nrc public meeting summary reportnrc public meeting summary report 11/09/11 – discussion of rpv...

NRC Public Meeting Summary Report 11/09/11 – Discussion of RPV Integrity Issues for Operating Nuclear Power Plants

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NRC PUBLIC MEETING SUMMARY REPORT

Date: November 18, 2011 Meeting Contact: Gary L. Stevens RES/DE/CIB 301-251-7569 [email protected] Subject: CATEGORY 2 PUBLIC MEETING – DISCUSSION OF REACTOR

PRESSURE VESSEL INTEGRITY ISSUES FOR OPERATING NUCLEAR POWER PLANTS

Meeting Date/Time: Wednesday, November 9, 2011 / 4:00 pm Location: American Society of Mechanical Engineers - Boiler Code Week Marriott St. Louis Union Station 1820 Market Street St. Louis, MO 63103 Regency C Room Purpose: The purpose of this meeting was to have technical discussions related to

evaluation of irradiation effects on RPV ferritic materials for operating plants, with particular focus on 10 CFR 50.61a and 10 CFR 50 Appendix G evaluations.

Summary: The announcement and draft agenda for this meeting were posted on the

NRC web site on October 19, 2011. They are available via ADAMS at Accession No. ML112790410.

The final meeting agenda is included in Attachment 1.

Meeting attendance is included in Attachment 2. Material presented at this meeting and referred to in the discussion below

is included in the attachments to this meeting summary. Gary Stevens (NRC) opened the meeting at 4:00 pm with introductions,

followed by statements summarizing the efforts being undertaken in the NRC Office of Nuclear Regulatory Research. These efforts include additional research being performed on reactor pressure vessel (RPV) integrity issues. This public meeting is a continuing effort by the NRC to solicit relevant input from interested parties on this subject. The specific

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purpose of this meeting was to have technical discussions related to the evaluation of the structural integrity of RPVs in operating plants. This meeting is a follow-on to previous public meetings held on this topic on July 26, 2011 (Summary Report available at ADAMS Accession No. ML112170262) and September 15, 2011 (Summary Report available at ADAMS Accession No. ML112790501). The NRC’s research efforts on RPV integrity relate to several documents, and this meeting included technical discussions related to these research activities, with particular focus on 10 CFR 50.61a and Regulatory Guide (RG) 1.161.

Mo Dingler (Wolf Creek Nuclear Operating Company) provided a presentation (see Attachment 3) introducing the objective of the Westinghouse presentation that followed (see Attachment 4). Collectively, these two presentations provided industry recommendations for Alternate Pressurized Thermal Shock (PTS) Rule (10 CFR 50.61a) implementation guidance from the Pressurized Water Reactor Owners Group (PWROG) and the Materials Reliability Program (MRP). In the first presentation, the following was noted:

• The Westinghouse presentation summarized a technical report (white paper) that the MRP is currently finalizing.

• The MRP White Paper presents suggested guidance and resolutions to Alternate PTS implementation issues (criteria and analytical methods) that Industry recommends for the contemplated regulatory guide.

• The MRP White Paper will be provided to the NRC for inclusion with the summary/minutes for this meeting within next week (this report was provided to the NRC on 11/18/11 via e-mail, and is included as Attachment 5).

• The Industry’s goal in providing these recommendations was to help avoid unnecessary burden on utilities and regulators when an applicant requests licensing under 10CFR50.61a.

Nathan Palm (Westinghouse) provided a presentation (see Attachment 4) on the specifics of the industry’s Alternate PTS Rule implementation guidance recommendations. The presentation covered the following:

• Background and Purpose • Recommendations

o Use of Sister Plant Data o Adjustments to ΔT30 values when statistical tests are failed o Criteria for not adjusting ΔT30 values o Direct calculation of Through Wall Cracking Frequency

(TWCF)


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o Plate Flaws vs. Weld Flaws o Solutions for flaw limit deviations

• Summary Additional discussion after the two industry presentations identified the following two items:

• The Industry requested that the NRC include information in the Implementation RG regarding personnel, procedure and qualification guidance for sizing and populating the “bins” in the Alternate PTS Rule flaw tables (Tables 2 and 3 of 10 CFR 50.61a).

• Attendees requested that the Industry’s White Paper include a sample problem for use of sister plant data.

Mark Kirk (NRC) provided a presentation (see Attachment 6) on recommendations for updating RG 1.161 and/or ASME Section XI Nonmandatory Appendix K. The NRC requested the following:

• The NRC is considering two options: o Revise RG 1.161; or o Revise Appendix K; then retire RG 1.161

• The NRC would like input: o Which approach is preferred? o Are the proposed changes appropriate? o Are there any other changes or updates needed beyond

those identified here? • If the preferred approach involves revision of Appendix K

o Are there parties interested in working on this effort? o How is the effort best coordinated within ASME?

There were no other presentations offered, nor were there any comments from any members of the public. Finally, based on the information gathered from the public meetings held this year on RPV integrity issues, as well as feedback from interested stakeholders, the NRC announced that there are no currently planned future public meetings on this topic. Interested stakeholders were encouraged to let the NRC know of any additional need to meet prior to the planned publication of the Alternate PTS Rule Implementation RG next spring. The meeting was adjourned at approximately 5:30 pm.


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Attachments: The following attachments are included with this report: Page No.

Attachment 1: Agenda .................................................................................................... 5 Attachment 2: Attendance Lists ...................................................................................... 6 Attachment 3: EPRI/MRP Presentation ........................................................................ 10 Attachment 4: Westinghouse Presentation ................................................................... 15 Attachment 5: MRP White Paper .................................................................................. 40 Attachment 6: NRC Presentation .................................................................................. 90


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Attachment 1 AGENDA

DISCUSSION OF REACTOR PRESSURE VESSEL INTEGRITY ISSUES FOR OPERATING NUCLEAR POWER PLANTS

Wednesday, November 9, 2011

4:00 p.m. – 7:00 p.m. Location

American Society of Mechanical Engineers - Boiler Code Week :

Marriott St. Louis Union Station 1820 Market Street St. Louis, MO 63103 Regency C Room Purpose of MeetingThe purpose of this meeting is to have technical discussions related to evaluation of irradiation effects on RPV ferritic materials for operating plants, with particular focus on 10 CFR 50.61a and 10 CFR 50 Appendix G evaluations.

:

Agenda

Time

:

Topic Organization Coordinator or

Presenter

4:00 Welcome and Introduction NRC Stevens

4:05 Brief Summary of NRC Research Activities on 10 CFR 50.61a and 10 CFR 50 Appendix G

NRC Stevens

4:15 Additional Recommendations for the 10 CFR 50.61a Regulatory Guide

EPRI Hardin

5:15 Recommendations for ASME Code Section XI Nonmandatory Appendix K and Regulatory Guide 1.161 Regarding Low Upper Shelf Energy

NRC Kirk

6:30 Public Comments TBD TBD

6:40 Next Meeting NRC Stevens

6:45 Summary and Review of Action Items NRC Stevens

7:00 Adjourn


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Attachment 2

ATTENDANCE LISTS

The individuals listed on the following 3 pages attended the meeting in person. The following individuals also participated via teleconference:

Name Organization E-mail Al Butcavage NRC [email protected]

Sarah Davidsaver AREVA [email protected] Carol Nove NRC [email protected] Brian Hall Westinghouse [email protected]

Mike McDevitt Southern California Edison [email protected]

ATTENDANCE LISTfor

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Attachment 3 EPRI/MRP PRESENTATION

Industry Recommendations for Alternate PTS Rule Implementation Guidance (Introduction)

Mo Dingler WCNOC; PWROG, MRP NRC Public Meeting on RPV Integrity Issues for Operating NPPs November 9, 2011

2 © 2011 Electric Power Research Institute, Inc. All rights reserved.

Background

• In a Public Meeting on RPV Integrity Issues in Rockville on July 26, 2011, NRC Research staff solicited Industry input for a regulatory guide being developed for implementation of 10CFR50.61a, the Alternate PTS Rule

• Several generic, non-plant-specific issues & questions on implementation have been identified as Industry has reviewed the Rule and contemplated its use

1. Issues regarding criteria and analytical methods

• The focus of today’s presentations

2. NDE issues regarding vessel inspection

• Addressed in other meetings


Goals

• The next presentation summarizes a technical report (white paper) that the Materials Reliability Program is currently finalizing

– Report presents suggested guidance and resolutions to Alternate PTS implementation issues (criteria and analytical methods) that Industry recommends for the contemplated regulatory guide

– Report will be provided to NRC for inclusion with the summary/minutes for this meeting (within next week)

• Industry’s goal in providing these recommendations is to help avoid un-necessary burden on utilities and regulators when an applicant requests licensing under 10CFR50.61a


Together…Shaping the Future of Electricity


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Attachment 4

WESTINGHOUSE PRESENTATION

1

Alternate PTS Rule (10 CFR 50.61a) Implementation Recommendations Developed for EPRI MRP

Nathan A. Palm Bruce A. Bishop November 9, 2011 St. Louis, MO

2

Agenda ● Background and Purpose

● Recommendations – Use of Sister Plant Data

– Adjustments to ∆T30 values when statistical tests are failed

– Criteria for not adjusting ∆T30 values

– Direct calculation of TWCF

– Plate Flaws vs. Weld Flaws

– Solutions for flaw limit deviations

● Summary

3

Background and Purpose ● Alternate PTS Rule (10 CFR 50.61a) published on January

4, 2010, including: – New embrittlement trend curve

– Surveillance data statistical checks

– ISI flaw limits

● Implementation Regulatory Guide being developed by NRC – NRC requested industry input at meeting on 7/26/11

– Industry has developed a white paper with recommendations for

Implementation R.G. – Will be formally provided following consideration of today’s discussions

4

Sister Plant Data ● 10 CFR 50.61a defines Surveillance Data as

“…data…included but not limited to, surveillance programs at other plants…,” commonly referred to as “sister plant data.”

● For RG. 1.99, Rev. 2 and 10 CFR 50.61 adjustments were made to sister plant measured ∆T30 values due to: – Differences in Cu and Ni content – Ratio of CF values

– Differences in irradiation temperature -

– Heat specific CF and credibility based on average of all measured shift values Values must be on consistent basis

5

Adjustments to Sister Plant Data under 10 CFR 50.61a ● 10 CFR 50.61a is based on larger set of data

● ETC takes into consideration: – Chemistry content explicitly rather than through CF

– Irradiation temp. through time averaged TC

– Fluence and flux

● Residual values (Measured ∆T30 – Predicted ∆T30) are calculated for each point of surveillance data prior to use in surveillance capsule data (SCD) statistical checks

● Adjustment of sister plant measured ∆T30 values is not required

6

Sister Plant Data Recommendations ● Allow exclusion of sister plant data with large differences in

measured material percent weight properties

● Permit a tolerance on TC

– Effect of TC on ∆T30 can be significant

– Values in Eason database were not accurate time weighted values per definition in 10 CFR 50.61a

– Suggested tolerance is ½ the range of cold leg temperatures experienced

Statistical checks can be performed using any TC value in range of plant cold leg temperatures

7

10 CFR 50.61a Surveillance Data Deviation Tests

● If any test fails, an evaluation of the data and its effect on ∆T30 and RTMAX values is required

● No prescriptive methodology for calculating ∆T30 when tests are failed

8

Adjustments to ∆T30 Values ● Application of SCD Tests to the Database for Eason-2007 ETC (ΔT30)

equation development showed they are overly conservative – Since all the SC data were actually used in the development of the ETC

equations, 99% should have passed all the statistical tests.

– Source of conservatism is that same uncertainty for populations of SC data in the database (~60-300 points) is also used for plant samples (3-8 points).

● Therefore, any needed adjustments to ΔT30 and RTMAX for failing the SCD statistical tests should be minimal.

