ET 23-0005, 10 CFR 50.55a Request Number I4R-08 for the Fourth Inservice Inspection Program Interval, Relief for Extension of Follow Up Examination Requirements for Reactor Pressure Vessel Head Penetration Nozzles with Mitigated Alloy 600/82/182 Peen

10 CFR 50.55a Request Number I4R-08 for the Fourth Inservice Inspection Program Interval, Relief for Extension of Follow Up Examination Requirements for Reactor Pressure Vessel Head Penetration Nozzles with Mitigated Alloy 600/82/182 Peened

ML23075A048
Person / Time
Site:	Wolf Creek
Issue date:	03/16/2023
From:	Boyce M Wolf Creek
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
ET 23-0005
Download: ML23075A048 (1)

Text

W o If C re e k -u\)sJffJl'f-Nuclear Operating Corporation Michael T. Boyce Vice President Engineering March 16, 2023 ET 23-0005 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 - 0001

Subject:

Docket No. 50-482: 10 CFR 50.55a Request Number I4R-08 for the Fourth lnservice Inspection Program Interval, Relief for Extension of Follow up Examination Requirements for Reactor Pressure Vessel Head Penetration Nozzles with Mitigated Alloy 600/82/182 Peened Surface Commissioners and Staff:

Pursuant to 10 CFR 50.55a(z)(2), Wolf Creek Nuclear Operating Corporation (WCNOC) hereby requests Nuclear Regulatory Commission (NRC) approval of 10 CFR 50.55a Request Number I4R-08 for the fourth ten-year interval of WCNOC's lnservice Inspection (ISi) Program.

During Refueling Outage 24 (RF24), which occurred during Spring 2021, WCNOC implemented the Ultra-High-Pressure Cavitation Peening (UHPCP) process on Reactor Pressure Vessel Head Penetration Nozzles (RPVHPNs) with Alloy 600/82/182 surfaces. In the Attachment to this submittal, WCNOC is requesting a change to the examination interval of the follow-up inspections for peened RPVHPNs and associated welds.

WCNOC requests approval by October 1, 2023, to allow for planning RF26 (Spring 2024) and RF27 (Fall 2025).

This letter contains no commitments. If you have any questions concerning this matter, please contact me at (620) 364-8831 x8687, or Dustin Hamman at (620) 364-4204.

Sincerely, Michael T. Boyce P.O. Box 411 I Burlington, KS 66839 I 620-364-8831

ET 23-0005 Page 2 of 2 MTB/jkt

Attachment:

10 CFR 50.55a Request I4R-08 cc: S. S. Lee (NRC), w/a R. J. Lewis (NRC), w/a G. E. Werner (NRC), w/a Senior Resident Inspector (NRC), w/a

Attachment to ET 23-0005 Page 1 of 9 Wolf Creek Nuclear Operating Corporation 10 CFR 50.55a Request I4R-08 Request for Relief for Extension of Follow up Examination Requirements for Reactor Pressure Vessel Head Penetration Nozzles with Mitigated Alloy 600/82/182 Peened Surface in Accordance with 10 CFR 50.55a(z)(2)

Attachment to ET 23-0005 Page 2 of 9 10 CFR 50.55a Request Number I4R-08 Request for Relief for Extension of Follow up Examination Requirements for Reactor Pressure Vessel Head Penetration Nozzles with Mitigated Alloy 600/82/182 Peened Surface in Accordance with 10 CFR 50.55a(z)(2)

Hardship Without a Compensating Increase in Quality and Safety 1.0 ASME CODE COMPONENT AFFECTED Component: Reactor Vessel Closure Head (RVCH) Nozzles Code Class: Class 1 Examination Category: Class 1 PWR Reactor Vessel Upper Head Code Item Number: B4.50 B4.60

Description:

Control Rod Drive Mechanism (CRDM) Nozzles Core Exit Thermocouple Nozzle Assy (CETNA) Nozzles Head Vent Pipe Size: 4.00 Inch (Nominal Outside Diameter - CRDM and CETNA)

NPS 1 inch schedule 160 pipe (Vent)

Material: RVCH SA533 Grade B, Class 1 Nozzle SB 167 N06600 (Alloy 600)

Alloy 82/182 weld material 2.0 APPLICABLE CODE EDITION AND ADDENDA Inservice Inspection (ISI) and Repair/Replacement Programs: American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, 2007 Edition through 2008 Addenda [1]. Examinations of the RVCH nozzles are performed in accordance with 10 CFR 50.55a(g)(6)(ii)(D), which specifies the use of ASME Code Case N-729-6 [2]

with conditions.

