RA-19-0138, Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1) for Reactor Vessel Cold Leg Dissimilar Metal Weld Inspections

Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1) for Reactor Vessel Cold Leg Dissimilar Metal Weld Inspections

ML19204A082
Person / Time
Site:	Robinson
Issue date:	07/23/2019
From:	Ellis K Duke Energy Progress
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML19205A203	List: RA-19-0138, Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1) for Reactor Vessel Cold Leg Dissimilar Metal Weld Inspections
References
RA-19-0138
Download: ML19204A082 (42)

Text

Kevin Ellis Manager - Nuclear Support Services

( ., DUKE H. B. Robinson Steam Electric Plant, Unit 2 ENERGY Duke Energy 3581 West Entrance Road Hartsville, SC 28550 843.951.1329 Kevin.Ellis@duke-energy.com PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 2 THIS LETTER IS UNCONTROLLED Serial: RA-19-0138 10 CFR 50.55(a)

July 23, 2019 United States Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 H.B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261 / RENEWED LICENSE NO. DPR-23

SUBJECT:

Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1) for Reactor Vessel Cold Leg Dissimilar Metal Weld Inspections Ladies and Gentlemen:

Pursuant to 10 CFR 50.55a(z)(1), Duke Energy Progress, LLC (hereafter referred to as Duke Energy) requests the NRC to grant relief from Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) and the augmented inspections of ASME Code Case N-770-2 as prescribed by 10CFR50.55a(g)(6)(ii)(F). Relief is being sought on the basis that the proposed alternative provides an acceptable level of quality and safety.

To support the current inspections scheduled during the Fall 2020 refueling outage, Duke Energy is requesting review and approval of this relief request by August 1, 2020.

Attachment 1 provides Relief Request RA-19-0138. Attachment 2 provides Westinghouse calculation LTR-SDA-18-016-P, Revision 0 which supports the Relief Request in Attachment 1.

Attachment 2 contains information that is proprietary to Westinghouse. In accordance with 10 CFR 2.390, Duke Energy requests that Attachment 2 be withheld from public disclosure. An affidavit is included (Attachment 4) attesting to the proprietary nature of Attachment 2. A non-proprietary version of Attachment 2 is included in Attachment 3.

This document contains no new Regulatory Commitments.

PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 2 THIS LETTER IS UNCONTROLLED

PROPRIETARY INFORMATION - WITHHOLD UNDER 10 CFR 2.390 UPON REMOVAL OF ATTACHMENT 2 THIS LETTER IS UN CONTROLLED U.S. Nuclear Regulatory Commission Page2 Serial: RA-19-0138 Should you have any questions concerning this letter, or require additional information, please contact Art Zaremba, Director - Nuclear Fleet Licensing, at 980-373-2062.

Kevin Ellis - Manager - Nuclear Support Services, H.B. Robinson Steam Electric Plant Attachments:

1. Relief Request RA-19-0138, "Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1) for Reactor Vessel Cold Leg Dissimilar Metal Weld Inspections.

2. One copy of Westinghouse LTR-SDA-18-016-P, Revision 0, "Technical Justification to Support the Extended Volumetric Examination Interval for H. B. Robinson Unit 2 Reactor Vessel Inlet Nozzle to Safe End Dissimilar Metal Welds." (Proprietary)

3. One copy of Westinghouse LTR-SDA-18-016-NP, Revision 0, "Technical Justification to Support the Extended Volumetric Examination Interval for H. B. Robinson Unit 2 Reactor Vessel Inlet Nozzle to Safe End Dissimilar Metal Welds." (Non-Proprietary)

4. Westinghouse Application for Withholding Proprietary Information from Public Disclosure CAW-18-4776, accompanying Affidavit, Proprietary Information Notice, and Copyright Notice.

cc: L. Dudes, NRC Regional Administrator, NRC, Region II M. Fannon, NRC Senior Resident Inspector, HBRSEP, Unit No. 2 N. Jordan, NRC Project Manager, NRR

U.S. Nuclear Regulatory Commission Attachment 1 Serial: RA-19-0138 Attachment 1 Duke Energy Progress, LLC H.B. Robinson, Unit 2 Relief Request RA-19-0138 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1) for Reactor Vessel Cold Leg Dissimilar Metal Weld Inspections

H.B. Robinson, Unit 2 Relief Request RA-19-0138 Attachment 1 Page 1 of 7 1.0 ASME Code Component(s} Affected The affected components are H.B. Robinson (RNP), Unit 2 Reactor Vessel Cold Leg Nozzle to Safe-End welds. These welds are Alloy 82/182 type welds covered by Code I '

Case N-770-2, Inspection Item B.

N-770-2 COMPONENT INSPECTION NOZZLE SIZE DESCRIPTION ID ITEM B 107/14DM 27.5" ID (Norn.) Loop "A" Cold Leg Nozzle to Safe End OM Weld B 107A/14DM 27.5" ID (Norn.) Loop "B" Cold Leg Nozzle to Safe End OM Weld B 107B/14DM 27.5" ID (Norn.) Loop "C" Cold Leg Nozzle to Safe End OM Weld Component materials and nozzle weld configurations are shown in Figure 1 below. The cold leg nozzle operating temperature is 547.6 °F with a thickness of 2.5 inches.

~ 2~_ R. (apprx.)

I\

.--~r--~-=--~---,~~

AJJoyl'1 I

I AuJt~Jrfc I Austt:nitk Stolrilt:ss Slttl / Stainles.s

{Safe End} J Stt:rl f crod fJ ------ -------

MACHJIVE THIS AREA TO 11*1/ir *Z/16*/-C" DIA.

PRIOR TOX-IIAY Of wno. - - -------- - - -- - - ~

Figure 1 - Cold Leg Nozzle to Safe-End Weld (Not to Scale}

H.B. Robinson, Unit 2 Relief Request RA-19-0138 Attachment 1 Page 2 of 7

2.0 Applicable Code Edition and Addenda

The Applicable Code edition in the 5th Ten-Year Inservice Inspection Interval for RNP, Unit 2 is the 2007 Edition of the ASME Section XI Code with 2008 Addenda.

3.0 Applicable Code Requirement

10CFR50.55a(g)(6)(ii)(F) requires holders of operating licenses or combined licenses for pressurized-water reactors as of or after August 17, 2017, to implement the requirements of ASME BPV Code Case N-770-2. Inspection Item B of Code Case N-770-2 requires unmitigated butt welds at Cold Leg operating temperatures ( 525°F and < 580°F) to be volumetrically examined every second inspection period not to exceed 7 years.

4.0 Reason for Request

The next required examination for the RNP Unit 2 Reactor Vessel Cold Leg nozzle to safe-end welds will be during Refueling Outage 32 (Fall 2020). This extension request is to allow this inspection to coincide with ASME Code Section XI Inservice Inspection (ISI) of the reactor vessel and the Materials Reliability Program (MRP)-227 reactor vessel internals inspection associated with license renewal commitments.

The Reactor Vessel Cold Leg welds (N-770-2, Inspection Item B) for RNP, Unit 2 are inspected from the Inner Diameter (ID) of the pipe. Therefore, to access the Cold Leg welds for examination, a critical Core Barrel lift must be performed. By receiving relief to move the Cold Leg inspections, RNP Unit 2 can leverage performing the inspection with the aforementioned required ASME Section XI ISI of the reactor vessel and MRP-227 inspections which also require this critical Core Barrel lift. Additionally, personnel dose savings for the site can be realized with this extension, promoting As Low As Reasonably Achievable (ALARA) practices associated with this significant and infrequently performed evolution.

EPRI document MRP- 349 (Reference 2) and a site-specific crack growth evaluation for RNP, Unit 2 (Reference 3, Attachment 2) provide the overall basis for extension of the current volumetric inspection interval for Reactor Vessel Cold Leg Dissimilar Metal (DM) welds. This technical basis demonstrates that the re-examination interval can be extended to the requested interval length while maintaining an acceptable level of quality and safety.