Number of Points N 3 4 6 8 10 12T(0.99,N) 4.54 3.75 3.14 2.90 2.76 2.68T(0.99,10N) 2.46 2.42 2.39 2.37 2.36 2.36Sqrt[F(0.99,N,60)] 6.49 4.43 3.02 2.49 2.20 2.03Sqrt[F(0.99,N,300)] 6.47 4.40 2.99 2.45 2.16 1.99

9

Adjustments to ∆T30 Values to Satisfy Mean and Outlier Deviation Tests ● Calculate the adjustment temperature X to just satisfy the mean or

outlier statistical tests

● Add the value of X to all the predicted SCD values , which would also reduce the residual values by the same amount, and show that all three statistical tests are now satisfied

● Add the same adjustment temperature value of X to the predicted values of ∆T30 and RTMAX for those vessel components with the same material heat as that in the surveillance capsule – Because different ETC equations were used to select the capsule materials,

the adjusted values of RTMAX for the surveillance capsule materials may not be the most limiting for the vessel belt line

10

Adjustments to ∆T30 Values to Satisfy Slope Deviation Tests ● No slope test deviations identified by NRC

● High fluence data evaluated by Westinghouse and no deviations identified

● Potential deviations could be addressed by using an industry consensus high fluence ETC

– Would be based on data from surveillance programs only

● If high fluence effects occur at fluence levels beyond EOLE fluence, adjustments to ∆T30 should not be made

11

Criteria for not Adjusting ∆T30 values ● Less than 3 points of heat-specific data

● Large differences in chemical % weight composition between data points cause failure of SCD tests

● Large differences in irradiation temperature between plants cause SCD test failures

● One point of data at low fluence (< 1 x 1019 n/cm2) causes

failure of the SCD tests

● Slope SCD test failing for fluence values beyond fluence values for end of license extension

12

Direct Calculation of TWCF ● RTMAX limits in 10 CFR 50.61a are based on a risk objective

of 1x10-6 events/yr for through-wall cracking frequency

● Exceptions are for circ weld and axial weld+plate limits – based on more restrictive criteria

● Plate limited plants will most likely reach circ weld limit (312°F) before reaching plate limit (356°F) – RTMAX for welds is dependent upon adjacent materials

– Fluence for plate and circ weld typically equivalent

● Calculation of TWCF and comparison to 1x10-6 events/yr risk objective should be allowed in lieu of RTMAX limits

13

TWCF Correlations – NUREG-1874

● TWCF95-TOTAL = αAWTWCF95-AW + αPLTWCF95-PL + αCWTWCF95-CW + αFOTWCF95-FO

● TWCF95-AW = exp{5.5198*ln(RTMAX-AW – 616) – 40.542}*β

● TWCF95-PL = exp{23.737*ln(RTMAX-PL – 300) – 162.38}*β

● TWCF95-CW = exp{9.1363*ln(RTMAX-CW – 616) – 65.066}*β

● TWCF95-FO = exp{23.737*ln(RTMAX-FO – 300) – 162.38}*β

● αXX is a function of RTMAX-XX

● β is a function of wall thickness

● Forgings not susceptible to under-clad cracking

14

Plate Flaws vs. Weld Flaws ● 10 CFR 50.61a contains separate flaw limits for “weld flaws”

and “plate flaws” ● Technical basis for plate flaw model in PTS Risk Study was

weld flaw model with reduction factors for density and size truncation limit

● Reduced size truncation limit was based on plate material being far removed from the effects of any welding

● Weld ISI volume, including ½ t of adjacent base metal (plates and forgings), was specified by ASME Section XI because of the concern for welding effects on adjacent base-metal flaws

15

Plate Flaws vs. Weld Flaws Recommendations ● Plate flaw limits should only be used and evaluated if and

when RPV base metal far removed from welds is examined by ISI

● Any flaws detected by ISI within + 0.5 inch of the outer boundaries of the ASME Section XI inspection volume can be treated as potential plate flaws

● If any of these potential plate flaws have a depth (TWE) of 0.4 inch or more, they should be evaluated as weld flaws

16

Solutions for Flaw Limit Deviations ● 10 CFR 50.61a requires analyses to demonstrate TWCF <

1x10-6 events/yr if flaw limits exceeded

● PFM should be reserved for the most extreme flaw limit deviations

● Regulatory Guide should contain criteria or methods for determining when deviations in flaw limits will satisfy TWCF < 1x10-6 events/yr

● Criteria or methods would consider flaw size, number of flaws, and level of vessel embrittlement

17

Flaw Deviations for “Low” Embrittlement Plants ● Deviations to flaw limits should be generically allowed for

plants with RTMAX values equivalent to or less than 1% of risk objective (TWCF < 1x10-8 events/yr)

● Maximum weld flaw size (TWE) must be less than 1.925” which was largest size simulated by FAVOR

RTMAX-AW 203°F RTMAX-CW 312°F RTMAX-PL 259°F RTMAX-FO 259°F

RTMAX Values Corresponding to a TWCF of 1 x 10-8 Events per Year

18

Flaw Orientation ● Axial weld flaws contributed 99.99% of TWCF for Palisades

and Oconee PTS pilot plants at 32 and 60 EFPY

● For Beaver Valley 1 pilot plant, plates contributed 28.52% to TWCF with axial flaws controlling

● 93% of the average density for all weld flaws would apply to the maximum density for axial flaws only

● 79% of the average density for all plate flaws would apply to the maximum density for axial flaws only

● Flaw limits in 10 CFR 50.61a should be applied to axial flaws only

19

Large Flaws ● Individual larger flaws (those greater than 0.475” in TWE)

have a larger individual contribution to total TWCF than small flaws

● Based on Palisades PFM results, maximum contribution to TWCF of a single flaw greater than those in the 10 CFR 50.61a weld flaw limits is 4.5 x 10-10 events per year

● A few relatively large flaws exceeding the corresponding size limit in 10 CFR 50.61a should not violate the risk objective of 1 x 10-6 events per year on total TWCF

20

Qualitative Evaluation for a Flaw Limit Deviation 1. Determine Flaw Orientation – If Circumferential

Insignificant effect on TWCF

2. Determine Location of flaw in RV, vertically and azimuthally

3. Determine material properties at flaw location. If located on weld fusion line, consider both plate and weld properties

4. Determine fluence at flaw location

5. Calculate RTMAX for flaw considering location specific fluence and material properties

6. Compare flaw RTMAX value to 1x10-8 RTMAX values

– If less Insignificant effect on TWCF

21

Qualitative Evaluation for a Flaw Limit Deviation (continued) 7. If flaw RTMAX is greater than 1x10-8 RTMAX values

– Compare flaw RTMAX value to RTMAX values for limiting vessel materials (welds vs. welds, plates vs. plates)

– If flaw RTMAX value is at least 5°F less than limiting RTMAX values for the vessel Insignificant effect on TWCF

– Reduction in RTMAX offsets increase in TWCF due to larger flaw

8. If qualitative criteria cannot be satisfied, or multiple flaws exceed limits, quantitative analyses should be performed

22

Quantitative Evaluation for Multiple Flaw Limit Deviations ● Contribution of flaw sizes to TWCF from FAVPOST output

● Assumes probability and frequency of failure is directly proportional to number of axial flaws in the RPV beltline

● Contribution to TWCF by size (TWE) is increased by ratio of ISI flaws to corresponding max. limits from Tables 2 and 3

● Developed worksheet procedure for ISI Flaw Factor and: TWCF95-TOTAL = ISI Flaw Factor * TWCF95-AW + TWCF95-PL + TWCF95-CW With: TWCF95-TOTAL < 1x10-6/year. Where: TWCF95-AW is calculated for RTMAX-AW, TWCF95-PL is calculated for RTMAX-PL and TWCF95-CW is calculated or set equal to 1x10-8/year per equations 3-5 and 3-6 in Section 3.3 of NUREG-1874.

23

Proximity Rules ● Per NUREG/CR-6817, multiple flaws that are combined into

one flaw per the previous ASME Section XI proximity rules are included in the VFLAW Code models

● Combined flaws should not be counted as multiple ISI flaws against the 10 CFR 50.61a flaw limits

24

Summary ● Implementation RG should include adjustments for

surveillance capsule data tests and criteria for not making adjustments

● Treatment of 10 CFR 50.61a ISI flaw limits should be con-sistent with risk-informed technical basis in NUREG-1874 – Application of limits only to axially oriented flaws

– Application of plate or weld flaw limits

– Evaluation procedure if flaw limits are exceeded

● Use of the recommendations in implementation RG should avoid any un-necessary burden on utilities and regulators


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Attachment 5

INDUSTRY WHITE PAPER

MRP Materials Reliability Program_________________________MRP 2011-030 (via email) November 18, 2011 Gary L. Stevens Senior Materials Engineer Component Integrity Branch Division of Engineering One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738 Subject: White Paper Regarding Implementation Guidance for 10 CFR 50.61a, Alternate PTS Rule Dear Mr. Stevens, In a Public Meeting on July 26, 2001, to discuss reactor pressure vessel integrity issues for operating nuclear power plants, staff from the Office of Nuclear Regulatory Research solicited industry input regarding technical issues that will be encountered by an applicant for licensing under 10 CFR 50.61a, “Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock (PTS) Events.” In response, please find attached the EPRI/MRP “White Paper on Alternate PTS Rule (10 CFR 50.61a) Implementation Recommendations.” Under contract to EPRI, Westinghouse Electric Company and other experts generated this white paper for your consideration as NRC RES develops a regulatory guide for 10 CFR 50.61a implementation. We believe that these suggestions will reduce the burden on utilities and regulators during implementation and review of the Alternate PTS Rule. We understand that this white paper will be included with the Meeting Minutes for the November 9, 2011, Public Meeting held in St. Louis, Mo, during which we presented the suggestions from this white paper. Sincerely,

Tim Hardin Project Manager, RPV Integrity Materials Reliability Program cc: Tim Wells, Southern Nuclear Anne Demma, EPRI Maurice Dingler, WCNOC/MRP

ELECTRIC POWER RESEARCH INSTITUTE 3420 Hillview Avenue, Palo Alto, California 94304-1338 ▪ PO Box 10412, Palo Alto, California 94303-0813 ▪ USA

800.313.3774 ▪ 650.855.2121 ▪ [email protected] ▪ www.epri.com

White Paper on Alternate PTS Rule (10 CFR 50.61a) Implementation Recommendations

November 2011

DISCLAIMER OF WARRANTIES AND LIMITATION OF LIABILITIES THIS DOCUMENT WAS PREPARED BY THE ORGANIZATION(S) NAMED BELOW AS AN ACCOUNT OF WORK SPONSORED OR COSPONSORED BY THE ELECTRIC POWER RESEARCH INSTITUTE, INC. (EPRI). NEITHER EPRI, ANY MEMBER OF EPRI, ANY COSPONSOR, THE ORGANIZATION(S) BELOW, NOR ANY PERSON ACTING ON BEHALF OF ANY OF THEM:

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TABLE OF CONTENTS

Executive Summary ................................................................................................................................... 4 1.0 Introduction and Purpose ..................................................................................................................... 5 2.0 Background .......................................................................................................................................... 5 3.0 Recommendation ............................................................................................................................... 10 3.1 Use of Sister Plant Data when Performing Surveillance Data Statistical Tests .......................... 10 3.2 Adjustment of ∆T30 when Mean and Outlier Deviation Tests are Failed on a Plant - or Heat-Specific Basis ...................................................................................................................... 12 3.3 Adjustment of ∆T30 when the Slope Deviation Test is Failed .................................................... 13 3.4 Criteria that can be used to Identify Situations in which Heat-Specific Adjustment to Generic ΔT30 Trends need not be Considered ............................................................................ 15 3.5 Calculation of Through Wall Cracking Frequency (TWCF) and Comparison to Risk Limits if RTMAX Limits are Violated................................................................................................... 16 3.6 Determining Whether ISI Flaws Should be Considered Plate or Weld Flaws when Comparing to Alternate PTS Rule Flaw Limits ........................................................................... 18 3.7 Solutions when Alternative PTS Rule Flaw Limits are not Satisfied .......................................... 19 3.7.1 Qualitative Solutions ........................................................................................................... 22 3.7.2 Quantitative Solutions ......................................................................................................... 26 4.0 Summary ............................................................................................................................................ 29 5.0 References .......................................................................................................................................... 30 Appendix A: Example Evaluations of Surveillance Data with Large Variations in Chemistry .............. 32 Appendix B: Example Worksheets for 10 CFR 50.61a Evaluation of Data ........................................... 34 Appendix C: Selected Output from VFLAW and FAVOR 06.1 Programs ............................................ 39 Appendix D: Slope Deviation Test Evaluation of High Fluence Data .................................................... 44 Appendix E: Calculation of TWCF per Weld Flaw ................................................................................ 48

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EXECUTIVE SUMMARY

The purpose of this white paper is to provide recommendations to the NRC for consideration in their development of an Alternate PTS Rule Implementation Regulatory Guide. The Alternate Pressurized Thermal Shock (PTS) Rule, 10 CFR 50.61a provides an alternative to the current PTS Rule, 10 CFR 50.61. 10 CFR 50.61a includes 1) a new embrittlement trend correlation (ETC) for use in predicting irradiation induced shifts in the nil-ductility transition reference temperature, 2) new requirements for the evaluation of plant and heat specific surveillance data to ensure the applicability of the new ETC, and 3) new requirements for the evaluation of in-service inspection data. To achieve the purpose of this white paper, recommendations are made in seven areas for inclusion into the regulatory guide: 1) Use of sister plant data when performing surveillance data statistical tests, 2) Adjustment of ∆T30 when Mean and Outlier Tests are failed on a plant-specific or heat-specific basis, 3) Adjustment of ∆T30 when the Slope Test is failed, 4) Criteria that can be used to identify situations in which heat-specific adjustment to generic ΔT30 trends need not be considered, 5) Calculation of through wall cracking frequency (TWCF) and comparison to risk limits if RTMAX limits are violated, 6) Determining whether flaws should be considered as plate or weld flaws when comparing to Alternate PTS Rule flaw limits and 7) Qualitative and quantitative solutions when Alternative PTS Rule flaw limits cannot be satisfied. While all of these recommendations may not be required for actual implementation, they certainly can be used for evaluating violations of any of the requirements in the Alternate PTS Rule 10 CFR 50.61a. Therefore, it is requested that these recommendations be incorporated into the regulatory guide to avoid any un-necessary burden on utilities and regulators in implementing and reviewing probabilistic fracture mechanics based risk-informed vessel applications. These applications include the Alternate PTS Rule 10 CFR 50.61a, revision of heat-up and cool-down limits in ASME Section XI Appendix G and the extension of the in-service inspection interval from 10 to 20 years for the PWR Owners Group.