Attachment to ET 23-0005 Page 3 of 9 3.0 APPLICABLE CODE REQUIREMENTS 10 CFR 50.55a(g)(6)(ii)(D)(1) requires that examinations of the reactor vessel head be performed in accordance with ASME Code Case N-729-6 subject to the conditions specified in paragraphs 10 CFR 50.55a(g)(6)(ii)(D)(2) through (8).

10 CFR 50.55a(g)(6)(ii)(D)(5) states the following:

Peening. In lieu of inspection requirements of Table 1, Items B4.50 and B4.60, and all other requirements in ASME BPV Code Case N-729-6 pertaining to peening, in order for a RPV upper head with nozzles and associated with J-groove welds mitigated by peening to obtain examination relief from the requirements of Table 1 for unmitigated heads, peening must meet the performance criteria, qualification, and examination requirements stated in MRP-335, Revision 3-A, with the exception that a plant-specific alternative request is not required and NRC condition 5.4 of MRP-335, Revision 3-A does not apply.

MRP-335, Revision 3-A (hereafter known as MRP-335 R3-A) [3] requires a follow-up examination to be performed in the second refueling outage (RFO) subsequent to peening for plants with heads that the reactor pressure vessel head penetration nozzles (RPVHPNs) and associated J-groove welds have experienced effective degradation years (EDY) < 8 and all RPVHPNs are free of pre-peening flaws.

4.0 REASON FOR REQUEST

1. Wolf Creek Nuclear Operating Corporation (WCNOC) performed ultra-high-pressure cavitation peening (UHPCP) on the RPVHPNs and associated J-groove welds in Refueling Outage 24 which occurred in Spring, 2021. The UHPCP was performed in accordance with the requirements of MRP-335 R3-A [3].

2. 10 CFR 50.55a(g)(6)(ii)(D)(5) requires that in lieu of the requirements of Table 1, Items B4.50, and B4.60, and all other requirements in Code Case N-729-6 [2] applicable to peening, peening must meet the performance criteria, qualification, and examination requirements stated in MRP-335 R3-A [3]. For reactor pressure vessel head penetration nozzles with EDY < 8, MRP-335 R3-A [3] requires a follow-up volumetric or surface examination, and a demonstrated volumetric or surface leak path assessment in the second refueling outage after peening (i.e., N+2). However, the warranty from the vendor that performed the UHPCP application specifies that a follow-up volumetric examination be conducted in the third RFO after peening (N+3).

3. WCNOC requests that the follow-up volumetric examination specified by MRP-335 R3-A [3] for the second (N+2) RFO be extended by one cycle to the third (N+3) RFO to align scheduled inspections. Performing the follow-up examinations during the N+3 refueling outage (18-months after the N+2 refueling outage) will permit any previously undetected

Attachment to ET 23-0005 Page 4 of 9 Primary Water Stress Corrosion Cracking (PWSCC) to grow sufficiently to be detected without increasing the likelihood of a through-wall crack (see 5.2).

4. Performing volumetric examinations during two sequential RFOs would add unnecessary dose for support personnel and would be contrary to As-Low-As Reasonably Achievable (ALARA) program practices. The RPVHPNs identified in Section 1 are located in a Locked High Radiation Area (LHRA) inside containment. The volumetric examination evolution results in personnel radiological exposure while setting up the equipment, conducting the examination, and demobilizing the equipment. Based on historical data at WCGS, the estimated increase in occupational dose due to performing RPVHPN volumetric examination in two separate outages would be approximately 732 mRem. A higher dose would be expected if tool breakdowns or issues requiring additional LHRA entry were to occur.

Conducting two volumetric examinations would also increase the potential for industrial safety issues and contamination exposure. Therefore, WCGS concludes that performing RPVHPN volumetric examinations during the second and third RFOs after peening would present a hardship without commensurate safety benefit.

5.0 PROPOSED ALTERNATIVE AND BASIS FOR USE 5.1 Proposed Alternative

1. MRP-335 R3-A [3] specifies that for reactor vessel heads operating at reactor cold leg temperature (Tcold) (EDY < 8) with no previously detected PWSCC that follow-up volumetric examinations be performed in the second (N+2) refueling outage after the baseline (pre-peening) inspection and peening are performed. As an alternative to the requirements of 10 CFR 50.55a(g)(6)(ii)(D) WCNOC proposes that the follow-up volumetric examination be conducted in the third (N+3) refueling outage after peening.