5.0 Proposed Alternative and Basis for Use 10CFR50.55.a(z) states, in part:

Alternatives to codes and standards requirements. Alternatives to the requirements of paragraphs (b) through (h) of this section or portions thereof may be used when authorized by the Director, Office of Nuclear Reactor Regulation, or Director, Office of New Reactors, as appropriate. A proposed alternative must be submitted and authorized prior to implementation. The applicant or licensee must demonstrate that:

(1) Acceptable level of quality and safety. The proposed alternative would provide an acceptable level of quality and safety;

H.B. Robinson, Unit 2 Relief Request RA-19-0138 Attachment 1 Page 3 of 7 Duke Energy believes that the proposed alternative of this request provides an acceptable level of quality and safety.

RNP Unit 2 proposes a one-time extension to the Code Case N-770-2, Table 1, Inspection Item B, volumetric examination from every 2nd inspection period not to exceed 7 years.

The extension requested is for a period not to exceed one ASME XI ISI interval. The inspection will be performed no later than the Fall 2022 refueling outage, approximately 9 years from the previous inspection (RO-28 in September 2013).

Technical Basis The overall basis used to demonstrate the acceptability of extending the inspection interval for Code Case N-770-2, Inspection Item B components is contained in MRP-349 and the site-specific flaw evaluation performed for RNP Unit 2. In summary, the basis for extending the inspection is: (1) there have been no incidents of cracking in large diameter butt welds operating at Cold Leg temperatures (< 575°F) that can be attributed to PWSCC (2) crack growth rates in Reactor Vessel Cold Leg DM welds are slow, (3) the likelihood of cracking or through-wall leaks, in large-diameter Cold Leg DM welds is very small, (4) the RNP specific axial and circumferential flaw evaluation showing any indication detected during the Refueling Outage 28 (RO-28) inspection (Fall 2013), as well as any flaw size which could have been reasonably missed during the RO-28 (Fall 2013) inlet nozzle weld examination would not grow to the allowable size flaw specified by ASME Section XI rules over the timeframe of the requested inspection interval. This technical basis demonstrates that the re-examination interval can be extended while maintaining an acceptable level of quality and safety.

Service Experience During Refueling Outage RO-25 in October 2008, RNP Unit 2 performed PDI (Performance Demonstrated Initiative) qualified ultrasonic examinations from the inner diameter (ID) surface with phased array technology on all the outlet and inlet (Hot Leg and Cold Leg) Reactor Vessel nozzles to satisfy MRP-139 examination requirements. RNP Unit 2 is a three-loop plant with three inlet and three outlet nozzles. All nozzles met examination volume requirements of essentially 100%. During the October 2008 UT examinations, two (2) adjacent axial flaws were discovered in the B Cold Leg inlet nozzle (107A/14DM): one that started in the stainless steel cladding and extended into the weld butter region and the second that appeared to be clearly embedded in the DM weld itself.

Based on ASME Section XI IWA-3300 proximity rules, the two subsurface indications were conservatively combined during the RO-25 outage, and were evaluated as one embedded flaw with length of 1.5 and total flaw depth of 1.254. The UT examination observed a separation of the embedded flaw from the ID surface, and follow-up eddy current examinations of the indication were performed as a supplement to the UT results. Based on the supplemental eddy current examination at the DM weld area of the combined embedded flaw, it was determined that there was no surface connection of the embedded flaw to the ID surface. A flaw evaluation, as required by ASME B&PV Code,Section XI, IWB-3600, was performed for continued service.

Re-examination of the reactor vessel inlet nozzle DM weld on the B Cold Leg was performed again during RO-27 (February 2012), as well as RO-28 (Fall 2013). The

H.B. Robinson, Unit 2 Relief Request RA-19-0138 Attachment 1 Page 4 of 7 comparison of the 2008, 2012, and 2013 data concluded that there was no growth in the existing flaws, no new flaws present, nor had any of the flaws become surface connected.

It should be noted that there were slight differences in the measurements, but they were well within inherent measurement error, when comparing two separate automated examinations. Based on examinations performed in 2008, 2012 and 2013, it has been demonstrated that the embedded flaws have not grown, and these indications are most likely fabrication flaws and not service induced flaws. Additional discussion on this embedded flaw is in section 7.0 of Attachment 2.

Crack Growth Rates (Flaw Tolerance)

Site specific analysis has been performed for the reactor vessel nozzles using residual stress distributions that were developed and based on the most recent guidance of MRP-287, entitled Primary Water Stress Corrosion Cracking (PWSCC) Flaw Evaluation Guidance. This analysis determines the maximum allowable initial flaw size (based on 9 Effective Full-Power Years (EFPY)) that could be left in service and remain acceptable in accordance with the ASME Section XI IWB-3640 acceptance criteria through the Fall 2022 refueling outage for RNP Unit 2 taking into account potential PWSCC growth since the last volumetric and surface examinations. For development of the residual stress profile, the sequence of 1) initial welding of the nozzle to safe-end joint, 2) an inside surface weld repair with a repair depth of 50% through the dissimilar metal weld thickness and 3) the adjacent stainless steel weld between the safe-end and stainless steel piping was included. The subsequent crack growth analyses have shown that the flaw tolerance of the location is high and a postulated axial or circumferential flaw will not reach the maximum ASME allowable depth within the requested extension time period.

The crack growth curves and maximum allowable inner diameter (ID) surface connected flaw sizes developed for RNP Unit 2 are shown in Figures 7-1 and 7-2 of Attachment 2.

The maximum allowable end-of-evaluation period flaw sizes are also shown on these figures for the axial and circumferential flaw configurations analyzed. Based on these crack growth results, the maximum allowable ID surface connected initial flaw sizes for the axial and circumferential flaws are tabulated in the table below.

Maximum Allowable Initial Flaw Size Axial Flaw Circumferential Flaw (Aspect Ratio = (Aspect Ratio = 10) 2)

Maximum Allowable 0.114 0.394 Initial Flaw Size (a/t)

Flaw Depth (in) 0.285 0.985 Flaw Length (in) 0.568 9.85 The maximum allowable ID surface connected flaw sizes shown in the table above are the largest axial and circumferential flaw sizes that could be left in service and remain acceptable for a service life of 9 years from RO-28 (Fall 2013) to RO-33 (Fall 2022). In accordance with the Ultrasonic Testing (UT) detection and sizing requirements in ASME

H.B. Robinson, Unit 2 Relief Request RA-19-0138 Attachment 1 Page 5 of 7 Section XI Appendix VIII, Supplement 10 (ASME Section XI Code 2007 Edition with the 2008 Addenda), the minimum required detectable flaw depth is 10% of the wall thickness.

Therefore, the maximum allowable initial axial and circumferential flaw sizes are above the minimum flaw depth requirement per the UT detection capabilities, and thus would have been reasonably detected at the previous inspection of the DM welds.

Based on the H. B. Robinson Unit 2 maximum allowable flaw sizes in the table above, the calculated maximum allowable initial axial flaw size (flaw depth = 0.285 and flaw length =

0.568) for 9 EFPY is large enough to have been detected during RO-28 (Fall 2013) examination of the RV inlet nozzle DM welds. Therefore, deferring the volumetric examination for the H. B. Robinson Unit 2 RV inlet nozzle DM welds from the RO-32 (Fall 2020) to the RO-33 (Fall 2022) is technically justified.

Probability of Cracking or Through Wall Leaks Sub-surface indications were also discovered on the RNP Unit 2 Hot Legs during RO-25 in October 2008 that were evaluated for continued service. Nine (9) separate sub-surface indications were discovered in three Hot Leg DM welds (one (1) in the A Hot Leg DM weld, four (4) in the B Hot Leg DM weld, and four (4) in the C Hot Leg DM weld). Re-examinations of the Hot Legs during RO-27 (February 2012), RO-29 (Fall 2015), and RO-31 (Fall 2018) all concluded, with slight differences all well within inherent measurement error, that there was no growth in the existing flaws, no new flaws present, nor had any of the flaws become surface connected. Results of the eddy current examinations performed during all of the refueling outages did not detect any planar surface connected flaws in any of the welds.

Analyses have been performed to calculate the probability of failure for Alloy 82/182 welds using both probabilistic fracture mechanics and statistical methods. Both approaches have shown that the likelihood of cracking or through-wall leaks in large-diameter cold leg welds is very small. Furthermore, sensitivity studies performed using probabilistic fracture mechanics have shown that even for the more limiting high temperature locations, more frequent inspections than required by Section XI, such as that in MRP-139 or Code Case N-770, have only a small benefit in terms of risk.

