin sti esses which are of concern for initiation but do not provide sufficitent energy to grow an-existing crack. TheSIASME Code requiiresthat the fatigue usage factor for all design basis'transients be less than one. Therefore, provided a plant remains Within it's design basis transieit paramete's and number of cycles', tile design bisis tr-iiisicnttsshould not initiate a craeký Slow transients; where the therimal stresses, arc

                  ýallowed to fully develop through the reactor vessel wallh.such as.heatup and cooldown.are of WCAP- 16168-NP-A                                                                                                            June 2008 Revision! 2

N-5 much more concern for fatigie crack growth. For this reason the primary transient chosen to be: evaluated with PROItS13FI for the WXestinghousce CT, and B&W pilot plants was the cooldown transient. While the fatiguc ewcle consists of hcatup and eooldown, the cooldown portion is used to calculate the chlioge in stress and stress intemsity factor because i! is the portion that results in tensile slresss on the inside of the vessel wall, For eacli pilot plant, the cooldown triansient that was.mevaluhlalt consisted or a 1001F/hour decrease in tlemperaiture rom full operating temperature to ambient temperatsnre, For the CE and B&W pilot plants this decrease in temperature Was coeiicident with a do-r-se iii pressure from normalxoperating pressure to atmospheric pressure. For tie Weestinghouse pilO plantprpssure Was reduced at a rite of approximiately 700 psi per hour starting at the time cooldown is initiated. tIese 0ooldussn cuires asieeloisistent with design basis cooldown curkse for.theXWustinghouse and CE pilot plant desi*ns. B&W desitun basis data was not available so the B&W pilot plant cooldoIwn transient wa's assumed to he:coimlparable to the cooldown transient for the CE design. Atler choosina the conoklwn transient as the representative tiansient to be evaluated using the PIROBS31:) co0ded. it War necessary to determine a number of.wcoldoiwn transients tb at ixould envelope the fhti 'ueera'k growith of all ol the design basis transient.s. For the Westinghouse desi-gTs, pres ious.Iaiiue crack Prowth analyses of flaws on the inside surface of the reactor vessel had showvn thiiat .br-all of the desivpi.basis transients, bnly the following design basis transients irsulted in measurable crack growth:

• Heatup'(.;ooldo.Ni n
• lPressure "Iests
• Feedwater .Cyetng
• iiadvy.itent Deprcssuriton IleatupiCooldown and Pressure Tests are,a common contributor for all NSSS designs

            ,(W'estingho use. CE, and B1W).. For dlie Westinghotuse NSSS desions Feedwater Cyclingg and.

Inadvertant Dcpressuri7atiOn may become domifiant eiter as a result of die original plant specific des*n basis*oadini analysis or for uprating considerations. 'able la provides a list of transients that were~considered for-the Westinghbuse-NSSS design in iddressing fatigue crack erowth. A description oft*ih four tiansients that were determined to contribute to crack girowth is provided, iri tabl lb. Individual plant RCS design specifications pros ide additional detail on the transients.. Table la: Westinghouse NSSS Design Transients, Transient I Transient Ptant:HeatupiCooldown-10OFihr Smnall Loss of Coolant Accident Pressure Test 3125.psia/250 p-sia Small Steam Break Feedwater Cycling I Complete LosS of Flow lnadv-ertcnt.Deprcssurization I Feedwater Line Break Unit Loading and Unloading Between 0 and Reactor Coolant Pipe Break. 15% of Full Pows_ __r: Plant Loadinhl Unloading at 5% of flail power Large Steam Line Break

               .,per minute                                           *,.

Step Load Increase. Decrease of. 10% of Full I Reactor Ct olrnt Pump LOcked*Rotor PowerColtPupLceRtr

                ,rS      tLadDeerea swithsteam dump). I Control Rod Ejection WCAP-16168-NP-A                                                                                                       June 2008 Revision 2

N-6 ______ T'able Ia: \Vestmchouse NSSS3 Desiun Transients t Irisient I rnin i ..* . -;I-_ G...............

                         *.-       .........                  .*.ii . ..
                                                                         ..       *ri ....

I..

                                                                                        .... f~ .......

_.... 7 7 . ...... .................................. I nap Out ol ii~ieni \ouinall.'Onp.Shutdown! Iurbine Roll I c.st Loss of Load_ _ _Refucnin_ Loss of Flow j Exce.sie Feedwater Iltow Rctrr fi-i. from fuull pwer I Inadvtltent Auxiliary Sprav

                       *..y

.*!.* *! n, i ]. *.. . '..*..*......

l .. ....................... ..,* t .;!........... _!.*t ........................................................................ Control Rod D _o_ \ Accumulator Injection Break lInadyoTtant Saficy Injuetion Actuation [ Stuadx Stat* FIluctuationLs Table lb: Westinihouse NSSS Design Transients Contributing to Crack Growth rnieli. Description, Plant teaiup'CooWn-iOOTlhr D1esisn hel*itupicooldown transients are eonsenvatively rcprtýeated by continuou.S operations petfoined at a. uniforn tniperattire ratite. he heatupu:onsidered going 60m. ambicn't turmpadriur and pr*s*teon cndition to the no-load temperatue and pressure condition. fhlie C:01ldown considerI scing ngfhomft no- oad temperature and pressure conditions to amnbient temperature and pressure conditions P'esaure Test.3125 psia.-2250 psia j-"hc pressure tests .include both shop and field hydrostatic tests that. occur as:a result of component and system __:testing. Feedwater Cycling 'ffis transient addresses intermittent fluctuations in leddwatcr temperature that cause the reactor coolant average.iemperature to dccruasseio a lower value and thenrietum to nioload conditions-Inadvertent Depressurization iSevteral events can be postulhtdedto occur during normal plant .operation Which ssilt cause rapid dipressurwation ot the reactor coolanctsystum Of ihse. thliepr-ssurizer safety valve actuation causes the most severe transicnt and is eommonly used cis an umbrclla case to consentatiyely represent the impact on dile system.r'om any of the inaidvertent depoiessiri7atioh evenits, Existing analyses of tlese transients had been:peiformed using:a.O% through-wall initial flaW. Therefore; senitivitv studies were performed on the,four contiib.uting transients using the PROBSBFD Code with, aninitial flawdepth equivalent toathe thickness of the. cladding (then iru.nded up to the nearest wholepe-rcent of the wall thickness). The analysisýshowed that tie only design basis transient that resulted in significant crack growth was the cooldown transient, The sensitivity study using the PROFsB 'D .indicated that ihi flaw growth contribution of the FeedwaterCveling. and Inadvertant Depressuiization transients wasiat least an order of magnitude less than the contr-ibution fromn the hcatup"cooldown transient. Pressure test transienis werec

             .*enseloped by thcheatup/cooldown transient To envelope tIe contribution of the Feedwater Cycling and iadcittent Depressuri~ation tranisients and any partial.cooldowns, 2 add itional WCAP-16168-NP-A                                                                                                                                June 2008 Revision 2

N-7 cooldown transients per year were conservati1vcly added to the design basis of 5 cooldown cycles per year. 'Therefore, 7 cooldown cycles per year wereceuated cy with PROBSlMFD to determine the so'lhce breakingflaw density for the Westinghonse NSSS design pilot plant Prcviouis 'tigue crack growth studies were not available for the CE NSSS designs and therefbre, all design basis transients were evaluated using the.PROBSBFD code: Plant Heatup and Cooldown Plant Loading amd Unloading at 5% mii 0* 10 %'Stc) *Lad Increase ani.DDecrcase Reactor Trlip, Loss ofilows and Loss of Load LoMs of Sondar* Pressure Hydrostatitc lest 3125 psia12250 psia

                      ',afety V\avc Relief Table 2 provides a lisi. of transients Ithat were considered for the CE-NSSS design in addressing fatigue crack rowth, including~a description of thetransientis Individual plant RCS design specifieations provide additional detait on the transientcs, Table 2CE NSSS Dusign T1rnsients-Transient.                                               )escption Plant lltatup/Cooldowxn-lO0W*,ir            Design heatup/cooldown itransients are conservatively rpresented by (.ontinuous operations pcrforned4at it uniormi temperature rate. I1 ceheatupconsidered going from ambient temperantie and piessure ckonditioh to the
                                                          -no-load temperature and pressure conditionm The cooldowtn considers, oing from the nioload temperatu'e:

and pressurn conditions to ambient tIcmnpcraturc and __________ rcs~ure. conditionsl Pressure Test 3125 psia/2250 psia tlhe pressure tests include both shop and field'hvdrostatiie teststhi t occur as a result of componentand system testuni. Plant Loadingi Unloading 5%'nmin The unit loading and unloadingcasesare conservatively rep:presentud bya continuous and uniform ramp:p6we* Change of 5 percent per minute betveen 15 percent load. and full load. 'ltis is the maximum possiblexrate% consistent with operation under automatic reactor control. 10% Step Load Increase! Decrease The 160%Step load increase/decrease itrasient is a uwhich is assumed to be a change in turbine control value Reaitor 'Trip.Loss oflw; I.,Thee. include reactor trips due to a number f* Loss of Load circumstiances over the lifelof the plant. LoAssof SecorndaryPtressure A.reaetorbtip will occur. as a result6f thieloss of

                                                           *.secondary side pressure:.

