RS-20-080, Request for Alternative: One-Time Deferral of 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)

Request for Alternative: One-Time Deferral of 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)

ML20199M304
Person / Time
Site:	Byron
Issue date:	07/17/2020
From:	Demetrius Murray Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RS-20-080
Download: ML20199M304 (33)

Text

4300 Winfield Road Warrenville , IL 60555 Exelon Generation 630 657 2000 Office RS-20-080 10 CFR 50.55a July 17, 2020 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Byron Station, Unit 2 Renewed Facility Operating License No. NPF-66 NRC Docket No. STN 50-455

Subject:

Request for Alternative: One-Time Deferral of 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)

In accordance with 10 CFR 50.55a, "Codes and standards," paragraph (z)(2), Exelon Generation Company, LLC (EGC), is requesting relief for Byron Station (Byron) Unit 2 from the current examination requirements of Reactor Pressure Vessel Head Penetration Nozzles (RPVHPNs) performed in accordance with 10 CFR 50.55a(g)(6)(ii)(D), which specifies the use of American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components," Code Case N-729-6 on the basis that compliance with these requirements would result in hardship without a compensating increase in the level of quality and safety.

Circumstances are present in that the Centers for Disease Control and Prevention (CDC) and the State of Illinois have issued recommendations advising isolation activities (e.g., social distancing, group size limitations, self-quarantining, etc.) to prevent the spread of the COVID-19 virus. The nature of the RPVHPN inspections prevents compliance with the CDC and State of Illinois guidance for social distancing by maintaining at least six feet from other personnel to limit the spread of the virus. EGC does not have the internal capability and equipment to perform the volumetric examinations of the RPVHPNs. These volumetric examinations require a specialty vendor that maintains unique and complex qualifications. An outbreak affecting personnel and limiting their availability during RPVHPN inspections could result in not meeting the examination scope.

As discussed in Relief Request 14R-17 provided in Attachment 1, EGC proposes an alternative to the requirements of 10 CFR 50.55a(g)(6)(ii)(D) and is requesting relief to defer volumetric examination of the Byron Unit 2 RPVHPNs (Item No. B4.60 of Table 4-3 of MRP-335, Revision 3-A) to the next refueling outage (B2R23), which is scheduled in spring 2022. After this deferral, the approved volumetric examination frequency of once per inspection interval (nominally 10 calendar years) per MRP-335, Revision 3-A, Table 4-3, Item No. B4.60, will be followed.

U.S . Nuclear Regulatory Commission July 17, 2020 Page 2 Performing this examination in B2R22 is a hardship due to expected challenges with obtaining and maintaining staffing levels sufficient for the examination in B2R22. Deferral of this examination would also reduce the risk of exposure (i.e., reduced personnel on site and personnel physical distancing) for critical contract and direct hire personnel to the COVID-19 virus. The proposed alternative is based on crack growth analyses, assessments of the implications of previous indications of cracking that were repaired , direct visual examinations for evidence of leakage, and online leak detection capability.

Due to the hardship presented by the COVID-19 pandemic, EGC requests approval for a one-time deferral of the next required volumetric examination from the fall 2020 refueling outage (B2R22) to the following refueling outage in spring 2022 (B2R23). Performance of a volumetric examination of the Byron Unit 2 RPVHPNs during the fall 2020 (B2R22) refueling outage would result in a hardship without a compensating increase in the level of quality and safety in accordance with 10 CFR 50.55a(z)(2). includes the following supporting supplemental evaluation:

Technical Note TN-4069-00-02, Revision 1, "Experience for Unmitigated CROM Nozzles in U.S. PWRs Evaluated for Margin Against Leakage Considering Additional PWSCC Growth if Indications Had Remained in Service," Dominion Engineering , Inc., dated July 2020 EGC requests approval of the proposed alternative by September 15, 2020, which will support the Byron Unit 2 fall 2020 refueling outage (B2R22). The requested approval date is based on ensuring adequate vendor support for B2R22 scope.

There are no regulatory commitments in this submittal. Should you have any questions concerning this letter, please contact Ms. Lisa A. Simpson at (630) 657-2815.