Though past service experience may not be an absolute indicator of the likelihood of future cracking, the experience does give an indication of the relative likelihood of cracking in cold leg temperature locations versus hot leg temperature locations. While there is some amount of Primary Water Stress Corrosion Cracking (PWSCC) service experience in hot leg locations, the number of indications in large-bore butt welds is still small relative to the number of potential locations. Therefore, if hot leg PWSCC is a leading indicator for cold leg PWSCC, and the higher frequency of inspections will be maintained for the hot leg locations, it is reasonable to conclude that a moderately less rigorous inspection schedule would be capable of detecting any cold leg indications before they became large enough to be a concern.

Conclusions Extending the required RNP Unit 2 Cold Leg DM weld volumetric examination from the Fall 2020 (RO-32) refueling outage to the Fall 2022 refueling outage (RO-33) is justified

H.B. Robinson, Unit 2 Relief Request RA-19-0138 Attachment 1 Page 6 of 7 given (1) there has been no service experience with cracking found in any Reactor Vessel Cold Leg DM welds, (2) the B Cold Leg DM weld embedded flaw has not changed in size over the last three inspections (October 2008, February 2012, and October 2013), (3) all of the Hot Leg DM weld indications have not changed in sizes for 10 years (4) the RNP Unit 2 specific axial and circumferential flaw evaluation showing any undetected flaw size that could have been reasonably missed during the Fall 2013 (RO-28) examination of the B Cold Leg DM weld would not grow to the allowable flaw size specified by ASME XI rules over the timeframe of the requested inspection interval. The use of this proposed alternative will provide an acceptable level of quality and safety. For these reasons, it is requested that the NRC authorize this proposed alternative in accordance with 10 CFR 50.55a(z)(1).

6.0 Duration of Proposed Alternative This request is applicable to RNP Unit 2 Inservice Inspection Program for the 5th Ten-year Interval. The proposed alternative is applicable until the Fall 2022 refueling outage (RO-33).

7.0 Related Industry Relief Requests Similar proposed alternatives were previously approved by the NRC for the following licensees:

1. Comanche Peak Unit 1 Reactor Vessel Cold Leg DM Welds - ML16074A001 (NRC Safety Evaluation). NRC provided relief to extend examination interval to 9 calendar years.

2. McGuire Unit 1 Reactor Vessel Cold Leg DM Welds - ML15232A543 (NRC Safety Evaluation). NRC provided relief to extend examination interval to 10.5 calendar years.

3. Point Beach Unit 2 Steam Generator Nozzle to Safe End DM Welds - ML16063A058 (NRC Safety Evaluation). NRC provided relief to extend examination interval to 7.5 calendar years.

4. Indian Point Unit 2 Reactor Vessel Cold Leg DM Welds - ML13310A575 (NRC Safety Evaluation). NRC provided relief to extend examination interval to 10 calendar years.

5. Indian Point Unit 3 Reactor Vessel Cold Leg DM welds - ML14199A444 (NRC Safety Evaluation). NRC provided relief to extend examination interval to 10 calendar years.

6. South Texas Unit 1 Reactor Vessel Cold Leg DM welds - ML15218A367 (NRC Safety Evaluation). NRC provided relief to extend examination interval to 7.5 calendar years.

7. South Texas Unit 2 Reactor Vessel Cold Leg DM welds - ML16174A091 (NRC Safety Evaluation). NRC provided relief to extend examination interval to 9.5 calendar years.

8. Farley Units 1 and 2 Reactor Vessel Cold Leg DM welds - ML16174A091 (NRC Safety Evaluation). NRC provided relief to extend examination interval to 9.5 calendar years.

H.B. Robinson, Unit 2 Relief Request RA-19-0138 Attachment 1 Page 7 of 7 8.0 References

1. ASME Code Case N-770-2,Section XI Division 1, Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated with UNS N06082 or UNS W86182 Weld Filler Material With or Without Application of Listed Mitigation Activities, Approval Date June 9, 2011.

2. Materials Reliability Program (MRP): PWR Reactor Coolant System Coolant System Cold-Loop Dissimilar Metal Butt Weld Reexamination Interval Extension (MRP-349),

August 2012.

3. Westinghouse LTR-SDA-18-016-P, Revision 0, Technical Justification to Support the Extended Volumetric Examination Interval for H. B. Robinson Unit 2 Reactor Vessel Inlet Nozzle to Safe End Dissimilar Metal Welds. (Proprietary).

4. Materials Reliability Program: Primary Water Stress Corrosion Cracking (PWSCC)

Flaw Evaluation Guidance (MRP-287). EPRI, Palo Alto, CA: 2010. 1021023.

RA-19-0138 Attachment 2 Attachment 2 Duke Energy Progress, LLC H.B. Robinson, Unit 2 Relief Request RA-19-0138 One copy of Westinghouse LTR-SDA-18-016-P, Revision 0, Technical Justification to Support the Extended Volumetric Examination Interval for H. B. Robinson Unit 2 Reactor Vessel Inlet Nozzle to Safe End Dissimilar Metal Welds. (Proprietary)

Note: Text and figures that are within brackets is information proprietary to Westinghouse

RA-19-0138 Attachment 3 Attachment 3 Duke Energy Progress, LLC H.B. Robinson, Unit 2 Relief Request RA-19-0138 One copy of Westinghouse LTR-SDA-18-016-NP, Revision 0, Technical Justification to Support the Extended Volumetric Examination Interval for H. B. Robinson Unit 2 Reactor Vessel Inlet Nozzle to Safe End Dissimilar Metal Welds. (Non-Proprietary)

RA-19-0138 Attachment 3 Westinghouse Non-Proprietary Class 3 LTR-SDA-18-016-NP Revision 0 Technical Justification to Support the Extended Volumetric Examination Interval for H.B. Robinson Unit 2 Reactor Vessel Inlet Nozzle to Safe End Dissimilar Metal Welds July 2018 Author(s): Tyler F. Gaydosik*, Structural Design Analysis III Alexandria M. Carolan*, Structural Design Analysis III Verifier: Anees Udyawar*, Structural Design Analysis Ill Manager: Lynn A. Patterson*, Structural Design Analysis III

• Electronically approved records are authenticated in the electronic document management system.

© 2018 Westinghouse Electric Company LLC All Rights Reserved

@ Westinghouse

• • • This record was final approved on 7/19/2018 6:03:38 PM. (This statement was added by the PRIME system upon its validation)

RA-19-0138 Attachment 3 Westinghouse Non-Proprietary Class 3 Page 2 of 21 LTR-SDA-18-016-NP Revision 0 July 2018 FOREWORD This document contains Westinghouse Electric Company LLC proprietary information and data which has been identified by brackets. Coding (a,c,e) associated with the brackets sets forth the basis on which the information is considered proprietary.

The proprietary information and data contained in this report were obtained at considerable Westinghouse expense and its release could seriously affect our competitive position. This information is to be withheld from public disclosure in accordance with the Rules of Practice 10CFR2.390 and the information presented herein is to be safeguarded in accordance with 10CFR2.390. Withholding of this information does not adversely affect the public interest.

This information has been provided for your internal use only and should not be released to persons or organizations outside the Directorate of Regulation and the ACRS without the express written approval of Westinghouse Electric Company LLC. Should it become necessary to release this information to such persons as part of the review procedure, please contact Westinghouse Electric Company LLC, which will make the necessary arrangements required to protect the Companys proprietary interests.

The proprietary information in the brackets has been deleted in this report. The deleted information is provided in the proprietary version of this report (LTR-SDA-18-016-P Revision 0).

• • • This record was final approved on 7/19/2018 6:03:38 PM. (This statement was added by the PRIME system upon its validation)

RA-19-0138 Attachment 3 Westinghouse Non-Proprietary Class 3 Page 3 of 21 LTR-SDA-18-016-NP Revision 0 July 2018 1.0 Introduction The potential for Primary Water Stress Corrosion Cracking (PWSCC) for reactor vessel nozzle Dissimilar Metal (DM) weld requires an appropriate assessment of the examination frequency as well as the overall examination strategy for nickel-base alloy components and weldments. ASME Code Case N-770-2 (Reference 1) provides the visual and volumetric inspection guidelines for the primary system piping DM butt welds to augment the current inspection requirements.