Safety Val;e Relief Several cvcntis can be postulated to occur during normal plant'operation which Will cause rapid depressurization of the reactor 0oolarit system. Of thcse. the pressurizer WCAP- 16168-NP-A June 2008 Revision 2

N-8 K Table2: Cl NSSS Desim Transient~s Transient ... I)escrlptin saety vadve actuation. causes thic most severe transient and is commonly used is an umbrella case to conscrativetv represent the impact on the system from

                                           ....................... .. an- of the other transients.         "            "

Cortsistnt With the Westinghouse design. the eooldowii trainsient produced the largest amount of fitfigue crack growh.th ']he loss of secondary pressure transient also produced measurable growth. Ilowever. the 12 eooidowns per year was considered to be conservative in comparison to the actuatnambe4r of eooldowNvs a plant miglt expcresce. in a given yearlof Operation. [lherefore, to envelope the contribution of the loss of secondary, pressure transicLtL only 1 additional cooldtownl transient was added to the design basis o( 12 cooldowns pri y ear! thus resulting in 13 cooldowns per year being evaluated with PROBSI3FI) to dletnrine the suriace breaking flaw density ibrthe C.. design pilot plant; For the B&W design twelsevc cooldown transients a vear were assumed and evaluated with PROBSI*IL) to determine the surface bre*king flaw dunsity. Ps stated in thie \WCMfiorra B&W plant to apply the intervas etension of this WCAP, it wouldtha-vis to be dem0otnstrated thait the 12 cooldo01w transients pe crenvelope the fatigue crack gyrowth Ironi;all of the design basis

transients.

b) The staffrequests that the \V,001,identify the initial flaw size, location and density assumed in the pilot. plaint fatigue emack growth inilysis and the basis totrthe initiad flaw sine distribution ond ldensity. Identify and provide anatysesiof all insetrvice inspection results and destructive test results that were used to determine the iitial flaW size,.location and density assumed in the pilot plani fatigue crack growti analysis. Resp6nse: In Revision I of WCAP-16168-NrP, the initial flaw distributions, including the surface breaking flaw distribution used for the'fatigue crack growth analysis., are discussed on pages 3-8 and 3-9 of section 3.2. [he distiributiorswfor the three repr6sentativxe plants wcei all generated usinp-the computer code VFLAW\03 developed by PNLL as described in Revision I of NUIEG,*R-6817, A GeneralizedPrtxrvchthre'firrGenautingI la- J ~latcdtnputs]Jir lhe PA I*,11? Code, 206. the txhnical bases for the surfice breakiii flaW distributions at6 described in Scction 8 of ihis report and the application of VFLAW03 Computer Code tfo surface-breaking flaws in single-pass cladding is decribed in Section 9.6 (pages 9.24 to 9.26). Figure9. 17 of this report provides thed input surface breaking flaw distribution for the Oeonee Unit I vcssel that can be compared with the input to the,PROIOSI31D Computer Program in Sections K-I and K-2 in Appendix K Of the WCAP Repo**t. Input -variable 1 (FIFDepth) gives the fiactiil uintial flaw depth as'003 ,,which corresponds to tlie non-zero density in the row for N'3 (percent of watl thickness) in Figure9. 17. Input

                     .Figurevariable 2 (IFlawDen) gives the initial flaw density as 0.0036589 flaws per square footin 9A717PROBSFD            input variables 16 to 19 (Pecr'enti,Precent4) gives the percentages for the 4 values. Of aspectratio specified in input variables8 to611 (Asp6ct I-spect4) of 2, 6, 10 and 99 (infinite) as 67.450, 20.769, 3-9642 and 7.8166, rcspcctivclv" which agrees with the values in IWCAP- 16168-NP--A                                                                                                                       June 2008 Revision 2

N-9 Figure: 9.17 of the NUIlMiGCR Report. ihesame input to V14AWV03 was used, except for plant-specificvalues ol'Vessel wall thickness and cladding Ihid ncss, which were set. equal to the bead size for a singlc-pass laidding, to generate the iitial surace flaw distribution input to tPROBSBFWD t1r he-n' r alle tUnit I in S~ections (-1 and (-2 of Appcndi\ C and tfor Palisades in Sections (- and G-2 of Appendi.k Q in the WCAP Report. As indicated in the WCAP RportO. the infoiination for dalcalating cldding (surface bieakin) flaws in Tables B-2 (page B-8), F-2 (page F-)an .1-i2 (pag J- 18) is taken dircctlyi from Taiblc 4 2 oft he Dc~ember 20)02 Draifi NUREG Report on the.Technical Bnis tur.Revisiori of thi PTiS Rule (ADAMS: ML 03009)0626). Note fat the eitation for this refu ence within App*ndies B. F and J in Revision 1 of WCXA.P 16 168-NP will be chaigcd frnom 1,7] to the concct rderlence nuinber of [81, e) "liestaffl rquesst,hatthll 'WOG'identitfy the fatigue crack gr6wth eures (crack growth versus: elnge in siress intnsity factor) used in thb pilot-plant studies. Response. The fatigue crack geow i raite cquations for feritic materials, suchas the vessel wall base metal arc takern hom Section 4.22 of the iThet.wreialand U'sua IMamadlfor!jc-P!4RISS (NURFG(C< 5864, Jlyd 1*99). A*s.notedl in i is ieport, these "equations provide .iprohbedilsifc, represeitalion of the fatigue. growfli rehatirMuSip firw eriitie nimderils inllwater caontained in AjXpecdix..A of Setion XI of the.AIME Boiler and Prcsu'e Vessel Code." .Figure A 43900 2 Rcference I.atigie Crack Growth Curves for Carbon and Loss Allo, Ferritie Steels Exposed to Water Enyironnients. from Appendix'A to Section Ml in the current edition of ihe ASME. Boiler and PNess=ui Vessel Code, is also0ip ovided belows for a graphical rel*esentatioh of these. equations. It should be noted that the latigue crack growth curves in Appendi\ A of Section X\l of the ASMIE Boiler and Piessure Vessel Code hive not chluged since they were originally includcst in the 1978 Edition olSection X. :Furthenoire; therc are presently4no known plans to revise the curves in the future. R - 0.25 do x10x *W*l*"5C 0 AK < V9i d,',, 95 i 0I x Itr"'AV' Q Kl Q -exp(-6.40X+0.542C.) 0.25 <R < 0.65 dd {AWA I- Ak-

• f, fA >

A 1.02 x lo(t26.9Rk-5.,72 5 j 7

f. =-01:X1T (j.75R+6.06 WCAP- 16168-NP-A June 2008 Revision 2

N-10 I > 0465 O; =(f., 1.25 0 2 + O = exp)[01.1025R- 0.43,3625 + (0.6875R +0,370125 y",r . A*' <10

                                                  '[2:5 -k1.0 10 -7'AK- .O AM 2!12 Q ep(- 0.367 +0.817CF
              ýIn  the.above. tqqodion'., R is Kýj, I         ik.K,nm - Kj and .F is normil1y distribute~d m-~   ilt a.

mecan of 0 atnd 'standard dcta~tion ot one. T1he units on the applied stes intentsity 1'actor. K. arc ksi-Onichb) and inches pei cxd~c 6n the. crack growth rate da'dN Notecthatthc nonnallh distributed random xviloe of'ý C "'hidt ,i used tocalculaeu tt. he unccrtaiznt fiacor Qj is spcified as input xdrirble: 14 (F( (R4U,() to Ihe PRQBSB'D111 Comiputtei Progranis firs~t showsn in Sectiton C I in AppiendiN C of flue XX( A.]Report. June 2008 WCAP- 16168-NP-A WCAP-16168-NP-A Revision:ý2

N-11

                                         ......... _.=.........: = * ,* ; . *.. ........... = .: = = *.........* * .:= *=:
                                                                   .....        - i . - --!- "- * ~ °* .. ... -. .. -....-.i....-i....
                                               .......    .. i: : i : "; : *' : * ......... ....... ........ -

i= ' j i '  ! i "'f: S............... 5; ' - it*', F.., 4 R  ::.,. igue Crack Growt " Curves i o and L o=================te=ls: iExposed

                                             ...I2E :I:I:I :: 7 Y'.2
                                        ':2L.R
                                        ...                                                toWater
                                                                                           "..

• i -Environments

                                                                                                               '   .. ..... t - "        ' ' * * "' ';* " ' " {
                                                                   ................. :. .-.*...;.. ,.:=:..*L*. :: < ]:Z =Z.Z :.Z 'L :2]...............:+      !::
                                                          ....i......... ..--        . ...   ....

n.L-:L. ...................

                                                                                                                                                  . ... +...i.

Exposed to wat~erEnvironiments (From Appendix A of...... ,F "?XI..ofthe1..ASME Section --......

                                                                                                                                 - BPV     :?-t!Code)

2. In Attachmaent Fiur .qu30i-2,b 'ito the June 8, 2006 letter maerilfeerence:Ftige Crackfromrowt'ho the'WOG, does. the WOOthe. staffo indicated andL thiat under-ctlad AlloysFepm crackis thCSeeWsG Figuw in forgings Refere~nce renie so*sh*allow* .that(TchanceFaiuCFfrack for themrowgimthCurves initiating during, fodrCarbon and Loweatlyer a *severe pressurized thenaal-riticSth~eefrals sho ck (PT s) transient would be fairly,:smnall. Analyses (NUREG-1874,. "Recommended Screening Limits for Pressurized Thermal..Shock (PTS)") performed by the staff indicates that for geverePTS transientsthie through wall crack ,frequency"'(TWCF), for forg ings with unader-clad cracks are greater tha those for. axiaj

        *welds with equivalent material reference temperature. How does the staff analysis impact the WOO mnalyses and the .T.WCF pilot-plant screening criteria inrAppendix.A of WCAP-1616 8-NP,.,Revision 1?9 Response:. The statement in the;Staffs question that "the through wall crack frequency (TWCF,)

for forgings with under-clad cracks are greater than those for axial welds with equivalent material reference temperature"* is only valid for reference temperatures greater than Below this P240. temperature,,the TWCF' of forgings'is equivalent to that of plates. This is confinredby plotting the TWCF-courelations in NUREG- 18774 for plates (PL), axial Welds(A W)', forgings (FO), and WCAP- 16168-NP-A June 2008 Revision 2

N-12 under-clad cracking (UC) on the same graph as shown helow. Relationship Between Max RT and TWCF 1.60lE-07 I.OOE- i I -.-- TWO 95-AWVV 55

                                                                                             ---      TWFT9`LFO LL                                                           -aý-      1V.CU95-LC
                                                                                                       ,   9VCO5-Pt 1.00E-13         0 1      2. 1.         3.4.

UOXE-19 1 .ODE.-21 ~ ~ 170 180 190 200 2110220 230 240 250 RT MAX (F) The correlations from NUREG-1 874 that were used to produce.this graph are as follows:

                                      .'exp{ 5..519h'tilnT;,*A;,-616) - 40:.542ýj TWVCF*:,,,* =expj2.3:737ln(RT,,,%j 4 ._- *.,0O).. 162,38NT"r1fl._ x l(t.....              . r fl(Th.

correlation is fir frgings with underclad cracking; rl91) TWCFUAT,,,, -{1:3 x !I)30I'"lfl(This is the underclad-cracking portion of the correlation a-ove. Without this underclad cracking portion.the forging-is equivalent to a plate) TWC*

                                     =*xp* 23737 a( T** z *        -   ,0-1"62.38)}

t For the graph,-Pwa:s ch6sei to have a yaluc of'*i".eorresponding to a reictor vcssel beltlinc walI thickness:of less than 9.5 inches per NUREG-1874, While. this selection is appropriate because most US PWNRs have a reactor vessel beltline wall thickness less than 9.5- for thicker vessels the conclusions discussed in thisresponse ddonot change. The graph was: plotted using degrees Rankin, However, the X-axis was adjusted to display degrees Fahrenheit. As-%shown in Table 3).4 of NURI*JiG-i874, the higthest RT',~Ai.~6 value for the innsorged plants in the domestic Pressurized Wate* Reactor (PWR) fleet is 1873F at,32 EFPy:and 198.6F at 48 EFIPy. Therefore, it is unlikely that the RTMAX.Fo value for any'domestic PWR will ever exceed 240'V: (even above:60 EIPY) and the TWCF value fr tbrgings xwill rem 'in below that for. axial welds with equivalent reference temperatures. Therefore; tle Staff analysis on the effects of und*r-clad cracking has. no impact on the WCAP analyses when applicd to the domestie.PWR flee In the unlik.ly ev-nt that the RTMAX~ value foi a plantexcds2400 F.this anlybis aiid the 207year inspection interval would not be applicable without further evaluation. WCAP- 16168-NP-A June 2008 Revision 2

N-13 Since the "TIVC' correlations have been revised from those in NURIEG-1806 and a correlation has been dctenrined for torgings (even though it has no impact for domestic P\VRs)7 the WCAPNwill be revised to. retlect the changec to the correlations. The pilot plant TWCF vatues, which are presented in Appendices B, F, and J and used in Appendix A, will be revised using the updated correlations. Since all vessel forgiigs in domestic plants will not be atfected by under-clad cracking. there is no need to detem*inc whether the cladding was .fabricated in accordance with Regulatory Guide 1 43 as directed in NUREG-l1874.