D~:i:-

Sr. Manager - Licensing Exelon Generation Company, LLC Attachments:

1) 10 CFR 50.55a Relief Request 14R-17 for Byron Station, Unit 2, Revision 0

2) Dominion Engineering, Inc. Technical Note TN-4069-00-02 , Revision 1, July 2020 cc: Regional Administrator - NRC Region Ill NRC Senior Resident Inspector - Byron Station NRC Project Manager - Byron Station Illinois Emergency Management Agency - Division of Nuclear Safety

ATTACHMENT 1 10 CFR 50.55a Relief Request 14R-17 for Byron Station, Unit 2 Revision 0

ISi Program Plan Unit 2, Fourth Interval 10 CFR 50.55a Relief Request 14R-17 for Byron Station, Unit 2 Revision 0 (Page 1 of 20)

Request for Relief One-Time Deferral of 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)

1. ASME Code Components Affected Component Numbers: Byron Station (Byron) Unit 2, Reactor Vessel 2RC01 R

Description:

One-Time Deferral of Post Peening Follow-Up Inspection for Reactor Pressure Vessel Head Penetration Nozzles (RPVHPNs) Having Pressure-Retaining Partial-Penetration J-groove Welds with Mitigated Alloy 600/82/182 Peened Surfaces Code Class: Class 1 Examination Category: ASME Code Case N-729-6 Code Item: B4.20 Identification: RPVHPN Numbers 1 through 78 and the vent penetration nozzle (P-1 through P-78 and the vent penetration nozzle)

Reference Drawing: Closure Head Assembly: 185282E Size: 4 Inch Nominal Outside Diameter, 2.75 Inch Nominal Inside Diameter (Vent Nozzle NPS 1)

Material: SB-167 UNS N06600 (Alloy 600), ENiCrFe-3 (Alloy 182),

and ERNiCr-3 (Alloy 82)

2. Applicable Code Edition and Addenda lnservice Inspection (ISi) and Repair/Replacement Programs: American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code,Section XI, 2007 Edition including 2008 Addenda [1 ]. Examinations of the RPVHPNs are performed in accordance with Title 10 of the Code of Federal Regulations ( 10 CFR) 50.55a(g)(6)(ii)(D), which specifies the use of ASME Code Case N-729-6, with conditions.

Code of Construction [Reactor Pressure Vessel (RPV)]: ASME Section Ill, 1971 Edition through Summer 1973 Addenda.

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3. Applicable Code Requirement 3.1. General Requirements ASME Code Case N-729-6 contains requirements for the inspection of RPVHPNs, with or without flaws, as conditioned by 10 CFR 50.55a(g)(6)(ii)(D). The specific Code requirements for which use of the proposed alternative is being requested are as follows:

10 CFR 50.55a(g)(6)(ii)(D)(1) requires (in part):

"Holders of operating licenses or combined licenses for pressurized-water reactors as of or after June 3, 2020 shall implement the requirements of ASME BPV Code Case N-729-6 instead of ASME BPV Code Case N-729-4, subject to the conditions specified in paragraphs (g)(6)(ii)(D)(2) through ( 8) of this section, by no later than one year after June 3, 2020."

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" [2], Figure 2, "Examination Volume for Nozzle Base Metal and Examination Area for Weld and Nozzle Base Metal," is applicable to the RPVHPNs.

ASME Code Case N-729-6, Paragraph -2410 specifies that the reactor vessel head penetration welds mitigated by a peening mitigation technique shall be examined on a frequency in accordance with Table 1, Items 84.50 and 84.60, of the Code Case (Refer to [2], hereafter known as N-729-6).

10 CFR 50.55a(g)(6)(ii)(D)( 5) requires:

"In lieu of inspection requirements of Table 1, Items 84.50 and 84.60, 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 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 [3], Table 4-3, provides inspection requirements for Alloy 600 RPVHPNs mitigated by peening. In accordance with 10 CFR 50.55a(g)(6)(ii)(D)(5), the NRC condition 5.4 requirement for inspection in the first refueling outage post peening application (N+1 ), which is reflected in MRP-335, Revision 3-A, Table 4-3, Note (11 )(b),

is no longer required. Thus, only a follow-up examination in the second refueling outage post peening application (N+2) is required for RPVHPNs that have experienced less than 8 effective degradation years (EDYs), regardless if indications attributed to primary water stress corrosion cracking (PWSCC) have been previously found .