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1.0 Introduction and Purpose The Alternate Pressurized Thermal Shock (PTS) Rule, 10 CFR 50.61a [1], was published on January 4, 2010 and became effective on February 3, 2010. This rule provides an alternative to the current PTS Rule, 10 CFR 50.61 [2], which has been effective for over 20 years. The purpose of this white paper is to provide recommendations to the NRC for consideration in their development of an Alternate PTS Rule Implementation Regulatory Guide. This regulatory guidance will help reduce the resources needed by utilities and the NRC to implement the Alternate PTS Rule, 10 CFR50.61a, and will provide a consistent and acceptable level of safety subsequent to Rule implementation, especially in those instances where alternate evaluations are required to demonstrate compliance with the Alternate Rule. 2.0 Background It is anticipated that the reactor pressure vessel (RPV) at several plants will have difficulty in meeting the requirements of 10 CFR 50.61 before the end of their license renewal period and will implement the alternative requirements of 10 CFR 50.61a. 10 CFR 50.61a provides less restrictive screening criteria than those included in 10 CFR 50.61. However, 10 CFR 50.61a includes 1) a new embrittlement trend correlation (ETC) for use in predicting irradiation induced shifts in the nil-ductility transition reference temperature, 2) new requirements for the evaluation of plant and heat specific surveillance data to ensure the applicability of the new ETC, and 3) new requirements for the evaluation of inservice inspection data. The US NRC has suggested that these new requirements may also be applicable for implementing the risk-informed revision to ASME Section XI Appendix G [12] and the extended inservice inspection interval for the reactor vessel provided for in WCAP-16168-NP-A, Revision 2 [11]. Alternate Rule, 10 CFR 50.61a, defines generic procedures and criteria to ensure compliance with the revised PTS evaluation requirements. If the event these generic criteria cannot be met the Alternate Rule allows additional plant-specific evaluations to be performed to demonstrate that the RPV has adequate resistance to fracture during PTS events. These additional plant-specific evaluations must be submitted to the NRC for approval. To assist in the implementation of the Alternate PTS Rule (10 CFR 50.61a), particularly for those instances where compliance with the generic procedures and criteria cannot be demonstrated, the NRC has begun the development of an “Implementation Regulatory Guide.” As part of the development of this regulatory guidance, the NRC held several meetings with industry and asked for formal input from industry. The two areas of the Alternate

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PTS Rule that are viewed by industry as most needing guidance are the evaluation of surveillance and inservice inspection data. This section summarizes the generic requirements specified in 10 CFR 50.61a for evaluation of the inservice inspection and surveillance results. The limits specified for the number and corresponding range from minimum to maximum size of flaws (through-wall extent or TWE) in the inner 1-inch or 10 percent of the wall thickness, whichever is greater, are listed in Tables 1-1 and 1-2 for welds and base metal, respectively. The basis for the limits in these tables is described in Section 3.6.

Table 1-1 Allowable Number of Flaws in Welds

Through-Wall Extent (TWE) of Flaw (in.)

Maximum number of flaws per 1000 inches of weld length in the inspection volume that are greater than

or equal to TWEMIN and less than TWEMAX TWEMIN TWEMAX

0 0.075 No Limit

0.075 0.475 166.70

0.125 0.475 90.80

0.175 0.475 22.82

0.225 0.475 8.66

0.275 0.475 4.01

0.325 0.475 3.01

0.375 0.475 1.49

0.425 0.475 1.00

0.475 Infinite 0.00

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Table 1-2

Allowable Number of Flaws in Plates or Forgings

Through-Wall Extent (TWE) of Flaw (in.) Maximum number of flaws per 1000 square

inches of inside surface area in the inspection volume that are greater than or equal to TWEMIN

and less than TWEMAX TWEMIN TWEMAX

0 0.075 No Limit

0.075 0.375 8.05

0.125 0.375 3.15

0.175 0.375 0.85

0.225 0.375 0.29

0.275 0.375 0.08

0.325 0.375 0.01

0.375 Infinite 0.00 The surveillance data evaluation requires that three statistical tests be performed. These include the Mean Deviation Test, Slope Deviation Test, and the Outlier Deviation Test. The tests evaluate the residual values of the surveillance data where the residual is defined as follows:

Residual (r) = Measured ΔT30 – Predicted ΔT30 The types of deviations evaluated by each of these tests are shown in Figure 1-1. The limits for the Mean, Slope, and Outlier Deviation Tests are shown in Tables 1-3, 1-4, and 1-5, respectively.

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Figure 1-1

Surveillance Data Statistical Tests Required by 10 CFR 50.61a

Table 1-3 Maximum Heat-Average Residual [°F] for Relevant Material Groups by Number of

Available Data Points

Material group σ [°F] Number of available data points

3 4 5 6 7 8 Welds, for Cu > 0.072 26.4 35.5 30.8 27.5 25.1 23.2 21.7 Plates, for Cu > 0.072 21.2 28.5 24.7 22.1 20.2 18.7 17.5 Forgings, for Cu > 0.072 19.6 26.4 22.8 20.4 18.6 17.3 16.1 Weld, Plate or Forging, for Cu ≤ 0.072 18.6 25.0 21.7 19.4 17.7 16.4 15.3

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Table 1-4

TMAX Values for the Slope Deviation Test

Number of available data points (n) TMAX

3 31.82 4 6.96 5 4.54 6 3.75 7 3.36 8 3.14 9 3.00

10 2.90 11 2.82 12 2.76 13 2.72 14 2.68 15 2.65

Table 1-5 Threshold Values for the Outlier Deviation Test

Number of available data points

(n)

Second largest allowable normalized residual value

(r*)

Largest allowable normalized residual value

(r*) 3 1.55 2.71 4 1.73 2.81 5 1.84 2.88 6 1.93 2.93 7 2.00 2.98 8 2.05 3.02 9 2.11 3.06 10 2.16 3.09 11 2.19 3.12 12 2.23 3.14 13 2.26 3.17 14 2.29 3.19 15 2.32 3.21

Note: r* is the residual divided by the standard deviation σ in Table 1-3

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3.0 Recommendations Recommendations regarding the following aspects of the Alternate PTS Rule will be presented in this white paper for requested incorporation into the new regulatory guide. While all of these recommendations may not be required for actual implementation, they certainly can be used for evaluating violations of any of the requirements in the Alternate PTS Rule 10 CFR 50.61a.

1. Use of sister plant data when performing surveillance data statistical tests required by Paragraph (f)(6).

2. Adjustment of ∆T30 to satisfy Paragraph (f)(6)(vi) when Mean and Outlier Tests are failed on a plant - or heat-specific basis.

3. Adjustment of ∆T30 to satisfy Paragraph (f)(6)(vi) when the Slope Test is failed.

4. Criteria that can be used to identify situations in which heat-specific adjustment to generic ΔT30 trends need not be considered to satisfy Paragraph (f)(6)(vi).

5. Calculation of through wall cracking frequency (TWCF) and comparison to risk limits to satisfy Paragraph (d)(4) if RTMAX limits required by (c)(3) are violated.

6. Determining whether flaws should be considered as plate or weld flaws when comparing to Alternate PTS Rule flaw limits as required by Paragraph (e)(1).

7. Qualitative and quantitative solutions to satisfy Paragraph (e)(4)(i) for instances when Alternative PTS Rule flaw limits cannot be satisfied.

Each of these recommendations is discussed in detail in the following sections.

3.1 Use of Sister Plant Data when Performing Surveillance Data Statistical Tests

Paragraph (6) of the Alternate PTS Rule requires that plants perform three statistical data tests of surveillance data to ensure that an appropriate RTMAX-X value has been calculated for each beltline material. Paragraph (f)(6)(i) of the Alternate PTS Rule requires that “The licensee shall evaluate the results from a plant-specific or integrated surveillance program if the surveillance data satisfy the criteria described in paragraphs (f)(6)(i)(A) and (f)(6)(i)(B)…” According to the Definitions section of 10 CFR 50.61a states that “Surveillance data results means any data that demonstrates the embrittlement trends for the beltline materials, including but not limited to, surveillance programs at other plants with or without surveillance program integrated per 10 CFR 50, Appendix H.” It was determined during a July 26 meeting between NRC and industry that this definition of surveillance data confirms the NRC’s intention to utilize sister plant data in the statistical surveillance data tests.

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In Regulatory Guide 1.99, Revision 2, and the existing PTS rule, the material characteristics are addressed by the chemistry factor term which is based solely on copper and nickel content, or on a fit of the measured ∆T30 values from surveillance data. This chemistry factor is then multiplied by a fluence factor to determine ∆T30. In the determination of chemistry factors using surveillance capsule data (R.G. 1.99 position 2.2), the measured ∆T30 values are multiplied by the fluence factor, summed, and then divided by the sum of the squares of the fluence factors in order to determine a surveillance data-based chemistry factor. The credibility of data is determined based on the deviation from this average. Since the credibility is determined about this average it is essential to make adjustments to the measured ∆T30 values to account for any bias in the data due to differences in irradiation temperature and material properties.

In a February 12, 1998 presentation [3] the NRC made recommendations regarding the treatment of sister plant data. These recommendations addressed the following issues:

1. Adjustments for differing chemical composition between that of the surveillance weld and that of the vessel weld – If the vessel heat best estimate chemistry differs from the surveillance material (matching heat number) best estimate chemistry, the NRC recommendation was to multiply the measured ∆T30 values by a ratio of the vessel chemistry factor to the capsule chemistry factor; the chemistry factors for this adjustment are to be taken from Table 1 of Reg. Guide 1.99, Rev. 2.

2. Adjustment for environmental conditions (temperature) – The recommendation is

that sister plant measured ∆T30 values be increased or decreased by 1°F for every

1°F difference in irradiation temperature between the plant and the sister plant.

The embrittlement trend curve (ETC) equation in 10 CFR 50.61a is dependent upon a greater number of variables and the dependency of ∆T30 on fluence is not as simplistic as it is in the ETC given in Regulatory Guide 1.99, Revision 2, or 10 CFR50.61. The fluence effects are embedded into the ETC rather than being considered with a single multiplier value and, for this reason, a simple ratio process such as that suggested in Reference [3] is not appropriate. Furthermore, in the 10 CFR 50.61a surveillance tests, a residual value is calculated for each specific material and measured value of ∆T30. For the Mean Test, the average is then taken on these residuals (deviations), as opposed to the R.G. 1.99, Revision 2 and 10 CFR 50.61 approach where the average is taken before the computation of the deviations. It should also be noted that the 10 CFR 50.61a ETC directly takes into consideration the material properties and the irradiation temperature in the calculation of ∆T30 and the residual for each point of plant or sister plant data. For these reasons, and since the new correlation is based on a very large database, it is recommended by industry that adjustments not be made to sister plant ∆T30 values for the purposes of performing the surveillance data tests in 10 CFR 50.61a.