In accordance with ASME Code Case N-770-2 guidelines, volumetric examinations are required for the unmitigated DM butt welds at the Reactor Vessel (RV) inlet nozzles every second inspection period not exceeding 7 years. A volumetric examination was previously performed for the H. B. Robinson Unit 2 RV inlet nozzle to safe end DM butt welds during the Fall 2013 Refueling Outage (RFO). The next volumetric examination for the RV inlet nozzle DM welds was planned during the Fall 2020 RFO.

The fracture mechanics evaluation in this report will determine the impact of performing the volumetric examination on H. B. Robinson Unit 2 during the Fall 2022 RFO. The time interval between the previous examination during the Fall 2013 RFO and the planned examination during the Fall 2022 RFO is 9 years, rather than the 7 years allowed by Code Case N-770-2. Therefore, H. B. Robinson is seeking relaxation from the ASME Code Case N-770-2 examination requirement to be able to defer the volumetric examination from the Fall 2020 RFO to the Fall 2022 RFO.

The technical justification to support this request is developed in this report based on a flaw tolerance analysis. The objective of the flaw tolerance analysis is to determine the largest initial axial and circumferential flaw sizes that could be left behind in service and remain acceptable until the next planned inspection. This maximum allowable initial flaw size can then be compared to a flaw size which would have been detected during the Fall 2013 RFO inlet nozzle DM weld examination based on the inspection detection capability.

The following sections provide a discussion of the methodology, geometry, loading and the flaw tolerance analyses performed to develop the technical justification for deviating from the volumetric examination requirements of ASME Code Case N-770-2.

• • • This record was final approved on 7/19/2018 6:03:38 PM. (This statement was added by the PRIME system upon its validation)

RA-19-0138 Attachment 3 Westinghouse Non-Proprietary Class 3 Page 4 of 21 LTR-SDA-18-016-NP Revision 0 July 2018 2.0 Methodology In order to support the technical justification for deferring the volumetric examination from the Fall 2020 RFO to the Fall 2022 RFO for H. B. Robinson Unit 2, it is necessary to demonstrate the structural integrity of the RV inlet nozzle DM welds subjected to the PWSCC growth mechanism. To demonstrate the structural integrity of the DM welds, it is essential to determine the maximum allowable initial flaw size that would be acceptable in the DM welds for the duration between examinations. This maximum allowable initial flaw size would be the largest flaw size that would remain acceptable until the Fall 2022 RFO. The maximum allowable initial flaw size for a given plant operation duration can be determined by subtracting the PWSCC growth for that plant operation duration from the maximum allowable end-of-evaluation period flaw size, which is determined in accordance with ASME Code Section XI (Reference 2).

To determine the maximum allowable end-of-evaluation period flaw sizes and the crack tip stress intensity factors used for the PWSCC analysis, it is necessary to establish the stresses, crack geometry and the material properties at the locations of interest. The applicable loadings which must be considered consist of piping reaction loads acting at the DM weld regions and the welding residual stresses which exist in the region of interest.

The piping loads at the RV inlet nozzle DM weld locations are used to determine PWSCC growth. In addition to the piping loads, the effects of welding residual stresses are also considered. For PWSCC, the crack growth model for the DM weld material is based on that given in MRP-115 for Alloy 182 weld material (Reference 3); this PWSCC growth model is also documented in ASME Section XI (Reference 2). Note that per MRP-115 (Reference 3), the Alloy 182 PWSCC crack growth rate bounds that of the Alloy 82 material. The nozzle geometry and piping loads used in the fracture mechanics analysis are shown in Section 3.0. A discussion of the welding residual stress distributions used for the DM welds is provided in Section 4.0. The determination of the maximum allowable end-of-evaluation period flaw sizes is discussed in Section 5.0.

The maximum allowable initial flaw size will be determined based on the crack growth due to the PWSCC growth mechanism at the RV inlet nozzle DM weld. The PWSCC growth is calculated based on the normal operating temperature and the crack tip stress intensity factors resulting from the normal operating steady state piping loads and welding residual stresses as discussed in Section 6.0. Section 7.0 provides the crack growth curves used in developing the technical justification to deviate from the ASME Code Case N-770-2 guidelines by deferring the volumetric inspection of the RV inlet nozzle DM welds from the Fall 2020 to Fall 2022 RFO.

• • • This record was final approved on 7/19/2018 6:03:38 PM. (This statement was added by the PRIME system upon its validation)

RA-19-0138 Attachment 3 Westinghouse Non-Proprietary Class 3 Page 5 of 21 LTR-SDA-18-016-NP Revision 0 July 2018 3.0 Nozzle Geometry and Loads The DM weld geometry for the H. B. Robinson Unit 2 RV inlet nozzles is based on the nozzle detail drawings (Reference 4). The operating temperature of the reactor vessel inlet nozzles is based on customer correspondence based on plant operation data. The RV inlet nozzle geometry and normal operating temperature used in the analysis are summarized in Table 3-1.

The piping reaction loads at the RV inlet nozzle DM weld locations are summarized in Table 3-2. These loads are used in determining the maximum allowable end-of-evaluation period flaw sizes and the PWSCC growth.

Table 3-1 H. B. Robinson Unit 2 Reactor Vessel Inlet Nozzle Geometry and Normal Operating Temperature Dimension Outside Diameter (in) 32.5 Inside Diameter (in) 27.5 Thickness (in) 2.5 RV Inlet Nozzle Normal Operating Temperature = 547.6°F Table 3-2 H. B. Robinson Unit 2 Reactor Vessel Inlet Nozzle Piping Loads a,c,e

• • • This record was final approved on 7/19/2018 6:03:38 PM. (This statement was added by the PRIME system upon its validation)

RA-19-0138 Attachment 3 Westinghouse Non-Proprietary Class 3 Page 6 of 21 LTR-SDA-18-016-NP Revision 0 July 2018 4.0 Dissimilar Metal Weld Residual Stress Distribution The plant specific welding residual stresses used in the PWSCC growth analysis are determined from the finite element stress analysis (FEA) in (Reference 5) based on the H. B. Robinson Unit 2 RV inlet nozzle DM weld specific configuration. Figure 4-1 shows a sketch of the H. B. Robinson Unit 2 inlet nozzle DM weld configuration. The FEA in (Reference 5) is based on a two-dimensional axisymmetric model of the inlet nozzle DM weld region. The FEA model geometry includes a portion of the low alloy steel nozzle, the stainless steel safe end, a portion of the stainless steel piping, the DM weld attaching the nozzle to the safe end, and the stainless steel weld attaching the safe end to the piping. The FEA model also assumes a 360° inside surface weld repair with a repair depth of 50% through the DM weld thickness, which is consistent with MRP-287 guidance (Reference 6). The following fabrication sequence was simulated in the FEA and matches the information provided in the reactor vessel nozzle details drawings (Reference 4):

The inlet nozzle was welded to the safe end ring forging using an Alloy 82/182 weld. The inner and outer diameter of the dissimilar metal weld is machined to finished size.

An assumed 50% inside surface weld repair 360° around the circumference was conservatively simulated in the Alloy 82/182 weld, which is consistent with MRP-287 (Reference 6).

Shop hydrostatic test was then performed at a pressure of 3110 psig and a temperature of 300°F.

The safe end was then machined for the piping side weld preparation.

The machined safe end was welded to a long segment of stainless steel piping using a stainless steel weld.

A plant hydrostatic test was performed at 2485 psig pressure with a temperature of 300°F.

After the plant hydrostatic test, normal operating temperature and pressure were uniformly applied three times to consider any shakedown effects, after which the model was set to normal operating conditions.

Based on the FEA model, residual stresses at the centerline of the DM weld were obtained. The recommended axial and hoop stress profiles were used in the generation of the crack growth charts to determine the maximum allowable initial flaw sizes (Section 7.0). The hoop and axial welding residual stresses for the recommended stress profiles at 100% normal operating conditions (operating pressure and temperature) are shown in Figure 4-2.