Previous relief has been granted to align the follow-up inspection for the 70 nozzles peened in spring 2016 with the follow-up inspection for the 9 nozzles peened in fall 2017

[4]. The current authorized follow-up inspection for all 79 identified nozzles is in the fall 2020 refueling outage. This corresponds to a nominal 4.5 calendar years since the last

ISi Program Plan Unit 2, Fourth Interval 10 CFR 50.55a Relief Request 14R-17 for Byron Station, Unit 2 Revision 0 (Page 3 of 20) volumetric examination for the 70 nozzles peened in spring 2016 (N+3 follow-up) and a nominal 3.0 calendar years since the last volumetric examination for the 9 nozzles peened in fall 2017 (N+2 follow-up).

The inspection frequency requirements for Item 84.60 for RPVHPNs mitigated by peening surface stress improvement (SSI) per MRP-335, Revision 3-A, Table 4-3 [3],

require a pre-peening baseline inspection, follow-up inspection, and subsequent in-service inspection . The Exelon Generation Company, LLC (EGC) proposed alternative under this relief only applies to the follow-up inspection for the 79 identified nozzles (Section 1) currently scheduled for the fall 2020 (B2R22) refueling outage, where a volumetric examination of 100% of the required volume or equivalent surfaces of the nozzle tube and a leak path examination are to be performed.

The proposed alternative would not modify the requirement of 10 CFR 50.55a(g)(6)(ii)(D)(5) and MRP-335, Revision 3-A, Table 4-3, Item 84.50 [3] that a bare metal visual examination (VE) of all identified nozzles be performed each refueling outage. Under the proposed alternative, a VE will be performed of all identified nozzles during the fall 2020 (B2R22) refueling outage.

3.2. Requirements Specific to Previously Repaired Penetrations In accordance with the NRC Safety Evaluation [5] addressing the embedded flaw repair method as applied to the Byron Unit 2 head, periodic ultrasonic testing (UT) and surface examinations have been performed of the previously repaired penetrations 6 and 68.

The long-term ISi requirement for the repaired nozzles is to perform UT examinations in accordance with 10 CFR 50.55a(g)(6)(ii)(D) supplemented by surface dye penetrant testing (PT) performed every other refueling outage. No indications of cracking or other degradation were detected during these examinations. Under the proposed alternative, PT would still be performed of the two repaired penetrations during the fall 2020 (B2R22) refueling outage in accordance with the schedule of the NRC Safety Evaluation [5]. In the same manner as for the unrepaired penetrations, the UT examination of repaired penetrations 6 and 68 would be deferred to the spring 2022 (B2R23) refueling outage.

Note (8) of both ASME Code Case N-729-6, Table 1, and MRP-335, Revision 3-A, Table 4-3 [3], require that "repaired areas shall be examined during the next refueling outage following the repair." As shown in Table 2 of this request, this requirement for a volumetric or surface examination of the repaired penetrations in accordance with Note (6) of N-729-6, Table 1 was satisfied prior to when peening was performed .

4. Reason for Request On January 27, 2020, it was determined that a Public Health Emergency (PHE) exists nationwide. This PHE was renewed on April 21, 2020. On March 9, 2020, Illinois declared a disaster proclamation over the Coronavirus outbreak. On March 13, 2020, the President of the United States declared a National Emergency due to the spread and infectious nature of the Coronavirus-2019 (COVID-19) virus and resulting pandemic.

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The most recent guidance from the Centers for Disease Control and Prevention (CDC) includes recommendations for physical distancing by maintaining at least six feet from other personnel to limit the spread of the virus. Furthermore, on March 28, 2020, the Department of Homeland Security identified workers in the nuclear energy sector as essential critical infrastructure workers .

On May 5, 2020, the Governor of Illinois announced a five-phase plan to reopen the State of Illinois, using health statistics and health care capacity. As of June 26, 2020, Ogle County, the county in which Byron is located and the North-Central Region of Illinois are currently in Phase 4 of the plan . The current limit on gatherings is 50 people.