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Though it is recommended that adjustments not be made to sister plant surveillance data (∆T30 values), it is recommended that limitations be placed on the variation in chemistry within a weld heat that must be considered. In other words, surveillance data that is from the same weld heat but has very different chemical composition should not be considered as sister plant data. An example of such a weld heat is shown in Appendix A. As shown in Section A.1 the mean surveillance data deviation test failed when the plant specific measured chemistry values are used. In Section A.2, the chemistry percent weight values for Plant B were used for all three plants and the requirements of all three surveillance data deviation tests were satisfied with substantial margin.

Large differences in irradiation temperature can also have significant effects on the calculated ∆T30 values. 10 CFR 50.61a defines TC as a time weighted average of the cold leg temperature under normal full power operating conditions. However, the values used in the development of the ETC may not have reflected an accurate time weighted average for each specific plant. Many plants have implemented power uprates which, in some cases, can result in a significant decrease in TC. If the surveillance data tests cannot be met using exact time weighted average TC values, a tolerance that is equivalent to plus or minus one half of the range of the operating temperatures over the irradiation time should be considered. If the surveillance data tests can be satisfied by using a temperature within this tolerance range on TC, then adjustment to ∆T30 values should not be required. This tolerance accounts for the uncertainty in the actual operating temperature when the largest changes in the embrittlement index occurred.

3.2 Adjustment of ∆T30 when Mean and Outlier Deviation Tests are Failed on a Plant - or Heat-Specific Basis.

Independent peer reviewers of the risk-informed technical basis for the Alternate PTS Rule asked that applicability of plant surveillance capsule data to the embrittlement trend curve (ETC) equations used in NUREG-1874 [4] be shown by those who want to use 10 CFR 50.61a. As a result, NRC developed four statistical tests for evaluating surveillance capsule data for 10 CFR 50.61a in 2008 [5]. Three of these tests, the mean and outlier statistical tests discussed in this section, and the slope statistical test discussed in Section 3.3, were incorporated into 10 CFR 50.61a. The fourth statistical test on uncertainty was not used because none of the surveillance capsule data was expected to fail this test.

Table 3.2-1 shows that when these surveillance capsule data (SCD) statistical tests are applied to the database used for development of the “EONY” ETC [13] (ΔT30) equations, they can be overly conservative. Since all the SCD were actually used in the development of the ETC equations, 99% of the data should have passed all the statistical tests. However, this was not the case. Specifically, eight PWR heats (~7% of all data in the database) failed mean statistical test, including four plate heats, three weld heats and one forging heat. No PWR heats (~1% of all data) failed the slope statistical test. Five

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PWR heats (~4% of all data) failed the outlier statistical test, including two plate heats, and three weld heats. Four of these failed heats also failed the mean test.

The source of this conservatism is that the same uncertainty for populations of SCD in the database (~60-300 points) is also used for a limited number of plant specific surveillance data samples (3-8 points). Table 3.2-2 shows the statistical correction factors at the 99% confidence level. The T Statistic is used for the uncertainty on the mean value while the F Statistic is used for the uncertainty on the variance, which is the squared value of the standard deviation. The T statistic value of 2.33 used in Equation 10 of 10 CFR 50.61a is for a large population of samples where, as shown in Table 3.2-2, a value of 4.54 would be appropriate for a heat specific set of 3 data points. Because of this conservatism, it is recommended that any needed adjustments to ΔT30 and RTMAX for failing the SCD statistical tests should be minimal. Accordingly, the following procedure is recommended for calculating the ΔT30 adjustment to meet the requirements of Paragraph (f)(6)(vi) when either the mean or outlier statistical tests are failed when applying the requirements of 10 CFR 50.61a, Paragraph (f)(6)(v):

1. Using a worksheet similar to the template in Section B.1 of Appendix B, calculate the adjustment temperature ΔTADJ to just satisfy the mean or outlier statistical tests.

2. Add the value of ΔTADJ to all the predicted surveillance capsule ∆T30 values, which would also reduce the residual values (measured – predicted) by the same amount, and show that both the mean and outlier statistical tests are now satisfied. Since the same adjustment value is added to each surveillance data point, the results of the slope statistical test would not change. If the slope statistical test is failed, then the adjustment procedure of Section 3.3 is recommended.

3. Add the same adjustment temperature value of ΔTADJ to the predicted values of

∆T30 and RTMAX for those vessel components with the same material heat as that tested in the surveillance capsules.

Note that because different ETC equations were used to identify the limiting materials per ASTM E185 at the time of capsule fabrication, the adjusted values of RTMAX for the surveillance capsule materials may not be the most limiting for the vessel beltline materials.

3.3 Adjustment of ∆T30 when the Slope Deviation Test is Failed A possible reason why the statistical Slope Test required by 10 CFR 50.61a is not failed with the current SCD could be that the accumulated fluence level of the current surveillance capsule materials is not high enough. However, for the purposes of developing this white paper, data from several plants, with fluence exposures of up to 8.5

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x 1019 were evaluated. The results of these evaluations are contained in Appendix D. No slope test deviations were encountered. In the future, with additional high-fluence surveillance capsule data, it is possible that the new surveillance capsule data could fail the statistical slope test, and possibly the mean and outlier statistical tests as well. The industry has developed a Coordinated Surveillance program [6] in order to obtain high fluence data from power reactors. The intent is to use this data to develop an industry consensus ETC that addresses irradiation induced shifts due to fluence values up to 1 x 1020n/cm2 (E>1Mev). Such an ETC would be used in the future, in the event that the slope test was not satisfied, to calculate the predicted values of ∆T30 for all relevant surveillance capsule data as discussed in Section 3.1.

Table 3.2-1 Heats of Material in Currently Operating PWR Plants That Exhibit Statistically

Significant Deviations [3.2-2]

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Table 3.2-2

Statistical Factors for 99% Confidence Level (α = 1%)

Note: T(0.99,n) = deviations above mean, T(0.99,infinite) = 2.33 and Sqrt[F(0.99,ns,np)] = maximum ratio of sample to population

deviations for statistical equivalency.

3.4 Criteria that can be used to Identify Situations in which Heat-Specific Adjustment to Generic ΔT30 Trends need not be Considered

The following are recommended as criteria for identifying situations when adjustments to the generic ∆T30 values, as calculated using the 10 CFR 50.61a ETC, would not need to be considered:

• The requirements of 10 CFR 50.61a to evaluate surveillance capsule data are only applicable if three or more surveillance data points measured at three or more different neutron fluences exist for a specific material. Therefore, for materials not meeting this requirement, the surveillance data tests could not be performed and there would be no basis for making any adjustment to the ETC predicted values.

• Some weld heats have large differences in material chemistry composition from one plant to another. Adjustments to plant specific ∆T30 values should not be made due to unsatisfactory results from the evaluation of sister plant surveillance capsule data with large differences in material properties.

• Similar to the discussion in the previous item, adjustments to plant specific ∆T30 values should not be made due to unsatisfactory results from the evaluation of sister plant surveillance capsule data with large differences in irradiation temperature.

• In some cases, a heat of material may fail the surveillance capsule data statistical tests due to one data point. If this point of data is associated with a low fluence (< 1 x 1019

n/cm2) and excluding this one point of data allows the tests to be passed, the credibility of that data point should be given further consideration and potentially disregarded.

Number of Points N 3 4 6 8 10 12T(0.99,N) 4.54 3.75 3.14 2.90 2.76 2.68T(0.99,10N) 2.46 2.42 2.39 2.37 2.36 2.36Sqrt[F(0.99,N,60)] 6.49 4.43 3.02 2.49 2.20 2.03Sqrt[F(0.99,N,300)] 6.47 4.40 2.99 2.45 2.16 1.99

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3.5 Calculation of Through Wall Cracking Frequency (TWCF) and Comparison to Risk Limits if RTMAX Limits are Violated

As described in NUREG-1874 [4], the RTMAX based PTS Screening Criteria in 10 CFR 50.61a are based on a risk objective of 1 x 10-6 events per year. Risk was measured using the 95th percentile value of through-wall cracking frequency, TWCF95, which was an output of the FAVOR probabilistic fracture mechanics code. There are several cases where the RTMAX screening criteria do not directly correlate to a TWCF value of 1 x 10-6 events per year.

The circumferential weld screening criteria value is based on a risk metric of 1 x 10-8 events per year. This lower risk limit was somewhat arbitrarily chosen due to their much lower contribution to TWCF. However, this limit may be problematic for plate limited plants. The calculation of RTMAX-PL requires the determination of ∆T30 based on the limiting fluence occurring along the inner diameter of the plate. The calculation of RTMAX-CW requires the consideration of the adjacent plate materials with the determination of ∆T30 based on the limiting fluence occurring along the inner diameter of the circumferential weld. These limiting fluence values, for the plate and circumferential weld, will in most cases be equal or very close. Therefore, plate-limited plants are likely to reach the PTS screening criteria for circumferential welds before reaching the plate screening criteria. Note that this discussion would also apply to vessel forging material that is compliant with Revision 1 of NRC Regulatory Guide 1.43 [10] and therefore not susceptible to under-clad cracking.

The combined screening criteria for plates and axial welds, RTMAX-AW + RTMAX-PL, is based on a simplification of the 1 x 10-6 events/year risk criteria that results in more conservative screening criteria. Plants that are not heavily biased towards either plates or axial welds could be limited by this more conservative criterion. In this case, calculating the actual value of TWCF95-TOTAL could show that the risk criteria is actually satisfied.

NUREG-1874 [4] provides correlations between TWCF95 and RTMAX. It is recommended that a provision be included in the proposed Regulatory Guide for direct calculation of TWCF95-TOTAL using the plant specific RTMAX-XX values, as a method for satisfying Paragraph (d)(4), when the PTS screening criteria are exceeded. The correlations to be used in the calculation of TWCF would be the same as those used in the derivation of the RTMAX screening criteria. The correlation for estimating the total 95th percentile value of TWCF is as follows:

TWCF95-TOTAL = αAWTWCF95-AW + αPLTWCF95-PL + αCWTWCF95-CW + αFOTWCF95-FO

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Where αXX is determined as follows (Note: “xx” is being used as a generic placeholder to represent AW for axial welds, PL for plates, FO for forgings, and CW for circumferential welds):

If RTMAX-xx ≤ 625°R, then αXX = 2.5

If 625ºR < RTMAX-xx < 875ºR, then αXX = 2.5 – (1.5/250)(RTMAX-xx – 625)

If RTMAX-xx ≥ 875ºR, then αXX = 1 and TWCF95-XX values are calculated as follows:

TWCF95-AW = exp{5.5198*ln(RTMAX-AW – 616) – 40.542}*β for axial welds,

TWCF95-PL = exp{23.737*ln(RTMAX-PL – 300) – 162.38}*β for plates,

TWCF95-CW = exp{9.1363*ln(RTMAX-CW – 616) – 65.066}*β for circumferential welds and,

TWCF95-FO = exp{23.737*ln(RTMAX-FO – 300) – 162.38}*β

+η*{1.3 x 10-137*100.185*RTMAX-FO}*β for forgings. In the above equations, RTMAX-XX values shall be expressed in degrees Rankin. Furthermore, if the subtraction from which the natural logarithm is taken results in a negative value, then TWCF95-XX is equal to zero.

η is equal to “0” for ring-forged vessels fabricated compliant with Regulatory Guide 1.43 [10] and equal to “1” for ring-forged vessels not fabricated compliant with Regulatory Guide 1.43. β is determined as follows (TWALL dimensions are in inches):

If TWALL ≤ 9½, then β = 1

If 9½ < TWALL < 11½, then β = 1+8(TWALL- 9½)

If TWALL ≥ 11½, then β = 17 Where TWALL is the thickness of the RV wall (inches), including the cladding.

If the calculated plant-specific TWCF95-TOTAL value is less than 1 x 10-6, then the screening criteria in the Alternate PTS Rule, 10 CFR 50.61a, should be considered to be satisfied.