3. \ppendix A-1 and 'ppcndix A-2 of WVs C -l.6168-NIP. Revision I identifies that tlhc'TWCF is a critical paramleter in deter'ininmng whethur the lidcnsce's reactor vessel is bounded by the analyses performed for the pilot-plants.

a) Licensees requesting to e-tend the instrvice inspection intLerva.ftom 10 years to 20. years inust prroyide the lollowine intfiomation ait the time'of their request: (1) determine their plant-specific TWCF using the latest methodology apl)pred by tdie staff for calculating the TWCF based on their plant-specific RTMAX and.NURILG- 1874; (2) deterainie the AlI.o values using the latest approved methodology documented in Regulatory Guide 1.99 or otherNRC-approved methodolog.v and (3) provide all material properties that were u.sed to dete*inme the plant speeufae 3A \X(c1 RI(.i.e. RR I R ,, I", W Ax-o.: RRTT',I,. w Al3 ~i.aale for limiting *naterials .in the beltline, maxisnin neuti on fluenee (cptFL) for limitiing materials in the beltlinef cold. leg temperatt"re under normalnbiperating donditions, neutron flux forlimiting. materials in thebeltine, and wt-i phosphorus, wt% manganesc, " '% nickel, :wt% copper for.limiting materia*ms in the beltlinc), Response: Appendix A of the W\CAP will be ses ised to. rquie thlat the plant speeific IAVC([I RTlMAX and i ml s-aloes be calculaled a4sstated above6 Appendix A willalso be revised to include all material properties required to determine the plant specific I WCX'I

                  'RT.,IAXv characicrizcs the reactor pressure vessel'sresistance to-fracture initiating from flaws found alone theexial weld fusion linscs and is evaluated forieaeh axiail weld fusion liiie.
                  'RT..a.Ai;. characierizes tlereactor pressure vesselr-sivsitancet6-fiiaetshe iniitiatifirfrom flaws found in plates that are not associated with welds, and is evaluated for each plate.

i'Ax~m~o characterizcs-the reactor pressure vesseSl resistance to fracturc initiatin" from flaws foind-in forgings that-are n6t associatcd with welds, and is evaluated0Tfr each forging.

                   'RTMAXCW characterizes tle reactor pressure vessel's resistance to fracture initiating from flaws found along the circumnlbrcntial weld fusion lines, and is evaluated for.eachcircumferential weld fusion line.

WCAP-16168-NP-A June 2008 Revision 2

N-14 (b) Licensees that have received approval 0ocxtend theinscrviec inspect ion interval from 10 years.to 20 years must provide within one year of complcting its net beltlinc inserN ice inspection, the analysis and data requested in Section (d) of thev oluntaru;PTS rule, 10 CFR 50.61 (a). Response; To addres tl. N1.A requirements fot rtporting and evahlution otinspecti on data,.the

                 .following rcquireinmenLs ,,vill bI. added to Appendix A. of.tei. WC!V :
                            "'Alldara on embeddedflmvis oj concern with a lhroucgh.irall avtenlt (TVE).greater than OJ inch shall be provided to ARC xivttln one year of completing the ne't vessel beltine insertice p&etou perAS , S&twt           XJ Appendui- 1,171, Siqyenent 4. [or potential yeS'" f!dihdtre due to PT7 , embeddedjflaws oftonaern re axially orientedpWanarJlaws in the ve~s'sl bcln within t~e inner 12 5% (TI) qf the wvssel wall rhickness:

Anuassesmentof the inseriCe inspeCoIon resulrt relative to the.flnt', distri*utions used in the pilot plant an'i ,es she all o tbxe provided. This atssavsnient shall be.pelbrmcdin aceoadodie,it/u the requirement. ofSection (d) i1 the final published version of the volunimarv P1'S rude, 10 CFR 30.61]a). The limitafion on the mirtniumi .\ E is taken from Seftion 2.10.22 on Probability of Detection and Figure 28 iii NURIU-01874. As noted, tflaws.%vith iasiaJIler TWF,were no(tielhided.in the vessel samples used for inspeetion quailicieation via the Pertbrmainee Deimonstrution Initiative. Note that Section (d) of the. voluntary VrS, ule.refers to Seetiqun (e)(2). which. provides the riquii-.*inins for moasurement and evaluation oI'surfacu breaking fIlaus, FIor potential vessel flailue due t .PTS. surface lreakmug flaixs oftconern are those Wxith fla w delpths all tlic way through the cladding'anrd into the base metal. The definition of "embedded flaws of concern" was.add6d to the WCAP insert to address the N-RC concern that critical flawconditions shiouldbe based on the flaw distribution, location and density that significantly contribute to the TWCF criteria for the pil6t-plants. This concern was stated in the purp0&eeof the technical basis docuciient for the embedded flaw limitations on densityand size for welds, plates, and forgings that were provided in Enclosure 1 of .SECY 104 6n June 25, 2007(.(ADAAMS: ML070570283). The teelmical basis document, was prov6ided by tIhe NRC.as.an Attachment to an April 200 7 NRC Memo, DevelopnI ofFlatt Size istribution Tablesfbr Draf*ProPosedTitle 10 ofthe Code ofFederalRegudations (10CFR) 50.61ta, AD..AMS: ML070950392. The definition is based upon those flaws which contribute most to TWCF as determined by the results of thelaitest trlS risk calculations that are sumnmarizcd in

                 .Section 3.3.1.3 of N-UR'.G-1874.

WCAP- 16168-NP-A June 2008 Revision 2

N-15

4. Section 5 ofWCAI- 16168-NP Revision I indicates ASMVE'ode. Section XI, Category B-J welds,

       'cure         Retaining Wetlds in Piping." at thereietor *essel noz7zes may be inspected at 20 year fequeney based on the analyses in tiie report. The Staff does not consider theanalysis perfbnned in accordance with this WCAP aipplicable for piping. The slail in a letter dated Denember 15" 1998, reviewed WCAP- 145.72, Revisionn1, "'Westinghouse Owners Group Application of Risk-lnformed Methods to Piping inservice Inispec~tion Topical Reportappoved Irlet hodolOgy for evaluating pipingý Ilhe stalf recomintamds tbat justification for increasing ihe inservicetinspeclion interval tiron 10 to 20 years for piping behjustidied in aeeordtrt with                           Revsion 1. Please revisk; WCX 16168-NP 5\.P1472,                                        .          accordingly.

Respnise: fhc PWR(XJ -udi ternoxe ( ateory B-Jxvelds hiorn thte applit-abilits of the 10 to 20 year intrval extension justiIied in WCAX - 16168-NP. Revision 1

5. WCAP- 161]68INP euvision I was written to justify increas ing lii inservice inspection intemal from 10 to 20 v*tars, for ASMI Code, Section Xl, Cateory, B-D welds, '"Full Pentlration Welded Nozzles in Vessils," Figures 3-1 and 3-2 indicate tlhat le beltline wvelds have the lovwest ratio of code allowable stress .ntensityvalucs ýI f ,pd, ligures do not include the full pt.entration nuzzle' to i) Thes.

w-sscl ' 1hds Ilie stafll -quests .that [lieNVOG prbovideuthe ratio of code allowable stress intensity value finr.full penetrationno;*.l .o-vcsselweld emtod.monstrate thait he bfltline welds are the linmiting: loeation*s Respon Th.e mar~nigin ratios o.stieSs intensity vsluies (Kid i.aN, sli) for the Category B1D. noLzle-to-vessel welds, are showfn in the.hable Weow. nett-i Noak i r~ Oriet aw Ainl j-170 Year 21.17 I MarnRai 1.20 30' T15i Cite __0I-.12 I Circumferential 30 SO [57 5-71

                                                      .x 140                                .5,71.

1)07

                                                                                    *67*                   :0.98 Circuiferential        110                                 8.62 20                            :8,62 30' 1                         8.62 The least limiting location iniie.reator vessel beltlinehas an N\SNE Code allowable stress
                  'intensit factor to applied stress intensit;' factormargin ratio of 0.504. Sinte this is less thanthe most limiting nozzle-to-vesseI weld location, these locations are not tile most muiti`gP:Of          region the reactor v:essel.

6. The Probabilisii Fracture Mechanics ýomputer To61 and Methodologyportion of Section 3.2 of WCAP-16168-NP. R-ision lindicates that the.fAiu're freq:uecy and distribution for all flaws i. the.

WCAP-16168-NP-A June 2008 Revision 2 Umfi:ien[ial

N-16 reactor vessels were calculated using the latest x,crsion (05 1))of the FAVOR code. This code has been signiifieintIy revised by Oak Ridge Nationil Laboratory. Providean analysis that demonstrates the impact of ttsing the lat6st Vetrsion of.FAVOR on the failure fi'quency ,nd distributions documented in WCAPI 16168-NP, Revision 1. Describe hotw tie results from the latt version of FAVOR code would impaet the conclusions in the X C(AP. Response: The hounding dideenees in through wall cracking frequency (TWVCF) and large early release Irequency (LERF) for different versions of the FAVOR ( ode are provided in the following table oti the three pilot plants. The bounding differences in TWCF and LERF were calculated in the tdslpmses to R XMs9 Part e and 12 Parts a and c; FAVOR Verions 02.4 and 03. 1 were used for Revision 0 of *VCAP- 16168-NP whilelFAVOR Version 05. lwas used for Reisioti 1 fbIl.bounding diflierenecs in TWCF mandl.EI.RF for FAVOR \Version 06.1. vhich was i used to .calculate the vatues ofT1VC I(or each plant in NUREG-I8(74.1 Recommennekd Scrceenring I rnimit~!r IPr&*uirize' Thermal Shoaek (P7.S) 2007, are pirovided1 im the response to RATM 8 .As can be seen in the table below., the estimated bounding ch ange inLLREF due to differentI ST Sintervals would itiil result in an insignificant chaing in risk(("<l0E-07Ayear) per the requirements of RegutatOrs Gmuide 1 174. "Iherefore, the risk infonned conclusions of! W XAP-16168-NP, Revision 1.remain valid for all versions of the FAVOR( Code that were tused in the risk svahuationsIThis conxclsion is also expeetcd to remain valid foi the next potential version of the. FAXVOR Code (possibly version .07. 1) that contains the modified embrittlement trend curves that. are proposed in the VoluntariyPTS'Rule (101C R5061a). Thisex ectation is-based upon a comparison of results in Tables 3.I and C(I in NUREG-1874 Thi. comparison-showed a maximum dififrenee in enibrittlement index (RNI ýx) of 5 1 and a maximinu diffierence in TWIrF of less than.20 percent at iisk analysis conditions (ais defined in Table C.1 of NURIG-1874) wsell beyond those shown in.the toll.owing table SElf.'Is oll FA\; OR V~rsion., m n WC F and LRF* HFfouiilins' D~iPffrences IRepresentative Plant Name Beaver Valley Palisades Oconee Unit 1 Unit. I FPTS Risk Analysis Cond ition 60 EPPY 60 EFPY Ext-A Bounding Difltrences.from FAVOR 3,44E-09 2.68E,08I N/A IVersion 03.1Rev.o of WCA), Bounding Diflfeences fromiFAVOR 2.49E-09 4.401-09 7.96E- 10 Version 05. 1 (Rev. I of WCAP) I Bounding Differences from FAVOR 9,37E-10 I.8E-0 1.26F-08 mion 06:1 (response to RAI 8) . ......