Face coverings and social distancing are required. To reach the fifth and final phase, a vaccine, highly effective treatment, or herd immunity would need to be established . This hardship is considered to apply until the fifth and final phase is reached .

EGC is adhering to State and CDC recommendations to keep employees safe. To prevent the spread of COVI D-19 at Byron and to protect the health and safety of plant personnel while maintaining responsibilities to support critical infrastructure, EGC intends to reduce the amount of personnel on-site. A line by line review of outage activities has been performed, and all items that are not necessary to be performed for nuclear safety or reliable generation have been deferred in order to minimize the number of people onsite. Several other measures being taken by EGC include temperature screening and self-screening questions of personnel prior to accessing the site, as well as select personnel at Byron being remote work enabled based on their work functions.

EGC does not have the internal capability and equipment to perform the volumetric examinations of the reactor vessel closure head penetrations. These examinations must be performed by one of the two qualified vendors in the U.S. The potential remains for the inadvertent introduction and spread of the COVID-19 virus by vendor nondestructive examination (NOE) personnel who recently traveled from other areas of the country with substantial numbers of active COVID-19 cases. The CDC has determined that COVID-19 poses a serious public health risk, with many U.S. states reporting community spread of COVID-19.

Twenty-four individuals from across the U.S. are required to mobilize on-site to support UT examination of the Byron Unit 2 RPVHPNs. Appropriate precautions will be taken based on lessons learned (e.g ., individuals traveling alone, driving instead of flying , and staying in individual hotel rooms). The examination vendor has formalized these precautions via company policy as a mitigation measure for COVID-19. Discussions with the vendor yielded no additional or new technology to conduct the RPVHPN inspections to reduce COVID-19 risk.

Industry vendors are also supporting multiple overlapping outages and thus do not have large numbers of additional support personnel on standby. The industry vendors also maintain unique and complex qualifications. An outbreak affecting personnel and limiting their availability during reactor vessel closure head inspections could result in not meeting the examination scope.

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The nature of the work prevents meeting CDC recommendations for social distancing by maintaining at least six feet from other personnel to limit the spread of the virus.

Although a majority of the work involved can be performed while maintaining social distancing, social distancing cannot be maintained for some tasks involved. Mitigation strategies do not eliminate the time that social distancing is not maintained. These individuals will wear the approved personal protective equipment (PPE) for COVID-19 in accordance with EGC policy. EGC believes that additional PPE has the potential to add individual effort to activities that increase human performance and safety risk.

EGC has assessed the possibility of some of the personnel who normally support the examination on-site with providing support remotely by data transmittal instead. The savings in personnel from this approach would only be expected to reduce the number of personnel traveling to the site by 3-4 employees (of the 24 individuals required to mobilize). This benefit is offset by the risk of slow data transmittal or interruptions, which would prolong the amount of time personnel need to stay on site and delay the completion of the examination.

EGC is contingency planning in case some of its workforce becomes unavailable due to the COVID-19 outbreak. With the current work scope and potential loss of personnel, there is the potential that the company may not be able to complete the refueling outage in a timely manner, which could negatively impact critical infrastructure that is needed during this time .

EGC has implemented the Ultra High Pressure Cavitation Peening (UHPCP) process at Byron Unit 2. EGC is requesting relief from the requirements of 10 CFR 50.55a(g)(6)(ii)(D) for a one-time alternative in which the volumetric examination for the nozzles identified in Section 1 is deferred by one fuel cycle to spring 2022. This alternative corresponds to a nominal 6.0 calendar years since the last volumetric examination for the 70 nozzles peened in spring 2016 (N+4 follow-up) and a nominal 4.5 calendar years since the last volumetric examination for the 9 nozzles peened in fall 2017 (N+3 follow-up). An interval of 6.0 calendar years is still less than the interval of 8 calendar years that would be acceptable under the requirements of 10 CFR 50.55a(g)(6)(ii)(D) for the Byron Unit 2 head if PWSCC affecting the head nozzles was not previously detected.