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3.6 Determining Whether ISI Flaws Should be Considered Plate or Weld Flaws when Comparing to Alternate PTS Rule Flaw Limits Independent peer reviewers of the technical basis for the Risk-Informed PTS Rule in NUREG-1874 [4] requested that the applicability of the flaw distribution data used by the FAVOR Computer Code be shown to be applicable by comparison to results from a specific plant vessel qualified in-service inspection (ISI). In response to this request, NRC completed a Memorandum [7] on the technical basis used for development of flaw size distribution limits in April 2007. When the limiting numbers of weld flaws that NRC developed in the Memorandum are compared with the number measured in the Shoreham vessel they were found to higher by a factor from 2.6 to 7.6, depending on their size. This is because on average, the FAVOR probabilistic fracture mechanics code simulated a larger number of flaws for the welds in the Palisades vessel beltline. Because there was a very limited amount of detailed flaw weld data available (about 65 inches examined vs. almost 60,000 inches in the U.S. PWR fleet of vessel), 1000 flaw distributions with large uncertainties were used with other conservative assumptions, like the higher flaw densities from Shoreham being combined with the larger flaw sizes in the Pressure Vessel Research User Facility (PVRUF) vessel. This was done to account for the statistical uncertainties, like those in Table 3.2-1 for extrapolating results from a small sample size to a much larger population. The NRC Memorandum [7] also says that similar results were also found for the plate flaws simulated by FAVOR for the Beaver Valley Unit 1 reactor vessel beltline.

Based upon this work by NRC, Tables 2 and 3 on ISI flaw limits for welds and plates (Tables 1-1 and 1-2, of this white paper), respectively, were then incorporated into the proposed Alternate PTS Rule 10 CFR 50.61a. Industry comments on these proposed flaw limits requested a meeting with NRC to further discuss the concerns with the NRC limits and their technical basis. However, the Alternate PTS Rule, 10 CFR 50.61a [1] was published without a meeting with industry to discuss these concerns with the proposed flaw limits. Even though some of the following concerns seem to counteract one another, each one will be specifically addressed in the Section 3 recommendations of this White Paper for the new Regulatory Guide on application of the 10 CFR 50.61a flaw limits:

1. The FAVOR Code flaw distribution output that was used (Palisades for welds,

Beaver Valley Unit 1 for plates) in the technical basis for the flaw limits in Tables 2 and 3 of 10 CFR 50.61a, respectively, is only the average of the 1000 flaw distributions generated by the VFLAW Computer Code for both weld and plate embedded flaws that are input to FAVOR as described in Revision 1 of NUREG/CR-6817 [8]. This means that for each 1000 vessel simulations by the FAVOR Code, 499 of the weld and plate flaw distributions are above average and would exceed the technical basis.

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2. Conversion of the embedded flaw size (through-wall extent or TWE) bin width from one percent (1%) of the total wall thickness in FAVOR to 0.05 inch in the technical basis did not consider the fact that the FAVOR Code used to calculate the through-wall cracking frequency (TWCF) for the screening limits in Table 1 of 10 CFR 50.61a only simulates the largest flaw size in each bin.

3. The flaw limits in Tables 2 and 3 of 10 CFR 50.61a are based upon the total number of simulated flaws with no distinction being made between the axially oriented flaws that contribute to failure and the TWCF and the circumferentially oriented flaws that do not fail (See discussion in Section 3.7.1). This is non-conservative in that if all the limiting number of flaws were axially oriented, then the TWCF calculated by the FAVOR Code would be about twice that used in NUREG-1874 [4], which is the technical basis for the screening limits in Table 1 of 10 CFR 50.61a.

4. The flaw limits for welds in Table 2 of 10 CFR 50.61a neglect the FAVOR Code simulation of much larger weld flaws well above (up to 1.925 inches) the truncation limit (0. 475 inch) used in the NRC technical basis document [7]. For the FAVOR Code category 2 flaw sizes, which are the basis for the NRC flaw limits, Section B.2 of Appendix B and Section C.4 of Appendix C show that the flaw sizes above the truncation limit contribute almost 50% to the calculated value of TWCF at 60 EFPY and should be considered in the evaluation of their effects on TWCF.

5. The NRC technical basis for the number of plate flaws is all flaws simulated by the FAVOR Code in the total vessel beltline inside surface area of almost 80,000 square inches normalized for 1,000 square inches. However, these plate flaws are not being examined during in-service inspections. Per the ASME Code Section XI requirements, the area that is inspected is adjacent to the welds because of the concern for large flaws due to the welding process, not for flaws resulting from just the fabrication of the plates or forgings per Revision 1 of NUREG/CR-6817 [8].

6. One FAVOR Code run [9] for a plant-specific flaw exceeding the plate size limit gave a TWCF many orders of magnitude below the risk limit of 1.0E-06/year. While this flaw was located in an area with low embrittlement of the vessel material, this area could still be of concern for failure due to large plate flaws that exceed the plate flaw size limits.

A very conservative definition of what constitutes “plate material” has been used for some 20 applications of the methodology for the PWR Owners Group Program on RPV

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ISI Interval Extension per Revision 2 of WCAP-16168-NP-A [11] (which also uses the FAVOR Code PTS results, NUREG-1874 [4] and the 10 CFR 50.61a ISI flaw limits). Specifically, all the flaws in the ISI Volume per ASME Section XI that were not in the weld itself were considered to plate flaws. With this conservative definition of plate flaws, some plants violated the plate size limit, but – as indicated in concern 6 discussed above – the FAVOR Code showed essentially no effect on the risk metric of TWCF. In general, the failure of the smaller plate flaws should be biased toward the regions of the vessel with the higher levels of embrittlement, but for the smaller number of much larger plate flaws, failure could occur at vessel locations with lower levels of embrittlement. Therefore, this was only an indication that there could be excess conservatism in this area that led to an assessment of how plate flaws were being evaluated for application of 10 CFR 50.61a.

The technical basis for the plate flaw model was a weld flaw model with reduction factors for density and size truncation limit per from NUREG/CR-6817 [8]. The reduced size truncation limit was based on plate material being far removed from the effects of any welding. Specifically, in Section 7.1 of NUREG/CR-6817 [8], the VFLAW Code plate flaw models only include flaws originated from the production processes for plates and forgings. The flaws with the largest depth (TWE) were from lack of fusion between the weld and base metal and were truncated at 25 mm (~1 inch) for SAW and SMAW and 50 mm (~2 inches) for repair welds, which is approximately 22% of the total RPV beltline wall thickness for Palisades. For plate flaws, the VFLAW Code truncation limit was only 11 mm (0.433 inch), which is about 5% of the wall thickness for Beaver Valley Unit 1. It was also noted that these truncation limits are twice the size of that measured by PNNL, which included data from PVRUF and cancelled vessels for Shoreham, Hope Creek and River Bend. For the NRC Memorandum [7], the basis for the plate limits was Beaver Valley Unit 1 where the 0.4 inch truncation limit is consistent with the 0.433 inch limit used by the VFLAW Code to generate the 1000 input flaw distributions for FAVOR.

Since the weld ISI volume, including ½ the wall thickness of adjacent base metal (plates and forgings), was specified by ASME Section XI because of the concern for welding effects on adjacent base-metal flaws, the initial recommendation was that the plate flaw limits in Table 3 of 10 CFR 50.61a should only be used and evaluated when RPV base metal far removed from welds is examined by ISI. However, 10 CFR 50.61a specifically requires that plate flaws within the inspection volume be evaluated. Therefore, the following procedure is now being recommended:

1. Any flaws detected by ISI within + 0.5 inch of the outer boundaries of the ASME Section XI inspection volume can be treated as potential plate flaws.

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2. If any of these potential plate flaws have a depth (TWE) of 0.4 inch or more, they would need to be evaluated as weld flaws. This requirement is needed to ensure that the small flaws due to fabrication of the plates and forgings, which is what was simulated in the FAVOR probabilistic fracture mechanics analysis code, have not been expanded to larger sized flaws by the effects of welding, which was simulated by FAVOR as weld flaws. Also note that the 0.375 inch limit on plate flaw size (TWE) in Table 3 of 10 CFR 50.61a is consistent with the 0.4 inch limitation because the next bin size of 0.425 inch would have exceeded it.

3. For the quantitative evaluation discussed in Section 3.7.2 when the plate flaw limits from Table 3 of 10 CFR 50.61a are exceeded, the plate area in square inches would be two times the inspected weld length.

Note that step 2 is required to be consistent with how the weld and plate flaw distributions were generated for input to the FAVOR Code per Revision 1 of NUREG/CR-6817 [8] and used to generate the PTS risk results (TWCF) in NUREG-1874 [4].

3.7 Solutions when Alternative PTS Rule Flaw Limits are not Satisfied

As previously stated, 10 CFR 50.61a, Paragraph (e), requires the evaluation of inservice inspection data. One of the most significant aspects of this requirement is the evaluation of inservice inspection findings against the flaw limits in Tables 2 and 3 of 10 CFR 50.61a. If the flaw limits in these tables cannot be satisfied, 10 CFR 50.61a, Paragraph (e)(1) requires that the licensee perform analyses required in Paragraph (e)(4) and Paragraph (e)(5) to demonstrate that the reactor vessel will have a TWCF of less than 1 x 10-6 per reactor year. This requirement seems to imply that a plant specific probabilistic fracture mechanics analysis be performed similar to the PTS pilot plant analyses documented in NUREG-1874 [4]. Such an analysis should only be required for the most extreme deviations to the flaw limits. The proposed Regulatory Guide should provide criteria or methods for determining whether the deviation in flaw limits will have a detrimental effect on the applicability of the PTS Screening criteria in Table 1 of 10 CFR 50.61a (i.e., would cause plant to be in excess of the 1 x10-6 events per year risk objective).

There are several scenarios that are envisioned in which flaw limit deviations could occur and the potential means for addressing the deviation should vary depending on the scenario. Some of the potential scenarios are as follows:

• Vessel with Low Embrittlement Exceeds Allowable Number of Flaws

• Vessel with High Embrittlement Exceeds Allowable Number of Flaws

• Vessel with Low Embrittlement Exceeds Maximum Flaw Size

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• Vessel with High Embrittlement Exceeds Maximum Flaw Size

One initial recommendation is to generically allow deviations to the flaw limits for plants that have low embrittlement. Low embrittlement can be defined as having RTMAX values that are equivalent to or less than a TWCF of 1% of the PTS risk objective, or approximately 1 x 10-8 events per year. These plants would still have high enough fracture toughness of the vessel material so that even large increases in the number of flaws above the 10 CFR 50.61a limits would not result in reaching the 1 x 10-6 events per year PTS risk objective. Two conditions on this generic exception would be that 1) each flaw identified in the in-service inspection would need to have a through-wall extent (TWE) of less than 1.925”, which, as discussed in Section 3.6, was the size of the largest flaw simulated by the FAVOR Code for the vessels in the PTS Risk Reevaluation pilot plants, and 2) the flaw would need to satisfy the requirements of ASME Section XI, IWB-3500. Therefore, plants meeting this condition and having RTMAX values equal to or less than those in Table 3.7-1 should not need to perform additional evaluations to demonstrate that the plant specific vessel TWCF due to PTS is less than 1 x 10-6 events per year.

Table 3.7-1 RTMAX Values Corresponding to a TWCF of 1 x 10-8 Events per Year

RTMAX-AW 203°F

RTMAX-CW 312°F

RTMAX-PL 259°F

RTMAX-FO 259°F Note that the screening criteria for circumferential welds is already based on a TWCF of 1 x 10-8 events per year. Other values were calculated using equations in NUREG-1874. Values are applicable for thicknesses up to 9.5”.

For plants that do not satisfy these criteria, additional evaluation would be required. For the scenarios in which the number of allowable flaws is exceeded, a quantitative approach is more appropriate. For scenarios in which the maximum flaw size is exceeded, a qualitative approach is more appropriate. Both qualitative and quantitative recommendations for addressing flaw limit table deviations are provided below for inclusion in the proposed Regulatory Guide.

3.7.1 Qualitative Solutions

Flaw Orientation

The output of the FAVPFM Module of the FAVOR Code provides statistics on the contribution of different flaw types to the conditional probabilities of initiation (CPI) and failure (CPF). These statistics are reported for the weld and plate materials, for the three

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flaw categories (Category 1 = surface flaws, Category 2 = flaws with crack tips in inner 1/8th of the thickness, and Category 3 = flaws with crack tips between the inner 1/8th and 3/8th of the thickness), and flaw orientations. As stated in the executive summary of NUREG-1874, “Axial flaws, and the toughness properties that can be associated with such flaws, control nearly all of the TWCF.” This statement is supported by the results shown in Table 3.1 of NUREG-1874. This table shows that for Palisades and Oconee at 40 and 60 years 32 and 60 EFPY, the axial welds, in which all flaws are axially oriented, contribute over 99.99% of the TWCF. All flaws in the axial welds are axially oriented. For Beaver Valley Unit 1, Table 3.1 of NUREG-1874 shows that the plates contribute up to 28.52% of the TWCF. However, it is the axial flaws in the plates that are providing this contribution. Similar results for 60 EFPY are shown in Sections C.4 and C.5 of Appendix C.