• • • This record was final approved on 7/19/2018 6:03:38 PM. (This statement was added by the PRIME system upon its validation)

RA-19-0138 Attc1chment 3 Westinghouse Non-Proprietary Class 3 Page 7 of21 LTR-SDA-18-016-NP Revision 0 July 2018 Dissimilar Metal Weld Nozzle Forging Safe-End Butter Figure 4-1: H.B. Robinson Unit 2 Reactor Vessel Inlet Nozzle DM Weld Configuration

• • • This record was final approved on 7/19/2018 6:03:38 PM. (This statement was added by the PRIME system upon its validation)

RA-19-0138 Attachment 3 Westinghouse Non-Proprietary Class 3 Page 8 of 21 LTR-SDA-18-016-NP Revision 0 July 2018 a,c,e Figure 4-2: H. B. Robinson Unit 2 Reactor Vessel Inlet Nozzle DM Weld 100% Normal Operating Residual Stress Profiles Through DM Weld with 50% Inside Surface Weld Repair

• • • This record was final approved on 7/19/2018 6:03:38 PM. (This statement was added by the PRIME system upon its validation)

RA-19-0138 Attachment 3 Westinghouse Non-Proprietary Class 3 Page 9 of 21 LTR-SDA-18-016-NP Revision 0 July 2018 5.0 Maximum Allowable End-of-Evaluation Period Flaw Size Determination In order to develop the technical justification to defer the volumetric examination of the RV inlet nozzle DM welds from Fall 2020 RFO to Fall 2022 RFO, the first step is the determination of the maximum allowable end-of-evaluation period flaw sizes. The maximum allowable end-of-evaluation period flaw size is the size to which an indication is allowed to grow until the next inspection or evaluation period.

This particular flaw size is determined based on the piping loads, geometry and the material properties of the component. The evaluation guidelines and procedures for calculating the maximum allowable end-of-evaluation period flaw sizes are described in paragraph IWB-3640 and Appendix C of the ASME Section XI Code (Reference 2).

Rapid, nonductile failure is possible for ferritic materials at low temperatures, but is not applicable to the nickel-base alloy material. In nickel-base alloy material, the higher ductility leads to two possible modes of failure, plastic collapse or unstable ductile tearing. The second mechanism can occur when the applied J integral exceeds the JIc fracture toughness, and some stable tearing occurs prior to failure. If this mode of failure is dominant, then the load-carrying capacity is less than that predicted by the plastic collapse mechanism. The maximum allowable end-of-evaluation period flaw sizes of paragraph IWB-3640 for the high toughness materials are determined based on the assumption that plastic collapse would be achieved and would be the dominant mode of failure. However, due to the reduced toughness of the DM welds, it is possible that crack extension and unstable ductile tearing could occur and be the dominant mode of failure. To account for this effect, penalty factors called Z factors were developed in ASME Code Section XI, which are to be multiplied by the loadings at these welds. In the current analysis for H. B.

Robinson, Z factors based on Reference 7 are used in the analysis to provide a more representative approximation of the effects of the DM welds. The Z-factors for Alloy 82/182 from Reference 7 have been incorporated into the ASME Section XI 2007 Edition with 2009 Addenda; this particular code edition has been approved for use in 10CFR50.55a. The use of Z factors in effect reduces the maximum allowable end-of-evaluation period flaw sizes for flux welds and thus has been incorporated directly into the evaluation performed in accordance with the procedure and acceptance criteria given in IWB-3640 and Appendix C of ASME Code Section XI. It should be noted that the maximum allowable end-of-evaluation period flaw sizes are limited to only 75% of the wall thickness in accordance with the requirements of ASME Section XI paragraph IWB-3640 (Reference 2).

The maximum allowable end-of-evaluation period flaw sizes determined for both axial and circumferential flaws have incorporated the relevant material properties, pipe loadings and geometry.

Loadings under normal, upset, emergency and faulted conditions are considered in conjunction with the applicable safety factors for the corresponding service conditions required in the ASME Section XI Code.

For circumferential flaws, axial stress due to the pressure, deadweight, thermal expansion and seismic loads are considered in the evaluation. As for the axial flaws, hoop stress resulting from pressure loading is used.

The maximum allowable end-of-evaluation period flaw sizes for the axial and circumferential flaws at the RV inlet nozzle DM welds are provided in Table 5-1. The maximum allowable end-of-evaluation period axial flaw size was calculated with an assumed aspect ratio (flaw length/flaw depth) of 2. The aspect ratio of 2 is reasonable because the axial flaw growth due to PWSCC is limited to the width of the DM weld configuration. For the circumferential flaw, a conservative aspect ratio of 10 is used.

• • • This record was final approved on 7/19/2018 6:03:38 PM. (This statement was added by the PRIME system upon its validation)

RA-19-0138 Attachment 3 Westinghouse Non-Proprietary Class 3 Page 10 of 21 LTR-SDA-18-016-NP Revision 0 July 2018 It should be noted that the resulting maximum allowable end-of-evaluation period flaw sizes were limited by the ASME Code limit of 75% of the weld thickness for both flaw configurations.

Table 5-1 Maximum End-of-Evaluation Period Allowable Flaw Sizes (Flaw Depth/Wall Thickness Ratio - a/t)

Axial Flaw Circumferential Flaw (Aspect Ratio = 2) (Aspect Ratio = 10) 0.75 0.75

• • • This record was final approved on 7/19/2018 6:03:38 PM. (This statement was added by the PRIME system upon its validation)

RA-19-0138 Attachment 3 Westinghouse Non-Proprietary Class 3 Page 11 of 21 LTR-SDA-18-016-NP Revision 0 July 2018 6.0 PWSCC Growth Analysis A PWSCC growth analysis was performed to determine the maximum allowable initial flaw size that would be acceptable based on ASME Section XI acceptance criteria (Reference 2) for the operating duration from the Fall 2013 to the Fall 2022 RFOs. The maximum allowable initial flaw size for the given plant operation duration is determined by subtracting the crack growth due to PWSCC for the specific plant operation duration from the maximum allowable end-of-evaluation period flaw size shown in Table 5-1.

Crack growth due to PWSCC is calculated for both axial and circumferential flaws using the normal operating condition steady-state stresses. For axial flaws, the stresses included pressure and residual hoop stresses, while for circumferential flaws, the stresses considered are pressure, 100% power normal thermal expansion, deadweight and residual axial stresses. The input required for the crack growth analysis is basically the information necessary to calculate the crack tip stress intensity factor (KI), which depends on the geometry of the crack, its surrounding structure and the applied stresses. The geometry and loadings for the nozzles of interest are discussed in Section 3.0 and the applicable residual stresses used are discussed in Section 4.0. Once KI is calculated, PWSCC growth can be calculated using the applicable crack growth rate for the nickel-base alloy material (Alloy 182) from MRP-115 (Reference 3),

which is also documented in ASME Section XI (Reference 2). For all inside surface flaws, the governing crack growth mechanism for the RV inlet nozzle is PWSCC.

Using the applicable stresses at the DM welds, the crack tip stress intensity factors can be determined based on the stress intensity factor expressions from API-579 (Reference 8). The through-wall stress distribution profile is represented by a 4th order polynomial:

2 3 4

= 0 + 1 xt + 2 xt + 3 xt + 4 xt Where:

0, 1, 2, 3, and 4 are the stress profile curve fitting coefficients; x = the distance from the wall surface where the crack initiates to the crack tip; t = the wall thickness; and

= the stress perpendicular to the plane of the crack.

The stress intensity factor calculations for semi-elliptical inside surface axial and circumferential flaws are expressed in the general form as follows:

4 a a j KI = Gj ac , at , tR , j Q t j=0 Where:

a = Crack depth c = Half crack length along surface t = Thickness of cylinder

• • • This record was final approved on 7/19/2018 6:03:38 PM. (This statement was added by the PRIME system upon its validation)

RA-19-0138 Attachment 3 Westinghouse Non-Proprietary Class 3 Page 12 of 21 LTR-SDA-18-016-NP Revision 0 July 2018 R = Inside radius

= Angular position of a point on the crack front. The deepest point ( = /2) along the crack front is considered in the analysis Gj = Gj is influence coefficient for jth stress distribution on crack surface (i.e., G0, G1, G2, G3, G4)

Q = The shape factor of an elliptical crack is approximated by:

Q = 1 + 1.464(a/c)1.65 for a/c < 1 or Q = 1 + 1.464(c/a)1.65 for a/c > 1 Once the crack tip stress intensity factors are determined, PWSCC growth calculations can be performed using the crack growth rate below with the applicable normal operating temperature.