7. On page 3-12, the.Topical.states.that "The:following effect%also need to b c:onsideredalong with the change in IS interval: Extent of inspection (percent coverage), Probability of detection (POD) with flaw size, Repair criterionhfor removing flaws fi6m service." Also on page, 3-12 it states that "For the pilot plant evaluations, examinations were assumed to be conducted in accordance with Section XI Appendix VIIl, so that Figure 4 could be used." Figure 4 is a graph of POD vs. flaw size. But on page3-1 Tit states "For examiple ifithe probability of detection fir the first inspection was 90 pertent, then the flaw density was effectively multiplied b'10 percent fbr:input to the next iteration." These calculations determine how many flaws and the sizes ofthose flaxvsltlit will be include1din the %s.dat".ilefor surface-breaking flaws WCAP-16168-NP-A June 2008 Revision 2

N-17 in the FAVOR code calculations. Because the Appendix VTII inspections are required for only welds and a small portion ofadpjacnt plate material, these inspections typically cover only a few percent of thc vessel surface in the bt-line region. lowever FAAOR typically models surface-breaking flaws as being randomly distirbuted across tle entire inner suilace of thie vessel in the belt-line region. It is not clear from the topical how tOle effecSL on the density of surltce-breakino flaws were moditied to reflect I fieIrction of tihe stufa areacov ered by the inspections please provide thoe peent-eoverage fr each of the pilotplants. Please provide a.elarifictition of the FAVOR calcula tions that explains how thle percent oer'agewas maS n)rporited, hi particular, please be dlea regarding assumptions About the presence ýand atlhcts of inspections on surthce-breaking flaws in the areas not subject to Appendix VIII inspectionls. Rbsponse. As disecussed in Section 2 10 1 of NUREG- 1874 the flaw models noiý unsd in version 06.1 of FAVOR. do. not directly eonsider the eflbcts of in-seriec inspection. roevxahate the eflects of fatigue crack growth and in-servicc inSptetion on any surface breaking flaws, thle flaw input file to FAVOR must be modifiud to include.these cffects. Howew'eiSection 4,4 of the Theory and Impeanentation Manl lfor vkrsion.o06.1 of the'FAIVOR Code staItes that the.flaw inlorlmliaoi in t*lheone input file (S.dait) ti the 1000 surthce breaking flaw distributions is applied in the same imnner to cladding over welds mad cladding,over basee metal (plates and forgings) Therefor.e fltheeffect or the-cry sniall inspection exveragein tile base metal xsas NOIcornsidered in ile PTS risk analyses discussed inRevisionz 1 of WCAP716168-NP If it had been eonsidered. then there would bcdabsolutelyzero difference in the iVCt due to inspection interval for the. stu-tacc breaking flaws in the base metal that are nscir.inspected. Aithough this effctc is small neglecting it.is nonethe-less conscerative, beciuse the,actu al dilkines.... in

                                                                                                      .r. C and
                                                                                                             *u             i
                  'to the l iehne, ni in.spection. inter,,al, .would* be. low em,than
                                                                                   . those
                                                                                        .. estinited.
                                                                                                .      . the "XVC*AP
                                                                                                                .n Rep.lort..

Thati'sý the actual djfferences in. TWCF and LERFRwould be even less statistically significant relativceto zero, as, discussed in the response to.Part d) of RAI 12 because the cffects, of tei uninspeeted base-metal flaws Were included. Risk

8. During.an October 11, 2005; public mecting.with tli.*uclear Regulaturv Commission (NRC)j (summarized in ML052910i48) the NRC staff and W\stinghouse discussed the rclationship betweenlthe:

proposed WCAP and.the PTS rulemaking work. The NRC staff noted that Nuclear Reactor Regulation's-(NRRWs comments regarding tiqqiressurized thermal shlickl(PTS)techlnical basis may, affect the.results of thecalculations in the WCAP- 16168. The NRC staff also noted :tat if tlie Westinghoiise Owners Group (WOG)'subiiitted \VCAP-1 16168 tpior to the resolutIion of NRRI s comments by RES the WOG would be expected to address N-1RA's cormnents as they affect the WCAP-16168 cafelulations. A critical component of the iistification of the requested inspectiofi intervl ecxtcnsion is.a fracture mchanics;evaluatibn of the rietor vessel. The PTS technical basis ,and.the Topicaltuse the FAVOR code to estimate the condilional probability of react6r essel faihlre. tlie resolutio ofi fNRS continuing review of the PIS rulemiaking technical basis has.caused theTFAVOR code to beinodified to correct deficiencies in the code. According to Referenee 26 in WCNAP-] 6168, FAV.OR code Version 05.1 wis used in the analsisiused in the Topicalr.the current version of the code used in the.PTS technlkat basis is FAVOR,6. t The changes made to verion 05.1 resulted in substantially incevased values of through wall cracking fre*quencyý (TWCF) for the pilot plants and significantly diftfeient eonrelations of TW.,CF to material reference. temperatures. Both of these factors are important when licensees relate the Topical analses to their plants. Please update the FAVOR computer code anal)ses .u.sing'ihe lates ersion of the FAVORl, code and make any corresponding changes to the anabyses presentedl'in'WCA1P-16168 (The, NRCQ does not WCAP- 16168-NP-A June 2008 1 Revision 2

N-18 expect the FAVOR code to undergo additional changes b fore the technical basis for the PTS nilemaking is.completedU but concludes that the version of the FAVOR code ultimately found acceptablein the.PyTS technical bask will hbethe*er'sion that%,ill be acceptable for reference by the W\OG in this Topical). Response: 1he pilot plaint ianalyss and change in risk calculations have been updated using version 06.I of the. AX OR (ode to he consistent with NIUREG 1874. The .esul.t of the analyses and change in risk calculations for the fltrce pilot plants are presented in the table below. CBeaer _______ TWCVFad 1.1'IF Re- iItSALV-n-ISie fftir) Valae Fnlt 1 Pawl~Jde% Ocorne Unit 1 S0yeariS ont¶ (Met' .5.0o4 5 -0'(9"2'541 -i j 7.62C-08/4A)gf.O 09 3,1.1E-08!2.551-09 l*.} .u!................ T

• 0 .. -

                                                                                                                                ......  ..1
                    \ alue;("Sandard Erio0       i                                   .                           I_

Lpov"r Bound 4a1.t9 1 63! 08 2.36E-08 The WCAP will be -revisedtl inehude the revised results. Appendiees V.E I and M will also be revised .to include the-F.ALVIVO)ST output from version 06. L

9. Paige 4-6 provides a deksciiption defincdias a consc vadi, c/bounding acceptance criteria, relating, Change in(Dlc C(hange inEl-.ERIncrease in firequLcy of through wasll crack growtlh<.E-.7*yr.

tPage 44'stdates that, "[It, "this evaduation, the CD ...and . RI wserc-ealeulated by cDF-LERF-iE*CPF where

         .CDF= Core LERI-          damage Large            frequency from a failure
                             ;early: release,"freqiuencyv              (events per year) e el frm aTfiue(vl LER                                                  dilute (events per year)

IE =.Initiating event frequeney. (events per year) CPF =Conditional probability of reactor vessel~failurc. a) Please precisely dcfinc CPF and fully describe theprocosscs used to calculate the values. Is this the conditional probability of failure given a PTS event on the last day of the last opierating year? Is this the average conditional probability offailure given a PTS event randomly occurring during the operating life of the plant? Or is thiis some Other' paramelci'7 Response: CPF is the distribution of theconditional Plrbbabilitics'of failure given thit:all postulated 1FeS events occur on the first day, of full power operation.following theircfuiling

                   *outage after the last operating year for th6 extcndcd licensc of the plant "lhis is a conscrvative
                   ,approach as discussed in the response to R A 9. The calculation.of CPF and TWVCF for dite two inspection intervals is sumnnat i7zd in the subsection entitletd "Pobabilistic Fractuir *c46clhanics Computer Tool and Methodology" (pages"3-14 to3-17) ititSecti.on 32 of the WCAP Report.. Tie calculation of CPF by the.FAVOR computer code is described in the following NRCAepoits: 1)

Sections 7.1 to 7. 10 of NUTREGýi1806, Technical BasisforRevision ofthe Pressurized'ThegrmaI Shock (PIS5i Scre¢nng Limnit~min PTS Rule (lXCFRXR0.61): Summary Re1port. 2006 2) Section 4 on Crack Initiation and. Seioin 5 on "firough-Wall Crackng in NUREG-1807, Prbbalbistic, F'ractu14r&Methnie.t -Models. .Pwecalinrisand Un .srtaihh'TreatnuintUZst*d in P:4 1VO/R Version WCAP- 16168-NP-A June 2008 Revision 2

N-19

04. , 2007 3). Setions3 and 4(Equations I to 132) of.NURICR-68S4 Jractwe 4lyis q/

Vesze/s -ak Ridge.F'l kOI v04. 1. Compnier Code: Theory andhnplenentaton of , /gorithms Metht's andtorrckazwn.*, 2006.and 4) Section 2 and Appendix A of NUREG 1874,

               *Recoa~mmnded Sc.ec'.ning I unrti /it      ersuri,:edTlgrmf Shock.(PS),2007. Appcndix A of

NURI (, 1874 deseribes the requested changes in going from FAVOR versi on 05 1, which was used in Re isuon I of W(CAP-16168-NP, to FAVORk vrsion 06. i. which was used to calculate the CPlF and TWCl results in NU1'IG-1874.