A worksheet evaluation in Section B.5 of Appendix B of the weld flaw density for 70,000 Palisades RPV simulations by FAVOR 06.1 relative to the mean and maximum weld flaw distributions of the 1000 distributions generated by VFLAW indicated the following:

• The output average weld flaw distribution from FAVOR is within 0.5% of the mean distribution from VFLAW in Section C.2 of Appendix C.

• For the size range of the 10 CFR 50.61a limits, the maximum VFLAW distribution in Section C.1 of Appendix C (99.9%) has about twice (2.08) as many flaws as the mean.

• For the 70,000 RPV simulations, 45% of the weld flaws are axial.

• The net effect of the previous two items above (2.08 times 45%) is that 93% of the average density for all weld flaws would apply to the maximum density for axial flaws only.

Because the weld flaw limits in Table 2 of 10 CFR 50.61a on the average number of axial and circumferential flaws is within 7% of the maximum number of axial flaws, these same weld flaw limits can be applied to the maximum number of axial weld flaws in any one vessel.

A similar worksheet evaluation in Section B.6 of Appendix B of plate flaw density for 74,000 Beaver Valley Unit 1 RPV simulations by FAVOR 06.1 relative to the mean and maximum weld flaw distributions of the 1000 generated by VFLAW indicated the following:

• The output average weld flaw distribution from FAVOR is within 0.2% of the mean distribution from VFLAW in Section C.3 of Appendix C.

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• For the size range of the 10 CFR 50.61a limits, the maximum VFLAW distribution in Section C.3 of Appendix C (99.9%) has about twice (1.58) as many flaws as the mean.

• For the 74,000 RPV simulations, 50% of the plate flaws are axial.

• The net effect of the previous two items above (1.58 times 50%) is that 79% of the average density for all plate flaws would apply to the maximum density for axial flaws only.

Because the plate flaw limits in Table 3 of 10 CFR 50.61a on the average number of axial and circumferential flaws is within 21% of the maximum number of axial flaws, these same plate flaw limits can be applied to the maximum number of axial plate flaws in any one vessel.

Also the maximum value is actually a 99.9 percentile value since 999 of the 1000 flaw distributions would be equal to or less than the distribution with the maximum values. Use of this upper bound value is consistent with using upper 99% bound values, instead of mean or average values, to evaluate the acceptability of RPV heat-specific surveillance capsule data per 10 CFR 50.61a, as discussed previously in Section 3.2.

In summary, the flaw limits in Tables 2 and 3 of 10 CFR 50.61a should be applied to axial flaws only. This proposed approach addresses what the real technical concern should be, that is, the maximum number of axial flaws that could lead to potential vessel failure. Large Flaws

Individual larger flaws (those greater than 0.475” in TWE) have a larger individual contribution to total TWCF than small flaws. In order to determine the contribution to TWCF on a per flaw basis, the results from an evaluation of the Palisades pilot plant vessel at 60 EFPY were analyzed. The percent contribution to TWCF of each of the individual weld flaw sizes was obtained from the FAVPOST output. The number of flaws of each size that were simulated was obtained from the FAVPFM output. The number of flaws of each size was divided by the number of vessel simulations to determine the number of flaws simulated per vessel. The percent contribution was then divided by the number of flaws simulated per vessel for each flaw size to determine the contribution of each flaw as a fraction of total TWCF. The TWCF for each flaw was then determined by multiplying this fraction by the total 95th percentile TWCF. The results of this evaluation are shown in Appendix E. As shown in these results, a single flaw of up to 0.875” in TWE can be tolerated without exceeding the risk objective of 1 x 10-6 events per year on total TWCF provided that a significant number of flaws does not exist in the

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smaller TWE flaw sizes. Multiple flaws and larger sized flaws can still be evaluated in more detail using the quantitative procedures in Section 3.7.2.

Flaw Location

As was discussed in Section 3.5, the TWCF of a particular vessel can be closely correlated to the RTMAX values for each of the particular product forms in a given reactor vessel. In other words, the limiting material controls the likelihood of reactor vessel failure due to PTS. Therefore, the flaws in the limiting material are important contributors to the TWCF and depending on the RTMAX values of the remaining materials, the flaws in the less embrittled materials may have a much less significant contribution to TWCF, as discussed below.

Qualitative Assessment Process

Based on the discussion above regarding flaw orientation, size and location, the following is a process that is recommended for addressing a flaw that exceeds the maximum size allowed in the flaw limit tables but is less than 0.875” in through-wall extent (largest TWE evaluated in the PTS pilot plant studies).

1. Determine orientation of flaw. If the flaw is circumferentially oriented, the flaw can be considered to have an insignificant effect on the total TWCF.

2. Determine location of flaw in the reactor vessel with respect to the reactor vessel core, both vertically and azimuthally.

3. Determine material properties at the flaw location. If the flaw is located on a weld fusion line, the properties of both the weld and plate should be considered to determine which material is more limiting.

4. Determine the fluence at flaw location using detailed fluence information, as opposed to using the peak inner diameter fluence.

5. Calculate flaw specific RTMAX value based on location specific fluence and material properties.

6. If the RTMAX value of the flaw is less than or equal to the appropriate RTMAX value in Table 3.7-1, the flaw can be considered to have an insignificant effect on the total TWCF.

7. If the RTMAX value of the flaw is greater than the appropriate RTMAX value in Table 3.7-1, the RTMAX value for the flaw location should be compared to the RTMAX values for limiting vessel materials (welds vs. welds and plates vs. plates). If it can be shown that the RTMAX value of vessel material where the flaw is

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located is at least 5°F less than the limiting RTMAX values for the vessel, then the flaw can be said to have an insignificant effect on the total TWCF. The reduction

in TWCF associated with this 5°F reduction in RTMAX offsets the largest possible increase in TWCF due to the larger flaw.

If the criteria of this process cannot be satisfied or multiple flaws exceed the limits, then more detailed quantitative analyses as recommended in the following section should then be performed.

3.7.2 Quantitative Solutions

If the maximum number of allowable axial flaws per Tables 2 and 3 of 10 CFR 50.61a is exceeded in multiple locations based upon plant-specific qualified ISI results per ASME Section XI, then it is recommended that the 95th percentile value of total TWCF for this vessel be estimated and shown to be acceptable – that is, less than the risk limit of 10-6 per year. The recommended estimation and evaluation is based upon the following considerations.

1. The RTMAX-XX embrittlement metrics and their contribution to TWCF from NUREG-1874 [4] as described previously in Section 3.5.

2. The increase in the axial weld flaw contributions is based upon several factors:

• The ratio of the number of axial ISI flaws found in a particular bin to the maximum number of axial flaws allowed for that bin in Tables 2 and 3 of 10 CFR 50.61a for weld and plate flaws, respectively. This ratio is defined as being at least unity, and can only be made to exceed unity based on the ISI result.

• The contribution of flaw sizes to failure and TWCF are taken from FAVPOST output for the basis plants (Palisades for weld flaws and Beaver Valley Unit 1 for plate flaws), where it is given for the maximum flaw depth (TWE in inches) in each bin.

• If the maximum number of axial flaws is zero in a given bin, then the full contribution from FAVPOST is used for each axial ISI flaw in that bin to calculate the increase ratio.

This recommended procedure is demonstrated in several example worksheets in Appendix A. Section B.2 shows the calculation of the percent increase for axial ISI weld flaws while Section B.3 shows the calculation of the percent increase for axial ISI plate flaws. Note that per 10 CFR 50.61a, plate flaw limits are also applicable to flaws in forgings that are not susceptible to under-clad cracking. Finally, Section B.5 shows the

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conversion of the calculated percent increase for weld and plate flaws to the corresponding ISI Flaw Factor (IFFXX) and the calculation and evaluation of the risk metric for TWCF per the following equation:

TWCF95-TOTAL = IFFAW * TWCF95-AW + IFFPL * TWCF95-PL + TWCF95-CW With: TWCF95-TOTAL < 1x10-6/year. Where: TWCF95-AW is calculated for RTMAX-AW,

TWCF95-PL is calculated for RTMAX-PL and

TWCF95-CW is calculated or set equal to 1x10-8/year.

Note that for these worksheet examples in Appendix B, the plant specific ISI results and RTMAX (RT-max) input values in the shaded cells are only examples to demonstrate use of the procedures for quantitatively evaluating exceeding the ISI flaw limits in Tables 2 and 3 of 10 CFR 50.61a.

With regard to the recommended procedures, please note the following considerations:

1. The procedures are based upon the assumption that the probability and frequency of failure is directly proportional to number of axial flaws in the RPV beltline.

2. The cumulative number of flaws given in Tables 2 and 3 of 10 CFR 50.61 had to be converted back to the allowable maximum number of flaws in each bin for comparison with the plant ISI data.

3. The percent TWCF contribution by flaw size (TWE) was obtained from the FAVPOST output for Palisades and Beaver Valley 1 at 60 EFPY from the PWROG ISI Interval Extension Program [8] as shown in Sections C.4 and C.5 of Appendix C, respectively. Similar results would be expected for the corresponding ORNL FAVOR 06.1 runs at 60 EFPY for the results in NUREG-1874 [4]. 60 EFPY was selected as being more representative of RPV conditions after license renewal operation to 60 or 80 calendar years. If much higher EFPY values near the PTS screening limits had been used for the evaluation, as was done in the NRC Memorandum [7], then there may not have been much margin to allow for any increases in TWCF due to exceeding the flaw limits.

4. Finally, per the excerpt below from NUREG/CR-6817 [8], multiple flaws that are combined into one flaw per the ASME Section XI proximity rules are included in the VFLAW Code models and should not be counted as multiple ISI flaws. Note that the recent changes to the ASME proximity rules should not be used in this case because they would be inconsistent with the prior proximity rules that were used in the technical bases for the FAVOR Code flaw distribution input.

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Also note that the recommended worksheet procedures in Sections B.2 for welds, B.3 for plates and B.5 for TWCF in Appendix A are conservative for the following reasons:

• Difference in FAVOR flaw size bin widths and those in the APTRS max. limits in that the larger percent contribution of the two bins is used,

• Contributions of any axial ISI flaws that exceed the 10 CFR 50.61a weld size limit of 0.475” are also included.

• If the calculated maximum number of flaws for the plant-specific vessel geometry (weld length or plate area) is less than 0.5, then it is set to zero.

• The TWCF is calculated assuming that all the flaws that exceed the maximum limits of 10 CFR 50.61a are located in the most embrittled region (highest RTMAX value) for that RPV beltline component.

• No credit is taken for the number of axial flaws from ISI being well below the number allowed by 10 CFR 50.61a.

Therefore, passing the worksheet procedure in Section B.5 of Appendix B by showing the calculated total TWCF95 is less than 1.0 x 10-6 per year should certainly be acceptable. However, if it is does not pass, then a more detailed evaluation that eliminates some of the above conservatisms could also be used to show acceptability of exceeding the 10 CFR 50.61 a flaw limits.

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4.0 Summary

It is requested that NRC Research incorporate the recommended adjustments to ∆T30 and RTMAX for vessel materials with surveillance capsule data that fail the mean and/or outlier tests and the recommended treatment for failing the slope test that are described in this White Paper into the Draft Regulatory Guide on Application of the Alternate PTS Rule.

It is also requested that the recommended treatment of 10 CFR 50.61a ISI flaw limits be consistent with and supported by how the risk-informed technical basis in NUREG-1874 was developed be incorporated into the new regulatory guide. This includes:

• Application of limits only to axially oriented flaws,

• Application of plate or weld flaw limits,

• Evaluation procedure if flaw limits are exceeded.

While all of these recommended treatments may not be required for actual implementation, they certainly can be used for evaluating violations of any of the requirements in the Alternate PTS Rule 10 CFR 50.61a. Use of these recommended treatments should avoid any un-necessary burden on utilities and regulators in implementing and reviewing FAVOR based risk-informed RPV applications, such as requiring any additional FAVOR PTS analyses with plant-specific flaw distributions.

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5.0 References 1. U.S. Nuclear Regulatory Commission Code of Federal Regulations, 10 CFR Part

50.61a, “Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events,” Volume 75, No. 1, dated January 4, 2010, effective February 3, 2010.

2. U.S. Nuclear Regulatory Commission Code of Federal Regulations, 10 CFR Part

50.61, “Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events,” Federal Register, Volume 60, No. 243, dated December 19, 1995, effective January 18, 1996.

3. U.S. Nuclear Regulatory Commission Presentation, “Generic Letter 92-01 and RPV Integrity Assessment,” Wichman, et. al., February 12, 1998.