The PWSCC growth rate used in the crack growth analysis is based on the Electric Power Research Institute (EPRI) recommended crack growth curve for Alloy 182 material (Reference 3):

da Qg 1 1

=exp - - K dt R T Tref Where:

da

= Crack growth rate in m/sec (in/hr) dt Qg = Thermal activation energy for crack growth = 130 kJ/mole (31.0 kcal/mole)

R = Universal gas constant = 8.314 x 10-3 kJ/mole-K (1.103 x 10-3 kcal/mole-°R)

T = Absolute operating temperature at the location of crack, K (°R)

Tref = Absolute reference temperature used to normalize data = 598.15 K (1076.67°R)

= Crack growth amplitude

= 1.50 x 10-12 at 325°C (2.47 x 10-7 at 617F)

= Exponent = 1.6 K = Crack tip stress intensity factor MPam (ksiin)

The normal operating temperature used in the crack growth analysis is 547.6F at the RV inlet nozzle. It should be noted that the fatigue crack growth mechanism is not considered in the crack growth analysis as it is considered to be small when compared to the crack growth due to the PWSCC growth mechanism at the reactor vessel inlet nozzle for the duration of interest. This is demonstrated by the low fatigue usage factor of 0.0007 at the inlet nozzle location of interest in the reactor vessel analytical report CENC-1111 (Reference 9). Therefore, it is not necessary to consider fatigue crack growth in the evaluation.

The PWSCC growth rate is highly dependent on the temperature at the location of the flaw, furthermore, the crack growth rate increases as the temperature increases. Therefore, during periods when the plant is not in operation, such as refueling outages or shutdowns, the temperature at the reactor vessel nozzles is low such that crack growth due to PWSCC is insignificant. Therefore, the PWSCC growth calculation should be determined for the time interval when the plant is operating at full power. The amount of time

• • • This record was final approved on 7/19/2018 6:03:38 PM. (This statement was added by the PRIME system upon its validation)

RA-19-0138 Attachment 3 Westinghouse Non-Proprietary Class 3 Page 13 of 21 LTR-SDA-18-016-NP Revision 0 July 2018 when the plant is operating at full power is determined based on previous plant operation data and the anticipated outages scheduled until the next inspections. This operation duration at full power is referred to as Effective Full Power Years (EFPY). However, the analysis herein will conservatively use a 100%

capacity factor (i.e. plant operating at full power); thus, the operation duration between Fall 2013 and Fall 2022 will be 9 EFPY.

• • • This record was final approved on 7/19/2018 6:03:38 PM. (This statement was added by the PRIME system upon its validation)

RA-19-0138 Attachment 3 Westinghouse Non-Proprietary Class 3 Page 14 of 21 LTR-SDA-18-016-NP Revision 0 July 2018 7.0 Technical Justification for Deferring the Volumetric Examination In accordance with ASME Code Case N-770-2 (Reference 1), the volumetric examination interval for the unmitigated RV inlet nozzle to safe end DM welds must not exceed 7 years. H. B. Robinson Unit 2 is seeking relaxation from the ASME Code Case N-770-2 requirement in order to defer the volumetric examination of the reactor vessel inlet nozzle to safe end DM welds from the Fall 2020 to Fall 2022 RFO.

Technical justification can be developed to support deferring the volumetric examination by calculating the maximum allowable initial flaw size that could be left behind in service and remain acceptable between the inspections. This maximum allowable initial flaw size can then be compared to a flaw size which would have been detected during the Fall 2013 RFO inlet nozzle DM weld examination based on the inspection detection capability.

The maximum allowable initial flaw depth is determined by subtracting the amount of PWSCC growth for a duration of 9 effective full power years (EFPY) from the maximum allowable end-of-evaluation period flaw depth shown in Table 5-1. The end-of-evaluation period flaw depth is calculated based on the guidelines given in paragraph IWB-3640 and Appendix C of the ASME Section XI Code (Reference 2).

The PWSCC growth at the Alloy 82/182 weld for the postulated axial flaw is calculated based on normal operating welding residual hoop stresses, while for circumferential flaw, the stresses considered normal operating piping loads (deadweight and thermal expansion) and normal operating welding residual axial stresses. The PWSCC crack growth model is based on MRP-115 (Reference 3). The maximum allowable initial flaw depth was calculated for an axial flaw with an assumed aspect ratio of 2. An aspect ratio of 2 is reasonable for the axial flaw due to the DM weld configuration since any PWSCC axial flaw growth is limited to the width of the weld. For the circumferential flaw, a conservative aspect ratio of 10 is used in the crack growth analysis.

The PWSCC growth analysis of the circumferential flaws considered two cases. The first case is normal operating piping loads (deadweight and normal thermal loads) with residual stresses from the profile shown in Figure 4-2. The second case is normal operating piping loads without residual stresses in order to obtain the most limiting crack growth results since a portion of the axial residual stress profile is compressive. It was determined that the case which included only normal operating piping loads without residual stresses was limiting for circumferential flaws. The exclusion of welding residual stresses in the evaluation is conservative for the circumferential flaw evaluation.

The PWSCC growth curves and the maximum allowable initial flaw sizes for an axial flaw and a circumferential flaw are shown in Figures 7-1 and 7-2, respectively. The horizontal axis displays service life in EFPY, and the vertical axis shows the flaw depth to wall thickness ratio (a/t). The maximum allowable end-of-evaluation period flaw sizes are also shown in these figures for the respective flaw configurations. Based on the crack growth results from Figures 7-1 and 7-2, the maximum allowable initial flaw sizes for the axial and circumferential flaws are tabulated in Table 7-1 for 9 EFPY.

• • • This record was final approved on 7/19/2018 6:03:38 PM. (This statement was added by the PRIME system upon its validation)

RA-19-0138 Attachment 3 Westinghouse Non-Proprietary Class 3 Page 15 of 21 LTR-SDA-18-016-NP Revision 0 July 2018 Table 7-1 H. B. Robinson Unit 2 Maximum Allowable Initial Flaw Sizes for 9 EFPY Axial Flaw Circumferential Flaw (Aspect Ratio = 2) (Aspect Ratio = 10)

Maximum Allowable Initial 0.114 0.394 Flaw Size (a/t)

Flaw Depth (in) 0.285 0.985 Flaw Length (in) 0.568 9.85 Note: Aspect ratio = flaw length/flaw depth Wall thickness (t) = 2.5 inches The flaw sizes shown in Table 7-1 are the largest axial and circumferential flaw sizes that could be left behind in service and remain acceptable from the Fall 2013 to Fall 2022 RFO (9 EFPY with 100%

capacity factor) for H. B. Robinson Unit 2. In accordance with the Ultrasonic Testing (UT) detection and sizing requirements in ASME Section XI Appendix VIII, Supplement 10 (Reference 2), the minimum required detectable flaw depth is 10% of the wall thickness. Therefore, the maximum allowable initial axial and circumferential flaw sizes are above the minimum flaw depth requirement per the UT detection capabilities, and thus would have been reasonably detected at the previous inspection of the DM welds.

[

• • • This record was final approved on 7/19/2018 6:03:38 PM. (This statement was added by the PRIME system upon its validation)

RA-19-0138 Attachment 3 Westinghouse Non-Proprietary Class 3 Page 16 of 21 LTR-SDA-18-016-NP Revision 0 July 2018

]a,c,e The evaluation in this report provides the technical basis to then further demonstrate that any inside surface connected flaws with flaw depth 11.4% of the wall thickness (Table 7-1) will be acceptable for continued operation for 9 EFPY. Since, there are no inside surface connected flaws detected in the cold leg, the technical justification developed in this letter report can be used to defer the volumetric examination for the H. B. Robinson Unit 2 RV inlet nozzle DM welds from the Fall 2020 RFO to the Fall 2022 RFO.

[

• • • This record was final approved on 7/19/2018 6:03:38 PM. (This statement was added by the PRIME system upon its validation)

RA-19-0138 Attachment 3 Westinghouse Non-Proprietary Class 3 Page 17 of 21 LTR-SDA-18-016-NP Revision 0 July 2018

]a,c,e Thus, based on the deterministic fracture mechanics analysis performed in this report along with the qualitative justification presented herein provides technical basis to defer the volumetric examination for Robinson Unit 2 inlet nozzle DM welds from Fall 2020 to Fall 2022 (total of 9 EFPY from the previous examination in Fall 2013).