This alternative corresponds to a nominal 4.5 calendar years since the last volumetric examination for these nozzles.

2. Supporting supplemental evaluations include:

• MRP-335 R3-A, "Materials Reliability Program: Topical Report for Primary Water Stress Corrosion Cracking Mitigation by Surface Stress Improvement," November 2016 [3].

• Dominion Engineering, Inc (DEI) authored Technical Note TN-4069 01, Revision 0, "MRP-335 R3-A Matrix of Deterministic Crack Growth Calculations for Tcold Reactor Vessel Top Head Nozzles Evaluated for Alternative Peening Follow-up Volumetric Examination Timing," August 2018 [5]. Reference [5] has been submitted to the NRC in its entirety in Precedents 2.a and 3.a.

Attachment to ET 23-0005 Page 5 of 9 5.2 Basis for Assessments and Supplemental Evaluation

1. The primary degradation mechanism addressed by 10 CFR 50.55a(g)(6)(ii)(D) is PWSCC.

This degradation mechanism occurs when a susceptible material is exposed to primary water environment, elevated tensile stress levels and elevated operating temperatures.

Previous examinations for the WCGS RPV upper head have identified no indications of cracking attributed to PWSCC. The WCGS RPV upper head was peened as a pre-emptive mitigation measure during the Spring 2021 RFO, and the UHPCP application followed the process of MRP-335 Revision 3-A without any deviations. A pre-peening volumetric examination was conducted during this RFO in accordance with Section 2.5.1 of MRP-335 Revision 3-A. This pre-peening inspection also did not identify any indications of cracking attributed to PWSCC for the WCGS RPV upper head. When applicable MRP-335 performance criteria are met, peening mitigation prevents initiation of PWSCC. The WCGS peening application met or exceeded the MRP-335 depth of compression requirements. In cases when a shallow preexisting flaw is located within a region of compressive residual plus operating stress, PWSCC growth of the pre-existing flaw would likely be arrested as documented in MRP-335 R3-A [3]. The deterministic and probabilistic analyses in MRP-335 (sections 5.2 and 5.3, respectively) also show that it is not necessary for growth of shallow pre-existing flaws to be arrested by the post-peening stress field.

Pre-existing flaws are effectively addressed by the combination of pre-peening and follow-up inspections. The likelihood that a pre-existing flaw exists below the depth of peening application is low since there have been no discoveries of PWSCC for the WCGS upper head.

2. DEI report, TN-4069-00-01 [5], is based on information included in MRP-335 R3-A [3].

The DEI report references [3] matrices of deterministic PWSCC crack growth calculations to demonstrate how the timing of volumetric examinations of the Alloy 600 nozzles after peening is sufficient to prevent pressure boundary leakage. The various cases evaluated how hypothetical, shallow PWSCC flaws of three different initial depths that are located in Tcold head Alloy 600 nozzle base metal and exist at the time of peening would grow after application of peening. The postulated shallow flaws are < 10% through-wall (TW),

therefore the pre-peening / baseline UT inspection cannot reliably detect the postulated flaws. The evaluation contained in [5] was performed by DEI for Exelons Byron and Braidwood reactor vessel heads. However, the results are directly applicable to the Wolf Creek Generating Station (WCGS) reactor vessel head since the MRP-335 R3-A [3]

results are not plant specific and the reactor vessel heads in use at WCGS, Byron, and Braidwood all operate per MRP-48 [7] within the Tcold temperature range (547 - 561°F (degree Fahrenheit)) evaluated by MRP-335 R3-A [3] and have an 18-month nominal fuel cycle. Table 5-13 through Table 5-15 in MRP-335 R3-A [3] provide the time to grow to 10% TW and time to grow from 10% TW to a leak. The tables also identify when the flaw would be detected in the Alloy 600 nozzle and show that the likelihood of flaw detection and through-wall leakage are the same irrespective of examinations being performed during the N+2 or the N+3 refueling outage. The result (i.e., detection or leakage) when performing an N+3 follow-up examination is the same as the result when performing an

Attachment to ET 23-0005 Page 6 of 9 N+2 follow-up examination with respect to the ability to detect and mitigate a crack prior to the crack becoming a through-wall leak.

3. The additional 18 months for an N+3 follow-up inspection at WCNOC has the advantage of allowing more time for potential shallow pre-existing flaws to grow and become more readily detectable at the time of the follow-up inspection. Considering that ultrasonic (UT) examination is not qualified to detect shallow flaws extending less than 10% through the nozzle wall, the N+3 follow-up inspection would permit more time for the potential slow-growing flaws to exceed the 10% threshold thereby making the UT examination more effective in detecting the flaws prior to implementing the long-term 10-year inspection interval.