DI is the disirilution of frequencies for each postulated PfS transient (initiating event) thatis combined with the CPF distribution to obtain the distribution of through-wall cracking fiequency JlTWC1') for that P1I"S transient. This combination of the CPF and IE distributions and summation. of Ih\VCF diibulions for altlcontibuting .PTS transients is peffirmed in the 1FAN;POST Module ol lAV OR as dcihcnbed in Sectiom 2,7 of NUjRJ.G/CR 685b, tractl-e bAnalyss qf 1'sqi rL...rAi*irdg1. 17 MR., V04., (.nimpder (.ode: Uisers (jwide, 2006. Section 2.7 also provides the methodology' and equations lor calculating thetstaLtitcil paramuters for the total 1ACl distributions that ame providcd in the AVIPOST output.(Appendiccs E. land V folithe three pilot plants in the WxVC'AXIreport respectr by),. The total I.TW( distribution beconies the ILERF distribution bccaus.li.e conditional prohability of lar"ge early release given vessel failure is tken as 1 0 as dc* enhed in Section 10.5 of NUREG- 1806; Thie W l?iC be revised to show thatý wsl Where: CDF= Core damnagi trequcny trom vessel failures due to all iI'STc-nct -nts per year) SLERF '~Large eadrl.**yrelease ~frequency fromvessel failures due to all PF'S events (events per year) IF _ýInitiating &eentfrequency (events-per year) for a given PTS transient i. CPFI-Conditional pfobability of reactor vessel failure for a given PTS tranisient i, and N "Ihe total number of posiulatied PIS transienis for a given ~plant. b) Please describe and justiJ th) opera-ting life selected fora) above Response: Because vcssel failure during a postulated PTS 0eventis morelikeIy to occur wvith a higher deg-ee of einbsittlesnet, which increases with operating timrie duelto.the aecuimulaieed neutron fluence,.an operating time that is. realisije but not 6oerly conservative wNas.desired. Another Consideration was trying to.bound most of the plants 6f eacli nuclear steamr supply system (NSSS)vendor's design using one of the.operating conditions in the P1'S Risk.Studv performed by the'NRC:(i.e. fromn Table 8. 1"5,of NUREG-1806) ihroughithe end of the first license renewal period (60 years), as requested byrthe PWR Owners Group. Usingthe:information in Tabie 9g5, Plant List for. Generailization Sfuiayj in NUREG-V806. Beaver Valley I at 60 EFPY was judged to beboixnding for enibiittlementitatall thie plants with a wsistinghiusc :NSSS desin, inl;iuding more enrbrittled Salan i. at 60 cal*ndar %rears. Likewise. Palisades at 60 EFFY was judged4t be bounding for-ernbrittlemfient at"All:the plnts widthe'a Coimh§onEiigneerin NSSS design, including more embrittled Fort Calhoun, at 60 calendar years. However, Oconee atr60 EPPY would not be bounding for emb-fittlemcnit at all the plants with a Babcoclk-& Wilcox NSSS design, specifitclly TMNI-L at 60 calendarIyears so thIe next highser extended condition A was used WCAP-16168-NP-A June 2008 Revision 2

N-20 for this pilot plant. 'llh.se FTPY.conditi6ns set the vessel accumulatcd fluencc and material . emrbittlenment .!eels. A maxinium operatingtime of 80 calenddar Years was used intstcad of 60

                  *cars. This longer operatilig time is not only onsidered to be,bounding but is also conservative tor two reassons. Firist. the transients in the plant design duty cycle that could produce fatigue crack afrowth are speclified using a gb,,en 'ale per.calendar year' as described in the response to Polt a) ofRAI L Therefore the fialigme cr ck growth would hi.about 33%          /higher due to the larger total numbeir of tdtiguetransients. Second, the effect-s of in-service inspections at60, 70-and 80 ai.endltri yeai are added in thi cases f6r ISI every 1.0 yerns (a 60% increase relativet0o the last inspection after 50 years), Both otitliese conscrv'atisms would tend to maximize the differences in 1\VC(F1 for tSl eiery 10 yr carsr*iativ to fixe ca.sus 101   l-year ISI only.

e) As: indicatied,in the use of a "conditiowil probiibility of reactorvessel Thilure," reactor vessel fiilure only occursr whien a demland (the PMS. event) is placed on a vessel that has become susceptible to faihure throuixh the growth oferncks. Cracks grow over time and may bccome largeenough to fail .given a FIB eavent but remain hidden until revealed iurough a reactor vessel weld inspection oi through, APT'S evoent and subs.-equent failure, Withot in event or an inspection, iheCPF incicases over time as the racks grow throughout the interval \ormally, the risk from unrevealed faults during aninspedtion interval is estimated bIasud on lhI randoni oc .unence of the upset event during the interval combined with the likelihood.of tihe unrt'vcaled fault as it inci'eascs throughout the inspection intervaL- Therisk .rsociatod with th . exut*ded interval i.i similariv estitiated. Tihe risk increase is the differene between these two riskcstimatcs, Please provide this estimate ntlite change' in isk associated. with extending the inspection interual fhom 10 to 20 cars. or jtistil' thesatthetii*ttc s iii the topical yields abounding estimate ofthis value. Rsponse: 1hi. e~stimiated .ant. in large early relerse, eqflUencyv assoiated with extendingý the inspection interval from 1t)i0o 20 years is.provided tfr the three pilot!plants for each of the NSSS vendor designs in the last rowof Table 4-1.on page4-8 in Revision I ofWCAP-16168-NP. All

                   'ofthesu*alucs aarc considcred to-be bounding estinates for the following reasons:

1)The values were calculated using thie samemethod6logy that was used in the PTS Risk Study pefformed by the N-RC,, Which has I! known *onservatisms: per. items (a) thlroughi (k0) on pages 12-11 to 12-12.in Section 12.4 of NUREG-1806, as compared to only 3 potential non-conservatisms per items (a) through (c) on page 12-12.

2) For most oftheplants, the tmbrittlement at 60 EFPY is used.to bound plants ai.the end of their first license extension (60 yeais). For. the remaining plants. the cnibrittlement at the EFPY at extended condition A is used to bound plants at t"he end0of their first license extension.

The l3 number ofdesim dultycycle transients'that could produce faiigue crack growth is about

                   ý33%higAier than the value for 60 years of operation.

4) Most plants are projected to not reach their design basis (40-year) number of transients after 60 year's of operation (first license extension).

5) The cffects of ISI are assumed to he cumulative, which is conservative because the reater-the effectivenessof the IST, theigreater the difference in TWCF due to the change: in inspection interval.

                   "6) The number of in-service inspectioni:that is cr6dited in the cases for ISI every 10 years is 8 (6,0% higher) rather than the expected value of 5.since no credit is usually taken for any inspectikns after the ee .nded oper,'aing license has expired (afner 60 years of operationi).

WCAP- 16168-NP-A June 2008 Revision 2

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7) 1The cases for only one inspection after 10 years of operation are conservatively used to under-estiinawt the ,ffeets of in-service inspection every,20 years (Sec Fignres 3=6 to 3-11 in the WCAP report).

S).( redit for the ivduction in flaw :density due to in-service inspections is conservatively applied to portions of the plates and forgyings tfit are not even inspected Wxhich would over-estimate the diffmencns due to chanugv in the inspection interv al per.the ressponse to RALI 18.

9) Upper 2 hound values (-97.5%) on the mean TWCI are conservatively used to estimate bsigma the bounding differences instead of the FAV'OR calculated mean dvalues tbr the.eases with ISI only after 10 years of operation (see response to RId 12 part e).

10) 1oivcr 2-sigm bound xV~aes (-2.5% ) on the meain.TWCE are consersatively used to estimate the boundina difercrirecs instead of tie F'AVOR calculaitd mean values fbr the cases with .I.S every'10 years (sw rcspoplsn, to RAI 12 part 6).

i1) Using sepurste ipper and lower 2-signia hounds in items 9) and 10) is about 40% conserviative relatit e to using an upper 2-sigma bound on tho combined uncertainties in the difference in mcan values of TWCF. It0. Paae 4-8 states'thalt 'ýthetrinsient inuitating frequency distributions Weie identified in theNRC PTS Risk Study [71 and afee included in Apendices D. KLand I for the pilot plants. C1ie Appendices include, the Qkroupcd) sequen**s but do no0t iLnICude the transient lrcquency distribution. Please provide the mean. values or the transient freqiucney in the Appendices to provide ithe link to the PTS technical basis results

       - that~are used in the Topien.

Response: 'rhe P*!S:transicnts are the same as those'in Appendix i..of NIJG-1806. T7he

                   'qH1-f   in Appendices R H. and- e1,corre nds to that in Appendix A.-of NURFoG- 1806. A column will be added to the tables in Appendices D 11, and L in tle WC:Al repori to include the.

mean initiating event frequency firom NUREG-1806 for eaeh of the transients.

11. Extending the interval for ixint pection of readoit vessel welds wsill to isXoetent, incrcasc lhu likelihood that'a PTS event will cause a reactor vessel failure.. A reactor vessel failure will tail the'reactor coolant fission product b6undarr. and may dirctlv faiiI thereactor hde fission pioduet boundarvy The diseussion on page 4-9 and 4-10-about maiintaining defense-in depth emphasizes 1) the low likelihood of a )Yit induced rupture, 2)"that a:%nmplingof plants" inevitabi undergo cmaminaiion's in:a given year so that unaknown degradatiorinmchlanisms will not be ignored for 20 veat's,, and ) tihat all reactor coolant pressure.boundaty failures oceun-ing to date have been identified though leakage; a) "lledefense-in-depth evaluation is performed in parallel with the risk.Ievaluation inthe integrated decision making proeess_ Please assess the propo'sed increase in inspection interval against each of the defense-in-depth elements listed on page 4-4 of WCAP-l6168..

Response: Page4-4 of WCAP-16168-NP.Re n'i.a states fion- Regulatory Guide 1.174 that "Defliscnin-depth philosophy is not expected to change untess:.

                      . A significantifiercasu in dhe existing ehalln-cs tb'fli iritigriiv. fi barriers occurs.
                     . The probability of failure of each barrierchaniges.significantly..

SNew. or additional failure dcpendencies arc introduced that inircase the likelihood of fiflure WCAP- 16168-NP-A June 2008 Revision 2

N-22 compared to the existing conditions.

             *,      lie overadl rdundancy and diversity in the barriers changms."