4. U.S Nuclear Regulatory Commission, NUREG-1874, Recommended Screening

Limits for Pressurized Thermal Shock (PTS), 2007 (ADAMS: ML070860156). 5. U.S Nuclear Regulatory Commission, Report, Statistical Procedures for Assessing

Surveillance Data for 10 CFR 50.61a, June 2008 (ADAMS: ML081290654). 6. Electric Power Research Institute, Materials Reliability Program: Coordinated PWR

Reactor Vessel Surveillance Program (CRVSP) Guidelines (MRP-326), Draft Report, August 2011.

7. U.S. Nuclear Regulatory Commission, Memorandum from Elliot to Mitchell,

Development of Flaw Size Distribution Tables for Draft Proposed Title 10 of the Code of Federal Regulations (10 CFR) 50.61a, April 3, 2007 (ADAMS: ML070950392).

8. U.S. Nuclear Regulatory Commission, NUREG/CR-6817, Revision 1, A Generalized

Procedure for Generating Flaw Related Inputs for the FAVOR Code, October 2003 (ADAMS: ML051790410).

9. U.S. Nuclear Regulatory Commission, Memorandum from EricksonKirk to Purtscher, Probabilistic Assessment of Flaw in Support of Calvert Cliffs Unit 2 Relief Request ISI-020, January 14, 2009 (ADAMS: ML090370140).

10. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.43, Revision 1, “Control of Stainless Steel Weld Cladding of Low-Alloy Steel Components”, March 2011.

11. Westinghouse Electric Company, WCAP-16168-NP-A, Rev. 2, Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval, June 2008 (ADAMS: ML082820046).

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12. Risk-Informed Method to Determine ASME Section XI Appendix G Limits for Ferritic Reactor Pressure Vessels: An Optional Approach Proposed for ASME Section XI Appendix G, MRP-250 and BWRVIP-215NP, EPRI, Palo Alto, CA: 2009, 1016600.

13. Eason, E.D. et. al., “A Physically Based Correlation of Irradiation-Induced Transition

Temperature Shifts for RPV Steels,” ORNL/TM-2006/530, Oak Ridge National Laboratory, November 2007 (ADAMS: ML081000630).

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Appendix A

Example Evaluations of Surveillance Data with Large Variations in Chemistry

A.1 Example of Mean Deviation Test for Weld Heat with Large Variations in Sister Plant Chemistry Weld Metal - W5214

Number of Data Points => 9

Capsule => T Y Z X Y V V T X Plant => A A A A B B C C C

Parameter % Copper 0.150 0.150 0.150 0.150 0.200 0.200 0.340 0.340 0.340 % Phosphorous 0.019 0.019 0.019 0.019 0.010 0.010 0.021 0.021 0.021 % Nickel 1.020 1.020 1.020 1.020 1.020 1.020 0.660 0.660 0.660 % Manganese 1.180 1.180 1.180 1.180 1.630 1.630 0.980 0.980 0.980 Fluence 2.630E+18 6.920E+18 1.040E+19 8.740E+18 4.550E+18 4.920E+18 5.300E+18 3.870E+19 4.490E+19 EFPY 1.40 3.20 5.50 15.50 2.30 8.60 3.18 7.27 20.39 Withdraw Cycle 1 3 5 12 2 8 3 8 20 Coolant Temperature (F) 543.00 543.00 542.00 540.50 543.00 528.00 547.00 547.00 547.00 Delta T30 96.87 152.52 171.74 175.62 165.22 194.90 167.29 225.76 230.66 Measured Value of Delta T30 151.60 172.00 229.20 193.20 195.00 204.00 209.32 288.15 265.93 Residual "r" 54.73 19.48 57.46 17.58 29.78 9.10 42.03 62.39 35.27

Type A Test Standard Deviation 26.40 Mean Deviation 36.42 Maximum Mean Residual 20.50 Pass/Fail? Fail

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A.2 Sensitivity Evaluation of Mean Deviation Test for Weld Heat with Sister Plant Chemistry Values Adjusted

Weld Metal - W5214

Number of Data Points => 9 Capsule => T Y Z X Y V V T X

Plant => A A A A B B C C C Parameter % Copper 0.200 0.200 0.200 0.200 0.200 0.200 0.200 0.200 0.200 % Phosphorous 0.010 0.010 0.010 0.010 0.010 0.010 0.010 0.010 0.010 % Nickel 1.020 1.020 1.020 1.020 1.020 1.020 1.020 1.020 1.020 % Manganese 1.630 1.630 1.630 1.630 1.630 1.630 1.630 1.630 1.630 Fluence 2.630E+18 6.920E+18 1.040E+19 8.740E+18 4.550E+18 4.920E+18 5.300E+18 3.870E+19 4.490E+19 EFPY 1.40 3.20 5.50 15.50 2.30 8.60 3.18 7.27 20.39 Withdraw Cycle 1 3 5 12 2 8 3 8 20 Coolant Temperature (F) 543.00 543.00 542.00 540.50 543.00 528.00 547.00 547.00 547.00 Delta T30 126.18 190.08 210.74 214.81 165.22 194.90 172.14 253.80 259.42 Measured Value of Delta T30 151.60 172.00 229.20 193.20 195.00 204.00 209.32 288.15 265.93 Residual "r" 25.42 -18.08 18.46 -21.61 29.78 9.10 37.18 34.35 6.51

Type A Test Standard Deviation 26.40 Mean Deviation 13.46 Maximum Mean Residual 20.50 Pass/Fail? Pass

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Appendix B

Example Worksheets for 10 CFR 50.61a Evaluation of Data

B.1 Template for Evaluation of Surveillance Capsule Data

10 CFR 50.61a Surveillance Data Evaluation for Plant Name Base Metal / Weld Wire Heat Number with an Adjustment of °F

Capsule Direction Log of Fluence

Adjusted Residual

"r" (x - xavg)2

Adjusted r*

(r/sigma)

Mean Deviation Test

Slope Deviation Test

Outlier Deviation Test Standard Deviation (sigma)

Slope (m) Largest r*

Mean Deviation Standard Error

of Fit Largest allowable r*

Maximum Mean Residual Standard Error

of Slope Pass/Fail?

Pass/Fail? T-Statistic Second largest r*

Critical T-Statistic Second largest

allowable r*

Pass/Fail? Pass/Fail?

35

B.2 Example 10 CFR 50.61a Flaw Evaluation of Weld ISI Data (Plant specific input in shaded cells only) Weld Axial Flaw Evaluation for ISI Weld Length (WL, inch) of: 1006

Min. TWE Max. TWE 1000" WL Act. WL No. ISI Percent Percent (inch) (inch) Max No. Max No. Ax. Flaws of TWCF Increase 0.076 0.125 75.90 76.36 78 20.91 0.45 0.126 0.175 67.98 68.39 70 20.91 0.49 0.176 0.225 14.16 14.24 16 10.25 1.26 0.226 0.275 4.65 4.68 4 10.25 0.00 0.276 0.325 1.00 1.01 2 9.61 9.50 0.326 0.375 1.52 1.53 1 9.61 0.00 0.376 0.425 0.49 0.00 0 9.27 0.00 0.426 0.475 1.00 1.01 1 10.34 0.00 0.476 0.525 0.00 0.00 0 10.34 0.00 0.526 0.613 0.00 0.00 1 7.94 7.94 0.614 0.700 0.00 0.00 0 5.22 0.00 0.701 0.787 0.00 0.00 0 3.86 0.00 0.788 0.875 0.00 0.00 1 1.90 1.90 0.876 0.963 0.00 0.00 0 4.94 0.00 0.964 1.050 0.00 0.00 2 1.60 3.20 1.051 1.137 0.00 0.00 0 2.78 0.00 1.138 1.225 0.00 0.00 0 1.67 0.00 1.226 1.925 0.00 0.00 0 8.03 0.00

Total = 24.74

36

B.3 Example 10 CFR 50.61a Flaw Evaluation of Plate ISI Data (Plant specific input in shaded cells only) Plate Axial Flaw Evaluation for ISI Plate Area (PA, sq. inch) of: 2012

Min. TWE Max. TWE 1000 PA Act. PA No. ISI Percent Percent (inch) (inch) Max No. Max No. Ax. Flaws of TWCF Increase 0.076 0.125 4.90 9.86 12 7.39 1.61 0.126 0.175 2.30 4.63 7 10.50 5.38 0.176 0.225 0.56 1.13 2 10.50 8.14 0.226 0.275 0.20 0.00 0 10.50 0.00 0.276 0.325 0.07 0.00 0 10.07 0.00 0.326 0.375 0.01 0.00 1 9.27 9.27 0.376 0.402 0.00 0.00 0 9.27 0.00

Total = 24.40

B.4 Example 10 CFR 50.61a TWCF Evaluation for Weld and Plate ISI Data (Plant specific input in shaded cells only)

TWCF for: Basis Plants at 60 EFPY

Type of RT-max RT-max ISI Flaw Alpha 95%TWCF Flaw (deg. F) (deg. R) Factor Factor per Year AW 247.4 707.07 1.2474 2.008 4.045E-07

PL 253.0 712.67 1.2440 1.974 9.219E-09 CW 312.0 771.67 1.0000 1.620 9.558E-09

Total = 4.234E-07 Limit = 1.000E-06 Pass? Yes

37

B.5 Comparison of FAVOR and VFLAW Output Information on Weld Flaw Densities

Palisades Embedded Axial Weld Flaw Density for FAVOR 06.1

FAVOR 70,00 Cat. Avg. No. Cat. Mean VFLAW Mean VFLAW Ratio VFLAW Max. VFLAW Max. VFLAW VFLAW Max. TWE (in) 2 Flaws 2 per Vessel No. / Sq. Foot No. Cat 2. to FAVOR No. / Sq. Foot No. Cat 2. to Mean Ratio

0.088 38465866 549.512 3.6072E+01 549.860 1.00063 4.1899E+01 638.683 1.16154 0.175 7597091 108.530 7.1206E+00 108.542 1.00011 1.4356E+01 218.834 2.01612 0.263 551377 7.877 5.1658E-01 7.874 0.99970 1.5987E+00 24.370 3.09478 0.350 137096 1.959 1.2883E-01 1.964 1.00270 1.9965E-01 3.043 1.54972 0.438 51294 0.733 4.7871E-02 0.730 0.99583 8.7474E-02 1.333 1.82729 0.525 20600 0.294 1.9388E-02 0.296 1.00426 4.1486E-02 0.632 2.13978 0.613 9256 0.132 8.7075E-03 0.133 1.00381 2.1384E-02 0.326 2.45581 0.700 4674 0.067 4.3642E-03 0.067 0.99631 1.1766E-02 0.179 2.69603

Subtotal = 8336858 119.098 119.110 1.00010 247.581 2.07859 Total = 46837254 669.104 669.465 1.00054 887.402 1.32554 All Total = 46844656 669.209

Notes: 1) Subtotal is for TWE from 0.175" to 0.438", Total is from 0.088" to 0.7", All Total is from 0.088" to 1.925".

2) Max. TWE of 0.7 " > 17 mm and gives average of 1 axial weld flaw for ISI of all 48 U.S. PWR Plants.

Weld 70K Region Vessel ISI Volume Type Weld Flaws Length (in.) Length (in.) Axial 1 8260605 60.71 48.71 Axial 2 8267896 60.71 48.71 Axial 3 8264926 60.71 48.71 Axial 4 12789082 93.98 81.98 Axial 5 12792965 93.98 81.98 Axial 6 12788065 93.98 81.98 Circum. 77393598 613.93 Total 140557137 1006.00 All Axial 63163539

Axial Ratio 0.44938 * ST Max/Mean 2.07859 is Net Effect of 0.934075

38

B.6 Comparison of FAVOR and VFLAW Output Information on Plate Flaw Densities

Beaver Valley 1 Embedded Plate Flaw Limits for FAVOR 06.1

FAVOR 74,00 Cat. Avg. No. Cat. Mean VFLAW Mean VFLAW Ratio VFLAW Max. VFLAW Max. VFLAW VFLAW Max. TWE (in) 2 Flaws 2 per Vessel No. / Cubic Ft. No. Cat 2. to FAVOR No. / Cubic Ft. No. Cat 2. to Mean Ratio

0.080 49092325 663.410 1.3996E+01 663.729 1.00048 1.7305E+01 820.651 1.23642 0.161 27747715 374.969 7.9129E+00 375.251 1.00075 1.0514E+01 498.603 1.32872 0.241 4741809 64.079 1.3522E+00 64.125 1.00073 3.5589E+00 168.773 2.63193 0.321 412306 5.572 1.1761E-01 5.577 1.00102 7.3123E-01 34.677 6.21741 0.402 51876 0.701 1.4809E-02 0.702 1.00179 3.0626E-02 1.452 2.06807

Subtotal = 32953706 445.320 445.656 1.00075 703.505 1.57858 Total = 81994155 1108.029 1108.683 1.00059 1522.703 1.37343

Notes: 1) Subtotal is for TWE from 0.161" to 0.402", Total is for TWE from 0.080" to 0.402".