• • • This record was final approved on 7/19/2018 6:03:38 PM. (This statement was added by the PRIME system upon its validation)

RA-19-0138 Attachment 3 Westinghouse Non-Proprietary Class 3 Page 18 of 21 LTR-SDA-18-016-NP Revision 0 July 2018

• • • This record was final approved on 7/19/2018 6:03:38 PM. (This statement was added by the PRIME system upon its validation)

--~------------~- ---/~ I 0.8

-- I I ASME Maximum Allowable End-Of-Evaluation Period Flaw Size 0.7

,,V I/

/

0.6

/'

Flaw Depth/Wall Thickness (a/t) 0.5

~

- /

Maximum Initial Flaw Size Allowed During I' 0.4 Fall 2013 RFO (a/t) = 0.114 /

I /

~

0.3 I V

,.V V l/

V I .,.,,..,.. ~

0.2 I

V i---

--~t- ~

0.1 9 EFPY to Fall 2022 RFO 0

0 1 2 3 4 5 6 7 8 9 10 Service Life (EFPY)

Figure 7-1: PWSCC Growth Curve for H. B. Robinson Unit 2 Inlet Nozzle Axial Flaw (DM weld), Aspect Ratio = 2

RA-19-0138 Attachment 3 Westinghouse Non-Proprietary Class 3 Page 19 of 21 LTR-SDA-18-016-NP Revision 0 July 2018

• • • This record was final approved on 7/19/2018 6:03:38 PM. (This statement was added by the PRIME system upon its validation)

I 0.8 I

ASME Maximum Allowable End-Of-Evaluation Period Flaw Size J

I 0.7 J

J I

0.6

,/ V Maximum Initial Flaw Size Allowed During 0.5 Fall 2013 RFO (a/t) = 0.394 V

Flaw Depth/Wall Thickness (a/t) i'----,..

r-----..

/

0.4 /

/

I/

J.,.. /

0.3 J......

~,,.

.... i...,,,,o" 0.2 i.---

9 EFPY to

~ Fall 2022 RFO

~

0.1 0

0 4 8 12 16 20 24 28 32 Service Life (EFPY)

Figure 7-2: PWSCC Growth Curve for H. B. Robinson Unit 2 Inlet Nozzle Circumferential Flaw (DM weld), Aspect Ratio = 10

RA-19-0138 Attachment 3 Westinghouse Non-Proprietary Class 3 Page 20 of 21 LTR-SDA-18-016-NP Revision 0 July 2018 8.0 Summary and Conclusions A volumetric examination of the reactor vessel inlet nozzle to safe end DM butt welds was performed during the Fall 2013 RFO at H. B. Robinson Unit 2. The next required volumetric examination is planned during the Fall 2020 RFO in accordance with ASME Code Case N-770-2 (Reference 1). However, the volumetric examination will be deferred to the Fall 2022 RFO for the reactor vessel inlet nozzle DM welds. Since the time interval between the previous examination and the planned examination exceeds 7 years, which deviates from the Code Case N-770-2 (Reference 1) inspection interval requirements, a flaw tolerance evaluation was completed to defer the volumetric examination of the inlet nozzle DM welds.

This letter report provides technical justification to support the relaxation request by performing a flaw tolerance analysis to determine the largest initial axial and circumferential flaws that could be left behind in service and remain acceptable between the planned examinations. This maximum allowable initial flaw size can then be compared to any flaw size which would have been detected during the previous inlet nozzle DM weld examinations.

Based on the PWSCC growth analysis results from Section 7.0 which is for a duration of 9 EFPY, the maximum allowable initial flaw sizes for the reactor vessel inlet nozzle DM welds are tabulated in Table 8-1 for H. B. Robinson Unit 2. These allowable initial axial and circumferential flaw sizes have been shown to be acceptable in accordance with the ASME Section XI IWB-3640 acceptance criteria through the Fall 2022 RFO for H. B. Robinson Unit 2 taking into account of potential PWSCC growth since the last volumetric and surface examinations.

In accordance with the Ultrasonic Testing (UT) detection and sizing requirements in ASME Section XI Appendix VIII, Supplement 10 (Reference 2), the minimum required detectable flaw depth is 10% of the wall thickness. Therefore, the maximum allowable initial axial and circumferential flaw size is above the minimum flaw depth requirement per the UT detection capabilities, and thus would have been reasonably detected at the previous inspection of the DM welds.

Based on the H. B. Robinson Unit 2 results in Table 8-1, the calculated maximum allowable initial axial flaw size (flaw depth = 0.285 and flaw length = 0.568) for 9 EFPY is large enough to have been detected during the last Fall 2013 RFO examination of the RV inlet nozzle DM welds. Therefore, deferring the volumetric examination for the H. B. Robinson Unit 2 RV inlet nozzle DM welds from the Fall 2020 RFO allowed by Code Case N-770-2 to the Fall 2022 RFO is technically justified.

Table 8-1 H. B. Robinson Unit 2 Maximum Allowable Initial Flaw Sizes based on 9 EFPY Axial Flaw Circumferential Flaw (Aspect Ratio = 2) (Aspect Ratio = 10)

Maximum Allowable Initial 0.114 0.394 Flaw Size (a/t)

Flaw Depth (in) 0.285 0.985 Flaw Length (in) 0.568 9.85 Note: Aspect ratio = flaw length/flaw depth Wall thickness (t) = 2.5 inches

• • • This record was final approved on 7/19/2018 6:03:38 PM. (This statement was added by the PRIME system upon its validation)

RA-19-0138 Attachment 3 Westinghouse Non-Proprietary Class 3 Page 21 of 21 LTR-SDA-18-016-NP Revision 0 July 2018 9.0 References

1. ASME Code Case N-770-2,Section XI Division 1, Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated with UNS N06082 or UNS W86182 Weld Filler Material With or Without Application of Listed Mitigation Activities, Approval Date June 9, 2011.

2. ASME Boiler & Pressure Vessel Code, 2007 Edition with 2008 Addenda,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components.

3. Materials Reliability Program: Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking (PWSCC) of Alloy 82, 182, and 132 Welds (MRP-115), EPRI, Palo Alto, CA: 2004.

1006696.

4. Drawings for RV inlet nozzles:

a. Combustion Engineering, Inc. Drawing E-232-275, Revision 11, Pressure Vessel Final Machining.

b. Combustion Engineering, Inc. Drawing E-232-276, Revision 3, Nozzle Assembly and Details.

c. Combustion Engineering, Inc. Drawing E-232-281, Revision 1, Material Identification (Pressure Vessel) Westinghouse Electric Corp. 155 1/2 I.D. Reactor Vessel.

5. Dominion Engineering, Inc. Calculation Note, C-8827-00-02, Revision 0, Welding Residual Stress Calculation for H. B. Robinson RPV Inlet Nozzle, Including Weld Repair.

6. Materials Reliability Program: Primary Water Stress Corrosion Cracking (PWSCC) Flaw Evaluation Guidance (MRP-287). EPRI, Palo Alto, CA: 2010. 1021023.

7. Materials Reliability Program: Advanced FEA Evaluation of Growth of Postulated Circumferential PWSCC Flaws in Pressurizer Nozzle Dissimilar Metal Welds (MRP-216, Rev. 1): Evaluations Specific to Nine Subject Plants. EPRI, Palo Alto, CA: 2007. 1015400.

8. American Petroleum Institute, API-579-1/ASME FFS-1 (APS 579 Second Edition), Fitness-For-Service, June 5, 2007.

9. Combustion Engineering, Inc. Report, CENC-1111, Analytical Report for Carolina Power and Light Reactor Vessel, January 1967.

10. Material Reliability Program: Primary System Piping Butt Weld Inspection and Evaluation Guideline (MRP-139, Revision 1). EPRI, Palo Alto, CA: 2008. 1015009.

11. [

] a,c,e

12. [

]a,c,e

13. ASME Codes & Standards, Record #09-1021. Technical Basis Document for Alloy 82/182 Weld Inspection Code Case N-770 and N-770-1, Authors: P. Donavin, G. Elder, W. Bamford. Date:

8/10/09.