4. In accordance with MRP-335 R3-A [3] WCNOC will perform a bare metal visual examination of each RPVHPN for evidence of pressure boundary leakage every refueling outage. This requirement ensures that in the unlikely event that through-wall cracking occurs before the alternative N+3 follow-up volumetric examination that the through-wall crack is identified after minimal time. This reduces the time for propagation of a circumferential crack in the nozzle tube at the top of the J-groove weld. The early detection also allows minimal time for conditions to develop producing low alloy steel corrosion due to the concentration of boric acid. Industry experience [6] with a leaking CRDM penetration affected by cracking of the J-groove weld illustrated the sensitivity of the demonstrated leak path assessment examination as an early indication of leakage.

5. The deterministic crack growth rate calculations in MRP-335 R3-A [3] evaluated the maximum allowed total (residual plus normal operating) stress of +10ksi on the nozzle OD, ID, and weld. The results show that flaws significantly deeper than the reduced-stress region below the peened surface tend to grow in depth at a rate similar to that for the unmitigated case.

6. The experience for unmitigated heads in the U.S. operating at Tcold, including that for WCGS prior to peening, shows that in practice and without taking credit for the peening surface stress improvement, through-wall cracking and leakage are unlikely to occur prior to an alternative N+3 refueling outage follow-up inspection.

7. ASME Code Case N-729-6 Item B4.20 [2] provides inspection requirements for PWR Reactor Vessel Upper Heads with Alloy 600 nozzles and Alloy 82/182 partial penetration welds that have not been mitigated by peening. It bases the frequency of inspection, in part, on calculated EDY [effective degradation years] and RIY [reinspection years] of the head. Each of these parameters is a function of the time and temperature history of the head. Susceptibility to crack initiation is represented by the EDY parameter, and potential for crack propagation is represented by the RIY parameter. WCNOC performed a calculation of RIY and EDY for an unmitigated head using the methods outlined in Code Case N-729-6. WCNOCs EDY calculation for the RPV upper head, which operates at Reactor Cold-leg Temperature (Tcold), produced a result below 8 EDY. WCNOCs RIY

Attachment to ET 23-0005 Page 7 of 9 calculation concluded that the next volumetric examination would not be required until the fall of 2028 (N + 5). The reinspection frequency for unmitigated heads is conservative to apply for peened heads, as peening reduces the possibility of PWSCC occurrence. An interval of 4.5 calendar years exam at N+3 (Fall of 2025), is still less than the RIY of 2.25 from Code Case N-729-6 for an unmitigated RPV head with no indications of cracking attributed to PWSCC.

8. The above assessments and supplemental evaluation demonstrate that an N+3 refueling outage follow-up inspection maintains the same level of safety as an N+2 refueling outage follow-up inspection.

9. On this basis, a 54-month (N+3) follow-up inspection for the RPVHPNs identified in Section 1.0 provides reasonable assurance of the low likelihood of leakage in RPVHPNs to support use of the requested alternative timing of the follow-up inspection.

5.3 Maintaining Defense in Depth

1) Under the proposed alternative, the required bare metal visual examination would still be performed each RFO. This examination for evidence of pressure boundary leakage would provide defense in depth in the unlikely event that leakage was to occur due to base metal or J-groove weld cracking or due to small flaws that are too shallow to be reliably detected in the pre-peening examination. The visual examination frequency ensures timely identification of through-wall cracking, before the development of conditions that may lead to substantial boric acid corrosion of the low-alloy steel RPV upper head.

2) The WCNOC boric acid corrosion program uses visual inspections to detect the boric acid leakage source, path, and any targets of the leakage. This program ensures that boric acid corrosion is consistently identified, documented, evaluated, trended, and effectively repaired. The boric acid corrosion control program provides both detection and analysis of leakage of borated water inside containment.

3) Furthermore, WCGS Technical Specification (TS) 3.4.13 requires reactor coolant system operational leakage monitoring, which includes containment sump monitoring and containment atmosphere radioactivity monitoring. Containment sump monitoring and containment atmosphere radioactivity monitoring devices are required to be operable per TS 3.4.15.