Thle extCnsi~n its inlspeeiotU initerwi! will not~e slcint any of the chainges ide~tifiied ab)ove. For this reason the detru.se in depth clemehnLs listed on page 4-4 will not be impacted. Additional afssessmentof the impact on each of the dekinse-in-depth elements from page 4-4 is provided in Areasonable balance among prevention of core danmge, prevention of containenit failure, and Colsequenicem figtation is preserve.d: The proposed increase in inspection would not cause. an increased reliance on any of the identified elements. Therelbre, the interval increase would not change the existing balance aniong pre-vention of core damage, prexention of containment failure, and cons'eqtjene"nmitigation. Over-relian"c onprogianunatic actiVities to Compensate for weaknesses in plant design is avoided-The change in inspection interval does not chlinge the robustness of the vessel design in y, It isbcecase 0tfthis robsitness tlhat the inspection. inteiral can be doubled With. no significant change in lililurc fireqtuency. Systern redundaiicy inidependence, amd diversity are preserved comrnmneisuratc with the expected fhequenc, and consc quenecs to the system (C.g.,.nomiisk-outliers): Theproposed inrease in.inspcetion iuiterval does not impact system r*idundancy, independence, or diversity in any way, since it is not clianging thc plant design orhow it is operated. Defenses agaihstpotential oiommon eausesfailures are preserVed and the potential for introiduction of new cormmon cause failure mechanisms is~assessed: The proposed incereasde in inspection inteival does not impact any: defenses iigainst any common cause-failures and there is no reason to txpect lhe introduction of any, nw comnmon cause failure mechanisms. This requirement applies toi.multiple active components, There is. only :one reactorxvessel per plant and it is a passive comlponent.

               . Independence of barriers is not degraded (the barriers are identified as the fuel cladding, reactor Conlant pressure boundary. and containmentstruct ure);

The increase in inspection .inteival does changelic relationship between the barriers in anyway and thereforedoes: not degrade the independence of the barriers. T'e changein inspection inteival does not change the robustness-of'the vesscl design in any way. It is because of this robustness thatthe inspection intervalcanlbe doubled with no significant change in fiahmeIrcquency. WCAP- 16168-NP-A June 2008 Revision 2

N-23

                     *Defenises again*st: human errors arc pre.ser'%ed:
                           'lThe increase in the RV inspection interval does not impact any defenses against human eirs in *muywa. Thc increase in the inspection interval reducesthe frequency for which the lower internals nieed to be removed. Reducing .thisfrequency reducet, the possibility. for human error and potentially daimaging the core.

b) It is likely that all plants ,will resquest to Uxtend the inspeetion interval from 1.0 to.20 years. Universal, ornear univer',A.a!doption of this option would, unls, aherwise.ananged, lead to a 10 year periodnwere no reactor vessel weld inspeetions vxiiuld be required, Please priosdc additional discussion. speui~l '1ug how a .'ampling of plants" perfinning reactor vissel welds inspection over the next. 10 yeais can be achieved. Response: On October34, 2006, the PWROG sulmitied to the NRC Letter (O-06-356, "Ilan for Plant Specifie Implementation o0.Extended Inservice Inspection Interval per WCAP16168-NP .Revision iOn of the Reactor Vessel In-S er ic Inspection Intervai',; This letter provides i.plan of when inspections will be performed provided that Revision 1.of WVCAPo16168-NP:is approved and that each plant'makes.a.plant.specific request to the Staffio Sexteid their intervil. As discussed in flte ettir and as previously'agreed upon by the.Staff the. the -WNIOG Staff will r17CViexplant speefic:requests to implcmentt the .20 year interval-against. Plan. Appio-al of tie plant. specific request is expeeded if the date requestedi'fthe plant siceific request is within one-refueling xycle of that ideitilied in the PWROG.Plan.

12. 'thbe foltoowingcrepeaLs part of Table 4-1 in tie'Topical.

Table 4-1 (mean vahies) LargeEarly Relcase Frequtiencies BVI Palisades% OC1 10-YXear1SI On3l, 5.04F-09 1.5,4E-08 2.06E-09*

          .ISiEye'     10 Years          4.)IOE-09                     1.67E-08't                 2.18E-09' For Beaver Valley Unit 1(BV1), the ISI Every 10 years (4. 10E-9) is less than the .10-year inservuce inspection (1S1) Only (5 04&-9). This seems reasonable becautse the repetitive ISTprovides 'oppoitunities, to find andremove growing crack6before they can lead toYssel.failure given a PTS event, Hoxever, for Palisades andOconee Unit 1(OCI) (and in , number ofindividual bin frequencies in the Appendieý) the situation is reversied.For ex*imple, for PIalisades above, thel.S1 every 10 years (i.e., 1:67E-08) is greater.

than the Il0-ycar I S Only (.1.54E-O8)J This appears to indicate that it is riskier to itnspect than tonot inspect. but it may demonstrate that the Mnte Carlo calcuilations by the FAVOR codc were'riot' converged sufficiently to reduce the uncertainty Ofthe meanwvalues of the total TIWCF to less~than the effect of detecting andi.retnovix esuface-tbreakinig:flawvs tfound byinspections at*,tn.year intervals. ai) Pleadexplain wh3,th~e mean failureestirriales are sometimires opposite of what!is expectetd. The explanaition should include a justification that this anal..ysis is preiise:enough to supportthbe charige in risk estintsifiitead of further investigatinIg the aplaient disrepancy and developing results that no longer WCAP- 16168-NP-A June 2008 Revision 2

N-24 appear contradictory. if the answer to b) below remove- this apparent discrepancy lhe answer to this question. can be retenred to b). Respouvw: The tc~hnical basis for the PWR Owners (iroup pr(lject to e"tend the vessel ISI inte-,val was always based on the premise that surface brakijng [lws would never be a significant contributoy to ,essvlfailurc :tbrough-wall crz cking) frecquency due to P1TS trinsients. This was based upon the fact that the frequency of sudace breaking flaws wouldlhavc to be very small, since none had ever been discovered during either pre-sernicc or in-service examinations and even if they did c\ist, their circumferential orientation due to thecladding welding process (see wSiction ýk61 of RNUREGiCR-6817) would lead to an-estbdtre through-wall fraceture (see Figure 9.7 of NURE --)806) Therresult pulorted in Revision 1 of WC'AP- 16168 NP just confirm this jiionise sen~ Wihetn potential.tiatigue iak growth is e'plicitiy considered Beeause oftho. uncertainty in how accurately an.insignifictnt (null) cillect can be ealculaied by FAVOR using standard.Monte-Carlo simul tiion methods a uen.servaaivet method of comparing upper and lower 2-sign a bounds was used as described in the respouse to Pant Q6f thislRAI. b) The T\Vi.C.F estinates are dominated by the more numerous embedded axial flaws, with little contribution fhorn the surfae"4-reaking circuinferetiail flaws that are varied by the WCMP FAVOR analysis. [he FA'VOR code treats eachflaw independently of0 vcr, othbei flai-,and it is possible to Calculate tile efleet of theTXTWCT contribution fbr. only suiface breaking flaws, without including any of tlleernbedded.flawsin the calkulation. Such in evalualion iouldlisotai.tle parameter of interest (the TWC1 caused by. sur'face flaws) and'the*reby eliminate the possibly dominant affect of the :uncertainty on the quantitafive r;sulLs associated 4iith the 1I',CI from tembedded flaws. In order to appropriately evaluate tlhe: unertainty of the T)VCF c~ontribution created by surface-breaking flaws as opposed to embedded flaw-S please ealuate these flas separately, so tht probability distfibutionis tor tie iTWCF contribution of surface breaking ftaws can be obtained and compared for the two..inspection cases. Response. [atthe reasons stated in the rePonSe toPra of thisR Altereshould be almost no contribution fronm surface-breaking circumferential flaws. However, Witli such few failuirs

resulting firomsurface breaking.flaws. there may be no way toobtain a converged solution using Mointe-Cairlo simulation because the.accuracy is,.based upon thienumber *if failures in the total number of 'essel simulations. That is,to obtain convergence and acceptable accuracy, a

                   ,significant number of fiures are required for the specified number of simulations. Note that 70,000.v(essel simulations were required for tie results provided in the response to Part c of this RAI. Evcn for 500.000 simulations w ithoui any embedded flaws, the value of throuh-wa ll cracking freqi*eney (T:WCF) c*alulated by FAVOR ,was zero (no failures) for both ISf cases. The change in TWCF.and the change in large early release frequency (LERF).would also be edssentially*zero,:whliic:woutld cetlainly be Considered,insignificant per the requirementls of Regulatory Guide 1.174.

c) Please describe~in detail how the.mean upper bound and mean lower boundparimeters (included in other ctitries in Table:4-4) are developed.

Respofise:.The inoitrmation in Tabld '4-1 is a summary of fie vessel failuirefrequency results for each of the three pilot plants thatare given in Tables 3-2, 3-3 and 3-4. rcspectively, in the WCAP. ror the first plant (Beaver Valley unit'l), the meati.vilue andstandard.eiTor for 10-YearJSI O(nly of5.04E-09 and.4.83El0. respectivelv, in Table 3-2 on page 3-18 were taken from the FAXPoST Output ialues of 5.0405E-09 and 4.827.2E.4.0, rvspectiely, on page E-5 in Section E-WCAP- 16168-NP-A June 2008 Revision 2

N-25 I of Appendix F. As described in the first paragrapli ofSection 3.3 (page 3-18), I'lhe Upper Bound Value was determincd by adding 2 timens the standard eror as reported by FAVT OST to the mean value of the 10-Ycar ISI Oniv ease." in Tible. -2 the Upper Bound Value of 6.0 17-09 camefiin e 504054)9 -0 2.' 4 827215-10 6.00594)209, The mean value and standard eror tor ISI Every 10 Y l.r of 4.101-09 and 2.891- I0) respectively, in Table 3-2 were talkem from thi FAVOO.ST Output sahi.* vf 4.09y9SE-(09and 28934E- 10,respLtneltvly on page E-1 in Sei~tion E-2 ol' AppendixE. Is described in the first paragraph of Seetion 3,3,. l-Alh*1 I.er Bound Value was determined/by Stubtiatiling,2 times th[idstimdard error as reporl&/dby FTAVpOST from the mean value of the ISI Ev-rv 10 Years cisc:' In TI.able 3-1 the Lower Bound Value of 3.52E-09 came from 4.09951 2X 28934E-10 3'52082E-09, As described in the first paragraph of Section 3 3ý"a chanmein failure fi-qucicy was eonservativcly calculated based on the diflkcrince between an Upper Bound ind a L,6wncr Bound InITable 3-2 the Boundidg, Differtlen of 2.49E-09 came from 6.005941 3520821-09= 2.48512E109 These same calculations were also perlortned for Pnisadtes in Table 3-3 in Section 'r4.usnig thie FTAV*kJST Output on pages 1-6 and 1-12 in1Sections.I-1 and I-2 of-Appendix I and for Oconec Unit 1 in TibleO-4 in Seclion -5 usine t*ie FAVIN)ST Output on pages.M-7 and M-14 in Sections.M-1 and NMo'of Appendix N. The lmotllowisg equatlions fr tile standard deviation andstiandard error are provided in Section 2,7 on FAVPOST Output. ii NN'REG/CR-6855.; FractureAnah.sis ofl*essels --Oak Ridge. FA4 "oR; v04.1, "2arnpiite'C(ole:

                               .'Grdd U5'bm¢e             2006:.3 Stndidi'd Devsrtiin4 .s=

Standard Error*

• n' I)

In these equationis, xj is the value of TXTCF for all PTS.transients; wlieidhis cailculated for eali vessel simulation i, and tle A swith the bar over it is the meanvalue of TWCF for all n vessel simUiation~s. Whiili. is typically gireatei thani 60,000."Note3 thai thet standii~d error, Which is a? measure on the umcertainty on tlhe mean value is equal 'to the standaid da'siation, which is a measure of.theuneertainty iin ll ,the sniulAted Nalues of TWCT divided by the squaire root of the. UImiber of simulations,. The uni~ertaintV on the mean ',,alueof TWCF is.used per the guidanee i Seetioh 2.2.5.5,Cot n'iparis-46iis with AcceptanceGuidelines, in Revision 1 of Regulatory. Guide 1i174. "lb1isseio.n states: ".Because of thie wivthe acceptance guidelines were developed, the appropriate numeri1rcal masues to use in the iniial comparison of the PRA results to the acceptance guidelints are mean values. The mean values referred to are the means of the probability distributions that.result fironi the propagation oiithc uncertitinties on the input-

              ,parameicrs and.tho.se modei unceirtaintiies.'cpiicitly rieprpsented in the model.'