14. [

] a,c,e

• • • This record was final approved on 7/19/2018 6:03:38 PM. (This statement was added by the PRIME system upon its validation)

RA-19-0138 Attachment 4 Duke Energy Progress, LLC H.B. Robinson, Unit 2 Relief Request RA-19-0138 Westinghouse Application for Withholding Proprietary Information from Public Disclosure CAW-18-4776, accompanying Affidavit, Proprietary Information Notice, and Copyright Notice.

RA-19-0138 Westinghouse Non-Proprietary Class 3

@ Westinghouse Westinghouse Electric Company 1000 Westinghouse Drive Cranberry Township, Pennsylvania 16066 USA U.S. Nuclear Regulatory Commission Direct tel: (412) 374-5541 Document Control Desk Direct fax: (724) 940-8542 11555 Rockville Pike e-mail: mercieej@westinghouse.com Rockville, MD 20852 CAW-18-4776 July 12, 2018 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE

Subject:

LTR-SDA-18-016-P, Revision 0, "Technical Justification to Support the Extended Volumetric Examination Interval for H.B. Robinson Unit 2 Reactor Vessel Inlet Nozzle to Safe End Dissimilar Metal Welds" (Proprietary)

The Application for Withholding Proprietary Information from Public Disclosure is submitted by Westinghouse Electric Company LLC ("Westinghouse"), pursuant to the provisions of paragraph (6)(1) of Section 2.390 of the Nuclear Regulatory Commission's ("Commission's") regulations. It contains commercial strategic information proprietary to Westinghouse and customarily held in confidence.

The proprietary information for which withholding is being requested in the above-referenced report is further identified in Affidavit CAW-18-4776 signed by the owner of the proprietary information, Westinghouse. The Affidavit, which accompanies this letter, sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CPR Section 2.390 of the Commission's regulations.

Accordingly, this letter authorizes the utilization of the accompanying Affidavit by Duke Energy Progress.

• Correspondence with respect to the proprietary aspects of the Application for Withholding or the Westinghouse Affidavit should reference CAW-18-4776, and should be addressed to Edmond J. Mercier, Manager, Fuels Licensing and Regulatory Support, Westinghouse Electric Company, 1000 Westinghouse Drive, Building 2 Suite 256, Cranberry Township, Pennsylvania 16066.

Edmond J. Mercier, Manager Fuels Licensing and Regulatory Support

© 2018 Westinghouse Electric Company LLC. All Rights Reserved.

RA-19-0138 CAW-18-4776 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:

ss COUNTY OF BUTLER:

I, Edmond J. Mercier, am authorized to execute this Affidavit on behalf of Westinghouse Electric Company LLC ("Westinghouse") and declare that the averments of fact set forth in this Affidavit are,true and correct to the best of my knowledge, information, and belief.

Executed on: 7/ 12--/ V I 1 Edmond J. Merci Manager Fuels Licensing and Regulatory Support

RA-19-0138 3 CAW-18-4776 (1) I am Manager, Fuels Licensing and Regulatory Support, Westinghouse Electric Company LLC

("Westinghouse"), and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rule making proceedings, and am authorized to apply for its withholding on behalf of Westinghouse.

(2) I am making this Affidavit in conformance with the provisions of 10 CPR Section 2.390 of the Nuclear Regulatory Commission's ("Commission's") regulations and in conjunction with the Westinghouse Application for Withholding Proprietary Information from Public Disclosure accompanying this Affidavit.

(3) I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged or as confidential commercial or financial information.

(4) Pursuant to the provisions of paragraph (b)(4) of Section 2.390 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the_

information sought to be withheld from public disclosure should be withheld.

(i) The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse.

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(b) It is information that is marketable in many ways. The extent to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information.

(c) Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense.

RA-19-0138 5 CAW-18-4776 (d) Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage.

(e) Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition of those countries.

(f) The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage.

(iv) The information is being transmitted to the Commission in confidence and, under the provisions of 10 CFR Section 2.390, is to be received in confidence by the Commi~sion.

(v) The information sought to be protected is not available in public sources or available information has not been previously employed in the same original manner or method to the best of our knowledge and belief.

(vi) The proprietary information sought to be withheld in this submittal is that which is appropriately marked in LTR-SDA-18-016-P, "Technical Justification to Support the Extended Volumetric Examination Interval for H.B. Robinson Unit 2 Reactor Vessel Inlet Nozzle to Safe End Dissimilar Metal Welds" (Proprietary), for submittal to the Commission, being transmitted by Duke Energy Progress letter. The proprietary information as submitted by Westinghouse is that associated with the technical justification to support the extended volumetric examination interval for H.B. Robinson Unit 2 reactor vessel inlet nozzle to safe end dissimilar metal welds, and may be used only for that purpose.

(a) This information is part of that which will enable Westinghouse to provide technical justification to support the extended volumetric examination interval

RA-19-0138 6 CAW-18-4776 for H.B. Robinson Unit 2 reactor vessel inlet nozzle to safe end dissimilar metal welds.

(b) Further, this information has substantial commercial value as follows:

(i) Westinghouse plans to sell the use of similar information to its customers for the purpose of providing technical justification to support the extended volumetric examination interval for H.B. Robinson Unit 2 reactor vessel inlet nozzle to safe end dissimilar metal welds.

(ii) Westinghouse can sell support and defense of industry guidelines and acceptance criteria for plant-specific applications.

(iii) The information requested to be withheld reveals the distinguishing aspects of a methodology which was developed by Westinghouse.

Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar technical evaluation justifications and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.

The development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Westinghouse effort and the expenditure of a considerable sum of money.

In order for competitors of Westinghouse to duplicate this information, similar technical programs would have to be performed and a significant manpower effort, having the requisite talent and experience, would have to be expended.

Further the deponent sayeth not.

RA-19-0138 PROPRIETARY INFORMATION NOTICE Transmitted herewith are proprietary and non-proprietary versions of a document, furnished to the NRC in connection with requests for generic and/or plant-specific review and approval.

In order to conform to the requirements of 10 CFR 2.390 of the Commission's regulations concerning the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets, and where the proprietary information has been deleted in the non-proprietary versions, only the brackets remain (the information that was contained within the brackets in the proprietary versions having been deleted). The justification for claiming the information so designated as proprietary is indicated in both versions by means of lower case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (4)(ii)(a) through (4)(ii)(f) of the Affidavit accompanying this transmittal pursuant to 10 CFR 2.390(b)(l).

COPYRIGHT NOTICE The reports transmitted herewith each bear a Westinghouse copyright notice. The NRC is permitted to make the number of copies of the information contained in these reports which are necessary for its internal use in connection with generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding. With respect to the non-proprietary versions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.

RA-19-0138 Duke Energy Progress Letter for Transmittal to the NRC The following paragraphs should be included in your letter to the NRC Document Control Desk:

Enclosed are:

1. LTR-SDA-18-016-P, Revision 0, "Technical Justification to Support the Extended Volumetric Examination Interval for H.B. Robinson Unit 2 Reactor Vessel Inlet Nozzle to Safe End Dissimilar Metal Welds" (Proprietary) 2.

3. LTR-SDA-18-016-NP, Revision 0, Technical Justification to Support the Extended Volumetric Examination Interval for H.B. Robinson Unit 2 Reactor Vessel Inlet Nozzle to Safe End Dissimilar Metal Welds" (Non-Proprietary)

Also enclosed are the Westinghouse Application for Withholding Proprietary Information from Public Disclosure CAW-18-4 776, accompanying Affidavit, Proprietary Information Notice, and Copyright Notice.

As Item 1 contains information proprietary to Westinghouse Electric Company LLC ("Westinghouse"), it is supported by an Affidavit signed by Westinghouse, the owner of the information. The Affidavit sets forth the basis on which the information may be withheld from public disclosure by the Nuclear Regulatory Commission ("Commission") and addresses with specificity the considerations listed in' paragraph (b)(4) of Section 2.390 of the Commission's regulations.

Accordingly, it is respectfully requested that the information which is proprietary to Westinghoµse be withheld from public disclosure in accordance with 10 CFR Section 2.390 of the Commission's regulations.

Correspondence with respect to the copyright or proprietary aspects of the items listed above or the supporting Westinghouse Affidavit should reference CAW-18-4776 and should be addressed to Edmond J. Mercier, Manager, Fuels Licensing and Regulatory Support, Westinghouse Electric Company, 1000 Westinghouse Drive, Building 2 Suite 256, Cranberry Township, Pennsylvania 16066.