2) Plate fabrication flaws also include fabrication flaws in forgings without undercald cracking.

Dimension Inches Vessel Inner Radius 78.5 Cladding Thickness 0.156 Inner Radius Base Metal R 78.656 Total Wall Thickness t 8.031 Outer Radius Cat. 2 Flaws (R+1/8t) 79.6599 Lower Weld/Plate Height 61.68 Upper Weld/Plate Height 99.45 Total Weld/Plate Height 161.13 Total Plate Volume (360 deg.) 81946.5

Axial Ratio 0.50000 * ST Max/Mean 1.57858 is Net Effect of 0.789291

39

Appendix C

Selected Output from VFLAW and FAVOR 06.1 Programs

C.1 Palisades Maximum (Largest) Embedded Weld Flaw Density from VFLAW

40

C.2 Palisades Mean (Average) Embedded Weld Flaw Density from VFLAW

41

C.3 Beaver Valley 1 Maximum, Median (50%) and Mean Embedded Plate Flaw Densities from VFLAW

42

C.4 FAVPOST Output for Palisades Weld Contribution to TWCF at 60 EFPY

43

C.5 FAVPOST Output for Beaver Valley 1 Plate Contribution to TWCF at 60 EFPY

44

Appendix D

Slope Deviation Test Evaluation of High Fluence Data

D.1 Slope Deviation Test – PLANT A, Surveillance Forging

Surveillance Data

Capsule Fluence

(x1019 n/cm2, E > 1.0MeV)

EFPY Measured ΔT30 (ºF)

Calculated ΔT30 (ºF)

Residual "r"

V 5.87E+18 1.4 0.0 17.98 -17.98 R 1.02E+19 2.6 20.1 23.70 -3.60 T 1.69E+19 6.9 0.0 31.14 -31.14 S 3.64E+19 17.0 76.8 45.15 31.65 N 5.80E+19 30.5 76.4 60.90 15.50

Slope Deviation Test Slope (m) 43.27 Standard Error of Fit 21.02 Standard Error of Slope 26.08 T-Statistic 1.66 Critical T-Statistic 4.54 Pass/Fail? Pass

D.2 Slope Deviation Test – PLANT A, Surveillance Forging

Surveillance Data

Capsule Fluence

(x1019 n/cm2, E > 1.0MeV)



Residual "r"

V 5.87E+18 1.4 34.7 17.99 16.71 R 1.02E+19 2.6 57.5 23.72 33.78 T 1.69E+19 6.9 33.6 31.17 2.43 S 3.64E+19 17.0 45.8 45.19 0.61 N 5.80E+19 30.5 91.1 60.95 30.15

Slope Deviation Test Slope (m) -2.95 Standard Error of Fit 17.60 Standard Error of Slope 21.84 T-Statistic -0.13 Critical T-Statistic 4.54 Pass/Fail? Pass

45

D.3 Slope Deviation Test – PLANT A, Surveillance Weld

Surveillance Data

Capsule Fluence

(x1019 n/cm2, E > 1.0MeV)



Residual "r"

V 5.87E+18 1.4 146.7 139.56 7.14 R 1.02E+19 2.6 156.2 154.33 1.87 T 1.69E+19 6.9 149.7 167.94 -18.24 S 3.64E+19 17.0 212.2 189.62 22.58 N 5.80E+19 30.5 216.9 212.08 4.82


D.4 Slope Deviation Test – PLANT B, Surveillance Plate

Surveillance Data

Capsule Fluence

(x1019 n/cm2, E > 1.0MeV)



Residual "r"

Y (LT) 6.05E+18 1.11 105.5 111.26 -5.76 U (LT) 1.73E+19 3.96 167.7 142.18 25.52 X (LT) 2.98E+19 6.43 164.8 158.17 6.63 W (LT) 4.92E+19 13.85 200.1 175.65 24.45 V (LT) 6.79E+19 19.01 214.2 191.90 22.30 Z (LT) 8.73E+19 21.82 218.3 206.17 12.13 Y (TL) 6.05E+18 1.11 124.0 111.26 12.74 U (TL) 1.73E+19 3.96 168.5 142.18 26.32 X (TL) 2.98E+19 6.43 200.1 158.17 41.93 W (TL) 4.92E+19 13.85 195.8 175.65 20.15 V (TL) 6.79E+19 19.01 231.0 191.90 39.10 Z (TL) 8.73E+19 21.82 215.3 206.17 9.13

Slope Deviation Test Slope (m) 10.84 Standard Error of Fit 13.40 Standard Error of Slope 9.85

T-Statistic 1.10 Critical T-Statistic 2.76 Pass/Fail? Pass

46

D.5 Slope Deviation Test – PLANT B, Surveillance Weld

Surveillance Data

Capsule Fluence

(x1019 n/cm2, E > 1.0MeV)



Residual "r"

U 6.05E+18 1.11 -28.4 23.46 -51.86 W 1.73E+19 3.96 7.0 42.02 -35.02 X 2.98E+19 6.43 -15.6 54.12 -69.72 Z 4.92E+19 13.85 10.2 68.22 -58.02 Y 6.79E+19 19.01 69.1 81.69 -12.59 V 8.73E+19 21.82 56.5 93.59 -37.09


D.6 Slope Deviation Test – PLANT C, Surveillance Weld

Surveillance Data

Plant Capsule Fluence

(x1019 n/cm2, E > 1.0MeV)



Residual "r"

C U 3.78E+18 1.09 161.1 145.66 15.44 C X 1.40E+19 4.30 170.6 197.31 -26.71 C V 1.93E+19 7.24 179.8 205.24 -25.44 C Y 2.64E+19 10.21 190.2 212.46 -22.26 C W 5.10E+19 19.22 208.0 229.74 -21.74

Slope Deviation Test Slope (m) -34.37 Standard Error 12.06 Standard Error of Slope 14.41 T-Statistic -2.39 Critical T-Statistic 4.54 Pass/Fail? Pass

47

D.7 Slope Deviation Test – PLANT C, Surveillance Weld

Surveillance Data

Plant Capsule Fluence

(x1019 n/cm2, E > 1.0MeV)



Residual "r"

D 263º 6.20E+18 2.98 59.0 92.25 -33.25 D 97º 2.64E+19 11.07 93.0 121.06 -28.06 C Y 6.12E+18 1.15 66.9 65.09 1.81 C U 1.73E+19 3.08 75.1 89.18 -14.08 C X 3.06E+19 6.11 87.4 104.39 -16.99 C W 4.75E+19 12.43 98.3 118.19 -19.89 C V 7.14E+19 20.16 117.5 135.38 -17.88 C Z 8.47E+19 24.26 113.5 144.36 -30.86

Slope Deviation Test Slope (m) -8.16 Standard Error of Fit 11.50 Standard Error of Slope 9.93 T-Statistic -0.82 Critical T-Statistic 3.14 Pass/Fail? Pass

48

Appendix E

Calculation of TWCF per Weld Flaw E.1 TWCF per Weld Flaw – Palisades, 70,000 Vessel Simulations, 60 EFPY

Flaw Depth (TWE)

# of Flaws

Simulated

# of Flaws Simulated per Vessel

% Contribution

to CPF

Fractional % Contribution

Per Flaw

TWCF Per Flaw Based

on Total TWCF of

7.0253 x 10-7 Events per

year 0.088 38465866 549.51 1.68 3.06E-05 2.15E-11 0.175 7597091 108.53 20.91 1.93E-03 1.35E-09 0.263 551377 7.88 10.25 1.30E-02 9.14E-09 0.35 137096 1.96 9.61 4.91E-02 3.45E-08

0.438 51294 0.73 9.27 1.27E-01 8.89E-08 0.525 20600 0.29 10.34 3.51E-01 2.47E-07 0.613 9256 0.13 7.94 6.00E-01 4.22E-07

0.7 4674 0.07 5.22 7.82E-01 5.49E-07 0.787 2632 0.04 3.86 1.03E+00 7.21E-07 0.875 1638 0.02 1.9 8.12E-01 5.70E-07


of 101

Attachment 6

NRC PRESENTATION

1

Low Upper Shelf Energy Methodology Update

Mark Kirk Michael Benson

U.S. NRC Office of Nuclear Regulatory Research

2 2

• Low USE is covered by two documents

– Regulatory Guide (RG) 1.161

– Previously Code Case N-xxxx

– ASME Section XI, Nonmandatory Appendix K

– RG 1.161 was issued before Appendix K

Low USE Methods

3 3

• Comments on applied J equations:

– Basis documents:

– Newman and Raju, J. Pressure Vessel Tech, 102:342-346, 1980.

– Welding Research Council Bulletin 413

– The J equations were developed only for PWR vessel geometries

– Confirmatory work shows that they remain valid for BWR geometries

RG 1.161 vs. Appendix K Similarities

4 4

• Acceptance criteria:

– Initiation: J = JR at ∆a = 0.1 in, SF = 1.15

– Stability: at J = JR, SF = 1.25

– Note that Anderson’s text shows

RG 1.161 vs. Appendix K Similarities

aJ

aJ R

∂∂

<∂∂

aJ

aJ R

∂∂

≤∂∂

5 5

• Material resistance properties:

– RG 1.161 includes JR correlations developed by Eason in NUREG/CR-5729, 1991

– Appendix K does not contain JR correlations

– States that the JR curve “shall be a conservative representation of the toughness of the controlling beltline material at upper shelf temperatures in the operating range”

• Evaluation of the crack stability criterion:

– RG 1.161 endorses one method: crack driving force diagram

– Appendix K endorses three methods: crack driving force diagram, failure assessment diagram, J/Tearing Modulus procedure

RG 1.161 vs. Appendix K Differences

6 6

Identified Issues Temperature Input into Eason Correlation

• Temperature input will have a large effect on the required USE value, so crack-tip temperature must be appropriately justified

• Model: RPV welds, Jd, Charpy, Fixed C4

• CVN = 33 ft-lb

• Bn = 1 in.

7

JR Model

Identified Issues

• Both the RG and Appendix K need additional guidance on model selection:

– The RG gives equations for the CVN model and says that other models may be used

– Cu-φt model should only be used for Linde 80 welds when Charpy data are not available

– The CVNp model should not be used to calculate a required USE number

– Jd models should be used instead of Jm models

– In general, the NRC feels that there should be more prescriptive guidance on the choice of model

8

Calculation of Required Upper Shelf Energy (USE)

Identified Issues

• Both the RG and Appendix K need an additional section: Calculation of Required USE

– Step 1: Calculate the minimum value of CVN that satisfies criterion 1, J = JR at ∆a = 0.1

– Step 2: Via an iterative method (next slide), calculate the minimum value of CVN that satisfies criterion 2, crack stability

– Compare the two, the largest one is the required USE

– Include an example problem

9

Guidance on Crack Stability Criterion

Identified Issues

Is J > JR for all ∆a?

CVNi+1 = CVNi + n ft-lbs CVN = CVNi

Yes No

i = i + 1

END

CVNi = 35 ft-lbs

Calculate JR

i = 0

Calculate J

CVNi-1

CVNi

CVNi-1

10

Update of Eason Correlations

Identified Issues

• Eason JR correlations were developed in 1991

– The correlations are based upon limited data beyond ~1e19 n/cm2, and therefore should not be extrapolated beyond this value

• According to scaled RVID2 data, there are 14 plants whose vessels have locations projected to exceed this fluence level at ~60 years

• NRC is planning to

– Conduct a comprehensive literature survey to identify data useful for JR curve correlation models

– Develop models based on these data

11

NRC Request

• NRC is considering two options:

– Revise RG 1.161; or

– Revise Appendix K; then retire RG1.161

• Would like input:

– Which approach is preferred?

– Are the proposed changes appropriate?

– Are there any other changes or updates needed beyond those identified here?

• If the preferred approach involves revision of Appendix K

– Are there parties interested in working on this effort?

– How is the effort best coordinated within ASME?

nrc public meeting summary reportnrc public meeting summary report 11/09/11 – discussion of rpv...

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