Thiis same approach 'was used to calculate'thle Upper hIounid. 6oer bound, and change in failure frequency for the FAVORversion 06.1 results presented in the response to RAI 8. WCAP- 16168-NP-A June 2008 Revision 2

N-26 d) Page3-18 states that; "[s]tatistically, thedil lrenceIbctween the mean failure frequencies for.the "ISI Every 10 Years- casc and the "lO-ycariSl Only" case is insignificant." Please describe the statistical techniques used to develop thiststatement, c,g., was hypothesis testing about the iwo means perlormed? How doe* this observation support the use of the Topicil methodology in demonstrating that the increase in tile inspection interval from 10 to 20 years satisfies die risk-infitried guidelines in RTG L.174. Response: The null hypothesis is that tlhe risk difference for the two ISI cases.s zero ]for the reasons stated in the responses to previous parts of R*l 1- For the differcnce in mean values to be statistically significant at the 99 percent confidence level,ithe1-statistic would requireits

                   .value to be equal to oi greater than 2.35 times the .sample Standard deviation. For the detailed 13-1 example in part e above, the sample standard deviations would be tihe square root of the sum pl'the squares oa the standard einors' li the tis o ISI cases, (4:8272' + 2.8934`)f x ti.E-1-=
                   '.6279E'1.0. This. value is 2.35 times the"Y)'o eonhdenee bound of 1.3226E-09. Ile actual difference in mean values is .50405E09 - 4.0995E-09= 0:94101 -09, .which is therefore not statistieally sigiuficatit relative to zero at the _99'%, conflidence level, Eeen if the results-were recvrscd and the.difference was   a 0.9410E-09, it would still not be statistically significant relative to zero.al. the"99%. co.
                                           . i denee leve.

Section 4 in Rcision . of\VCAP~t16168-NP ineluding . themethodology to ea eulite thie ch'ange in risk in Table 4-1 as described in the response to Part e of this RAI, clearly demonsti'ates that

                   .the incrcase in tie inspection interval from 10 to 20 years satisfies the risk -informed guidelimes in Regulatory Guide 1.1.74.
          .13, The Tables in Appendix A. appear to.illustrait dite inform.ition that the WOG proposes Nvill be contained in:individual licensee irlief requests. \Ve note th~at the "Plant Speciei B ,sis' prop~osed by WOG in the Tables in Appendix A refers to the "Pt]'S CGeneralization Stu*d!ys.i document that was not submitted by thle WVOG as part of die Topical and is not being revicwed by the stafffor usein relif rcquest to extend the:inspection interval ofwreactor vessel welds. We also note that there is a plant specific.

qluantitatve' estimate'oftlthie "hrouah Wall'Cracking Fequecy" in the taw o examples that implies a plant: specific calculation. In particular. please explain the value of 2 15E- 12 events/iear provided for ile W61f Creek~example plant in TalIe 1 of Appendix A-1. Also, please exphain ti eiviue af4.67F9 dveitsN uar provided for the pilot plant in the same table, Which does not Seem to match other pilot plant information elset~hcre in the Tolpical. Plase describe tie analysis that the WOG proposes that icensess -will need to perform to support a plant speeific relief requesL and relate these analvses to themethodologyand results in tile Topical for which ilite WOG is requesting approval. Response: The n PTS CGeneralization Study" (ADM)AMS Accession number: NM,042880482) is Re!reiiee 25 aoft CAP- 16168-NP, Reisi on 1. This study "Was performed as part of tde NRC PTS Risk Re-mealuation as described in NUREG1806: Thepuipose and conclusions: offlte Generalization Study arestated on pages 3-6 and 3-7.of the WNXCAP, respectively. The purpose is consistent with t!at stated in the first paragraph in Section:9.3 of NLREG I806: "Our aim was to identify whether the design and operational features thatitae theIkey contributors to PTS risk (see Section 8.6) vary significantly enough in the larger populhtion ofPWRs to. tquestion the generality of our results." "h~over0all eonelsion ii consistlen waith that stited intlie last paragraph in Section 9.3.3 of NLREG 1806: "These combined observations support.the overall conclusion that the TWCF esiiniates prodticed far the*eltailed analy.sis plants are sufficient to characterize (or boUnd) the TWCF estimates for thiefivE gencraliza Lion plants and. thus. by interj,'nce, PWRs in genei-al." The Generalization Study was reviews'ed by the Staff 's part of tie PTS Risk.Re-WCAP- 16168-NP-A June 2008 Revision 2

N-27 cvalua tion, and it is.thc basis tbr thei flet-wide apjdiability of the proposed PTS rule in N\J-UCG- 1806 and NURE- 1.874. Th'ereforc, the:Generalization Study was not submitted for review with XNCAP- 16168-NP. Rexieision 1. The through-wall cracking frequency value of 2.1 5V-]2 Ior Wolf Creek was calculated using the I*WCF- coyrelations in NUTREG-1806 and the infl onalion in the table below. The beltline mateaial properties in. this tabl were ta.kentrom the.NRC Rtactor Vessel Integrityl atabasc (It's,ID)and the floumee projections were taken from WUAP-16030 E luatlonfIresuYi ed TltermalSh,.ockfbr 1oCree.k May 2003. WCAP- 16168-NP-A June 2008 Revision 2

N-28 Rea*ctor Vessel Beltl Material Properties for.Wolf Creel Major" Materiail Region Description U n4rradiatod RT,ý, rIlwvý4 ID Comrlponent Heat Locatlion Flux Typel*I CU HI P [*Fl M~ethod  ?;urttt' le Type253 Base metal twt*39

                                                                                                ]7 [0t540     0110401                         -I      M.VI 2    R2506-1       Mate            B8759-2            Lo',er           A 5338       00.90      0.570   0.009      0.0   Plant Specl-,              .5"I 3    R2505-2       Plate           C4640.2            Lower           .A 5       o0 0, *    [0040. 0.Ow      10.0   Pla**ISpecilc           2.5 4    R205-2        Plate         NR61 783-1       tntermediate         A 533B       0     .0400,1640 0.0        -20.0   Plrt Spe*rl*r           3.51 5    R2005-.3      Plate         NR11 799-1       lntermeliate         A 533.B       0.050 [0O10       0.007    -. 0    Pian!Specifxc           3.51.

t P-2005-1 Plale NR61 a36-1 Internnediate A ~5 0,00 O 0.C-0 0006 .20.0 P sanl.5peo*c, 3.51 7 101l142A Axial Weld 9"146 Lovrn UinCeI0C0tl 0.0W 0.60 0.005 ,50.0 Plant*Specific 1.58

       $    101-1428   AxialWekld          90146             Lowel         Linde06I        00.40   10 0.. 0.005    -50.0   Plant Specifrj          3,08 9    101-142C   AxCial VWelI        950146            Lowef          tin e.0091      0.040     0.      0.005    -.500   Plan Specitis           308 10   101,124A   AxialWeld           90146       . Irtermediate      Uinde0O91       0,040   J000-M 0.005         50.0   Plan'tSpecifi             i..55 11   101-1248   Axial    eVld       90146          Intermediate       kndeOSI       .0.040     O.60    0,005    -50.0   PlantSpecft             3.08 12   101-124C   Axial Veld          90146         Intermedinte.      Linde 0091n     0,040   Iý00601   0    .00 50.0    Plant Specilic          3.05 13    101-171   Circ Weld           90146           InVLovxr          Linde. 124     Oj005 1         0   007IWI456.0 Plant$perific     J        3 WCAP- 16168-NP-A                                                                                                                                                June 2008 Revision ,2

N-29 The pilot plant 'I'WCF value of 4.67E-9 was also obtained using the TWCF CoWTelatiOns in NUREG-1806. Tlhis value does not match the values in the tables in the WCAP because the valnus in the tables were determ inid using the F-AVOR Code rather thia nthe TWCF correlation based upon maximum values of RT.rrbr the beltline eomponcn.tsý The pilot plaint TWCF values .were recalculated usin,-,t.flic wcI correlations in .NUREi-1874 and are presented in the Table below. 1he WCAJ will be rev1sed to includeit these valucs. IDeaver Valle, Unit I ] Palisaides Oconec Unit i Conditioi_ 60 ElPY __ 60 EFPY x-204 1 RVmAw CF) 247____ 253 ___ q* ___,,__,. R53 253 2 209 231 ~ _ _*7.7 __5 _ iRTMAX-F OFI) 0 I0 -~I0

                      ~.WFS:W4ý49E-09                               1        1.57L-07                     2.23E*-07 TI,~ g m                          7.54E- iT                    7.11E-12            j        S.72E-1t)

I ~3.661"-09 TCF~pj 735l~I 7'5E-lZ~ I TWG . .. 0001E+00 0 00E-00 0.001+ 00 IWC~SOT~-AL1,761, 08 I ,1607 4 2 7 To implement die ,itenic,dinservice inspection istervidl justified in thc.WCAP, a licensee would hav"e to desnonsti ate that the pilot plilnt afialyses are bounding 1he criteria to.be uvaluated to detenninc whether ate bounding are identitied in Table A-I of Appendix arc.pilotplantalves A. (i' the XWR(AP These criteria were selected based on feedback finm the Staff during inectings prior to tdiesubmittal of the WCAP otr reviewv. Dominant PT" Tranmients mn lhe ,RC MTS Rik S a~re

                                                                            -idt' applicable:

Tbie transients evaluated in the W\CAP pilot plant analyses were the PTS transients from the NRC PTS Rkis Re-evaluation. For thiscriterion. it is necessary toWdnemonstratethailhese transients are applicable to a specifi plant. 'At Ithe,fiie Revision 0 6f the W AP was issued. the: Generalization Study had not yet been completed. Thereflore it would have been necessary for each plant t comparemdesign features to determine it fthe pilot plantrPTS transients were, applicable to the specific plant. However. thecGencraliation Study has now. been peifonned and the pilot plant PTS transients have been found to be representatir of all the PWR plants in the domesfic fleet. As stated in the last paragraph in Sectios o874. this study demonstrate that risk-significant PStransients do not have any applreciablle plant specific di 1rerices withinthetpopulation otPWRIs currently operating in theIUnited Statcsl Therefoie, plant specific analyses are no longer needed for this criterion. Through "3! plant f'd~ akn rqec

                         *peciilk-T\\CF               TC) using the co0rrelations fin NLIREG-1874 must be lower value detcntuincd than the pilot plant TWCFWvalue calculated using th rTwCF correlations in NUREG-1874... T'lc
             '1,VCF:is esý`sentially alniasureof the embrittlemeht of the reactor vessel and bydemonstrating that the pilot plant has'a hihbhr TWC1F*vlue'the pilot plant change in risk-calculation is bounding.,:

WCAP-16168-NP-A June 2008 Revision 2

N-30

                  !requency a*ndaeii ofDsig fn              Basis Tranients:

It is necessarv to demonstrate thit the amount of fatigue crack growth considered in the pilot plant analyses is hounding for a specific plant. Since the amount of fitigue craek growth was calutlaed using the d*.*s basis transients, a comparison of design basis transients must be performnnd to ensure that thc assumed number of iealup-coodow\n transientsper. year is also applicable to the specific plant, ( I:addoa, cry'.r

                                         .\.i,gtA'4uitluplc The pilot plant analyses were plerl"ormed assuming a single layer of cladding because the probability of havin, a surface bteaking fliv in multibavcr cladding is much less than that of sime-.iax.er cladding: Si *ce the pilot plant anal ses Were performe With singlc-laver. all plants are bounded by this :parameter and this criteria is documented strictly for informiational proiposes.

Table A-2 provides.additional criteria relative to inspection. The purpose of theInspection 4-'khudelogym.vmbrfIit l.pections* and Number of tmfiatioms I 'ouadfields is discussed in the reponse to Pirtb of RAI 3. The purpose o* the PrOpcekd instxr1con scheufueJralwnce ofplant 1i~f&field is far comparison to the inspection plan. contained in PW ROG letter OGr÷6 356, as discussed in the response to p irt b ofRAAl 11

         ý14, "he ficqucn~y ofPTS challenges isa primary input to the chlange in risk                                  with wstimatesassociatcd extending the inspection intermal 1or reactor vess*cl Swelds iro6n 10 to 20 )yars. The Topical states that the transient fircque.ney results devcioped wrthe PI' S technical basis'arc used in the risk increase calculations in the~'I pical: Rculator*I (iuid         1 1i74 statcs that a probabilistie risk analysis used to support each risk-informed application should be technically adequate, Teclinically adequate is defined" at Oie highest level, as 3an analysis ttiat'is performed correctly, inaiamnnei' coisistent siith accepted practices'.

commensurate.With theI scope and level of detail required to support the rcquested changc a) Pleasedescribe hoWt heTopical proposes; that individual licensees" w ill obtain or develop PTirS transient frequency estimates to use in support of tiiciqi'equcst tar relicf Response. Individual licensees will not he required to obtain or develop PTS transient frequency estimates to like itn support of their request for interval extension. As'discussed in the rissponse to RAI i3, it .was originallv intended that individual licensees would have to compare significant design:features such as PORV capacity ani RWST temperature'to determine if the pilot Iant p1s transients were.applicable to their specific plant, floweverý:since that timethe P'lS Gcneriilization Study has been completcd as summiArxed oin pahcs 3-6 and 3-7 hind at the WCAP Rcport, .s stated in the last paragraph in Section 3.2.1 of NUREG-1874. this"'studv dchionstratc~i that risk-significantl.VS trarsientsldo not havv any appreciable plant-specific differences within the population of PW-Rs currentlyoperating in die United States.. TFurfmtc ioer the overall conclusion from this study is lprovided in tie last.paragrapi iii Section 9.3.3 of NUREG-1806: "the TWCF estimates produced.for hie:detailed analysis plants are

                   .sfficient io charicterize (or bound) the TWCF estimates for the five. generalization plantls and,

thus. by ,inference. PWs in geieralv.

b) Given the response toa,), please propose how the probabilistie risk assessme;nt analvsesthat:will be relied upon to support the relief requests willbe demonstrated to be ofsufficient teclmicnil adequacy so WCAP- 16168-NP-A June 2008 Revision 2

N-31 that there is c(nfidence that the increases in core damage frequency or iisk. caused by the etension of the reactor vessel weld inspcetion interval from I0 to .20 yearsis small. One acceptable approach to asscs technical adequacy is to assciss the analysis against cndorsed standard is desciibed in RG . 200. Response: The conditional probabilities for core damage and large early release assumed in the W\CAP and PtS Risk Ro-evaluation a*c 100% for a tluough-wall crack in the vcsscl. Giien this

                .assumption, the onclusions of fie17     P1S Genleralizaon Study,. ind the response to Part a olfthis R.M. no probabibstie riskoassessmet analyscs will beh.elied upon to support tih. requests for interval exten s ion. hereforic*it will not bCrequired that liccoqscs detnonstiai.e the thecthnical adequacy othe PRA. 'f.is *isonist.ent with,ti. NsR( proposexd voluntary P*IIS Rule, wvhich is not expected to rcquirc fle plaint PRA to satisfy R.G. 1.200 requirements.

15. \rhen iiterascs F due to incrases in neutron fluenc, and its resulting einblittlunment the fractional contributionw to tVVF i'tom difierent flwtypes (e.g., surface-breaking vs cinhedded circumferential s, axial,snldl vs. targe) can chainre substantialls.. Is th.e WCAP analysis applicable to plants whichame T\V(F xitu bsiitintially greaterithan the TWIC of the pilot plants in the WCAIP? Is sop please proxide an example to illustrate the application and specify any TWCF limit to the range of applicabiliiy.

Re'sponii'e 1he pilot.ptaits were chosen With the intent that there *xould be rno domcstic, PWR. plailts wih hi her TW CFsalues Since.the time the pilot plants were- chosen analyses perfonned by MNRResearch using piant (ti1t iidble in RV`IDhaveshown thiat'ther~e ae,severa lI

                 .Westinghouse plants that could have TWTi* values higher than those of the pilot. plant by end.of their operating license. For these plants. additional cx aluation: would be required to demionstrate that, e*een though the TNVCF values are hipher than the pilot plant values, the conclusions from the pilot plant aiialsesvare still applicable Futrthemiore, plants that have implemented the
                 *xctndcd inspection interval will he requirdl to reeValuatb theiri TWCF;alie     iconsistent Wi ith the response to AI* 16. In the event that a planitspecific TWOC value exceedstihe' appropriate NSSS

pilot plant value'as.a result f this reevaluation, iadditional evaluation Would also be requifed.

This discuss ion Will be included as part of the clarification to be added to Appendix A as part of the response to R.0 13.

16. New indutstry experience or inf6rmation mavarise that indicaGs that. t IAe'VCF estimates may need to be reevaluated. For example, licensees that utiize relaiations available under the new PTS rulemaking (50.6 1a. may make changes to dicir plants that could increase the TWCF above thfevalues in the WsCAP pilot plant, (due to increases in neutron fluence and its resulting embritflement) whieh were intended to be bounding examiples. Principle 5 of RG 1.174 is to provide.a monitoring program to assure that parameters critical to the conclusion of aecepiahility remain'at acceptable values during die life offthe change to thenlicense requiremeitsý. What type ofinontbiing and.feedback process is pioposed in the Topical that would call for aire-evaluation of the TXVCF as appropriate to ensure that, over time the validity of ihe analysis demonstrating an acceptable increase in risk is maintaiined'?

Response.: The PWROG proposes that for planits unplementing tile extended iniernal, TXCF be re-evaluated anly time fluence is projeted to increase by more than 10 percent, which isless than. one standard deviation on'the.global .fluence that is input to FAVORII Fluence may be projeced. tio increase as a resultof dcor reloiadirig, core loading pattern, Power.upriting,.or when a surveillance capsule is removed from the reactor vessel and evaluated. 'Ilfis is-consistent with the WCAP-16168-NP-A June 2008 Revision 2

N-32 current approach fbrri-.valuating pricsure-temnperature limit curvs and R'T, ,aiues. The N\CAII will be revised to include this requirement.

17. C'urreit analyses of crack stresses during plant operations indaiter that embedded cracks wxill not grow,*vith time. Beciause the.stafls FAVOR analvsis indicate* thait embedded axial cracks contribute nearly all oft.the TWCF, it follows that the assumption that cmbedded axial cracks do not grow with time isan important modeling assumption that contributes to the small risk increase estimated for extending the itspection inter al. RG 1.174 *rcotmmends address important miod*ling assumnptions by performing sensitivity studies or using qtitative arguments Please diseuss how sensitive the quantitative results of the Change in risk analysis are to theoassumption tihat enedded eracks will not griw ?

Responst: There is no sensitivity to growth of embedded flaws, to suberitical crack girowth relative to embrittled Vtessel failure due to postulated P'TiS tratsients. This lack of sensitivity is based:opon the NRC evaluation described in Section 3.2 "Assumption of No Suberilical Crack Growth." of NU G-1807. ProbabilisticFrocture*,,chamcxs -Mfodels. Poarmoeters,(nd U zray Treatemien ~sevd1in 1A VOR Pvsion 04.1 200 Note that fhis e)aluatio concluded that there isno sigerfic-it suberitie al crack growth o' either surface breaking or elmbedded flaws due to stress conrosioi cracking or fattigue .13caUiSIf embedded flaws are not exposed ito the rimar oolant, .their crack, growth, is.substantially less'than that for surface breaking flaws sbijcted io the same loading. flosse er becausethi highlensitivity of TWCFdcue to.any potential increase in.the site of the enibedded lawss(April 2007 NRC Memo. Developmnen of lI i"sSize Lixt, butiron Yah.befbi Druai~IPrOJeosr Tutle 10. Mfe code of Fedc~lRegukmi.uons (1(X I I? 50.16*ADAM,:MS ,Ml)070950392), periodic inspection every, 20. years is,,prioxsed to ensure no embedded flaw crack growlirihas otccurred. This is beirng done eventitongh the risk analysesin Revisioon I of"WCAP-16168-NIP show that no inspectionstare requited except tie initial one after 10 vears of operation. to satisfy the acceptably smaU change in risk,(LERF) criteria per Regulitory Guide 1.174. WCAP- 16168-NP-A June 2008 Revision 2

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