5.4 Conclusions

1. The following assessments and supplemental evaluations provide reasonable assurance of the low likelihood of leakage affecting the RPVHPNs at WCGS under the proposed alternative:

• The deterministic crack growth results presented within Section 5.2.3.2 of MRP-335 R3-A [3] demonstrate how the N+3 refueling outage follow-up inspection timing is as

Attachment to ET 23-0005 Page 8 of 9 effective as the N+2 timing in the case of heads operating at Tcold with a nominal 18-month fuel cycle to prevent through-wall cracking and pressure boundary leakage.

• The additional cycle for an N+3 refueling outage inspection has the advantage of allowing more time for potential slow-growing flaws to become more readily detectable during the follow-up inspection when compared to the N+2 refueling outage.

• Without taking credit for the application of peening surface stress improvement, the experience for unmitigated heads in the U.S. operating at Tcold demonstrates how through-wall cracking and leakage are unlikely to occur during an alternative N+3 refueling outage inspection.

• Bare metal visual examinations for evidence of leakage are required and will be performed every refueling outage. The demonstrated leak path assessment examinations required whenever a volumetric examination is performed provide defense-in-depth to identify leakage through both the J- groove weld and nozzle base metal.

2. Therefore, WCGS concludes that performing RPVHPN volumetric examinations during the second and third RFOs after peening would present a hardship without commensurate safety benefit and performing a single RPVHPN volumetric examination during the third RFO after peening would provide an acceptable alternative.

6.0 DURATION OF PROPOSED ALTERNATIVE The proposed alternative will be utilized to perform the follow-up inspection in WCNOC Refueling Outage 27 (Fall 2025) instead of Refueling Outage 26 (Spring 2024). WCNOC will assume the inspection schedule required by MRP-335 R3-A [3] with the conditions of 10 CFR 50.55a following Refueling Outage 27.

7.0 PRECEDENTS

1. ML19155A060, Safety Evaluation, Braidwood Station, Unit 1 - Relief from the Requirements of the American Society of Mechanical Engineers Code (EPID L- 2018-LLR-0126)

a. ML18270A066, Request for Alternative Follow-up Inspection for Reactor Pressure Vessel Head Penetration Nozzles with Mitigated Alloy 600/82/182 Peened Surfaces in Accordance with 10 CFR 50.55a(z)(2) for Braidwood Station, Unit 1

2. ML19035A294, Safety Evaluation, Byron Station, Unit No 2, Relief from the Requirements of the ASME Code (EPID L 2018-LLR-0118)

a. ML18248A060, Request for Alternative Follow-up Inspection for Reactor Pressure Vessel Head Penetration Nozzles with Mitigated Alloy 600/82/182 Peened Surfaces in Accordance with 10 CFR 50.55a(z)(2) {Byron Station Unit 2}

Attachment to ET 23-0005 Page 9 of 9

8.0 REFERENCES

1. ASME Boiler and Pressure Vessel Code,Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components," 2007 Edition including Addenda through 2008

2. ASME Code Case N-729-6, "Alternative Examination Requirements for PWR Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial- Penetration Welds,Section XI, Division 1"

3. Materials Reliability Program: Topical Report for Primary Water Stress Corrosion Cracking Mitigation by Surface Stress Improvement (MRP-335, Revision 3-A), EPRI, Palo Alto, CA, 2016. 3002009241

4. Letter from K. Hsueh (U.S. Nuclear Regulatory Commission) to M. Sunseri (EPRI), "Final Safety Evaluation of the Electric Power Research Institute MRP-335, Revision 3,

'Materials Reliability Program: Topical Report for Primary Water Stress Corrosion Cracking Mitigation by Surface Stress Improvement [Peening]' (TAC No. MF2429),"

dated August 24, 2016 [NRC ADAMS Accession No. ML16208A485]

5. Technical Note TN-4069-00-01, Revision 0, "MRP-335 R3-A Matrix of Deterministic Crack Growth Calculations for Tcold Reactor Vessel Top Head Nozzles Evaluated for Alternative Peening Follow-up Volumetric Examination Timing," Dominion Engineering, Inc., Reston VA, August 2018

6. Letter from A. J. Vitale (Entergy) to U.S. Nuclear Regulatory Commission, "Licensee Event Report # 2018-001-00, Penetration Indications Discovered During Reactor Pressure Vessel Head Inspection, Indian Point Unit No. 2," dated May 21, 2018 [NRC ADAMS Accession No. ML18149A126]

7. Materials Reliability Program: PWR Materials Reliability Program Response to NRC Bulletin 2001-01 (MRP-48), EPRI, Palo Alto, CA, 2001), EPRI, Palo Alto, CA, 2016.

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