During the initial peening application in spring 2016 (B2R19), 9 RPVHPNs at Byron Unit 2 did not receive complete peening coverage to meet the performance criteria of MRP-335, Revision 3-A [3]. The specific affected nozzles are identified in a response to NRC Request for Additional Information submitted by EGC on July 14, 2017 [6]. Eight control rod drive mechanism (CROM) penetrations and the vent line penetration were affected (9 total penetrations). EGC successfully peened the inner diameter (ID), outer diameter (OD), or both surfaces of these 9 penetrations (as needed) during the fall 2017 refueling outage (B2R20) to satisfy MRP-335, Revision 3-A requirements. Therefore, all 79 identified nozzles at Byron Unit 2 are now successfully peened in accordance with MRP-335, Revision 3-A.

This request is submitted due to an expected hardship obtaining and maintaining onsite staffing levels sufficient to prepare, perform, and demobilize from the volumetric examination of the RPVHPNs. In addition, the proposed alternative would reduce the

ISi Program Plan Unit 2, Fourth Interval 10 CFR 50.55a Relief Request 14R-17 for Byron Station, Unit 2 Revision 0 (Page 6 of 20) number of outside personnel introduced to the site during the refueling outage, which would have the benefit of reducing the possibility that an asymptomatic person infected with COVID-19 is present. Furthermore, the proposed alternative would reduce the extent of activities in containment during the refueling outage, reducing the workload and physical concentration of plant and other personnel. Therefore, deferral of the volumetric examination would have multiple benefits to reduce the risk of the COVID-19 outbreak affecting other maintenance activities during the refueling outage.

In summary, performance of a volumetric examination of the Byron Unit 2 RPVHPNs during the fall 2020 (B2R22) refueling outage is considered a hardship without a compensating increase in the level of quality and safety in accordance with 10 CFR 50.55a(z)(2) based on the assessments and supplemental evaluations described below.

5. Proposed Alternative and Basis for Use 5.1. Proposed Alternative EGC is requesting relief to defer volumetric examination of the Byron Unit 2 RPVHPNs (Item No. B4.60 of Table 4-3 of MRP-335, Revision 3-A) to the next refueling outage (B2R23), which is scheduled in spring 2022. After this deferral, the approved volumetric examination frequency of once per inspection interval (nominally 10 calendar years) per MRP-335, Revision 3-A, Table 4-3, Item No. B4.60, will be followed. Byron Unit 2 operating cycles are approximately 18 months in duration.

Performing this examination in B2R22 is a hardship due to expected challenges with obtaining and maintaining staffing levels sufficient for the examination in B2R22.

Deferral of this examination would also reduce the risk of exposure (reduced personnel on site and personnel physical distancing) for critical contract and direct hire personnel to the COVID-19 virus. The proposed alternative is based on crack growth analyses, assessments of the implications of previous indications of cracking that were repaired, direct VEs for evidence of leakage, and online leak detection capability. Based on these factors, EGC has identified performance of the volumetric examinations of the Byron Unit 2 RPVHPNs in B2R22 as a hardship without a compensating increase in the level of quality and safety in accordance with 10 CFR 50.55a(z)(2).

5.2. Background The primary degradation mechanism addressed by 10 CFR 50.55a(g)(6)(ii)(D) and Code Case N-729-6 is PWSCC. This degradation mechanism occurs when a susceptible material is exposed to a primary water environment, elevated tensile stress levels, and elevated operating temperatures. In the case of RPVHPNs, the periodic examinations required by Code Case N-729-6 are intended to provide reasonable assurance that any potentially existing PWSCC will be identified before it can cause significant degradation through the ejection of a penetration nozzle or significant degradation of the low-alloy steel head through leakage and consequent boric acid corrosion. Any leakage that could result due to a through-wall PWSCC condition would be small due to the morphology of PWSCC type flaws and not directly challenge the safety of the plant.

ISi Program Plan Unit 2, Fourth Interval 10 CFR 50.55a Relief Request 14R-17 for Byron Station, Unit 2 Revision 0 (Page 7 of 20)

Operating experience demonstrates the susceptibility of Alloy 600 nozzles to PWSCC.

Indications of PWSCC have been detected during previous examinations affecting two of the RPVHPNs at Byron Unit 2 (References [7] and [8]). Both affected CROM nozzles were repaired using the embedded flaw repair technique. Table 1 shows size and location information for each of the repaired PWSCC indications based on the data summary sheets for the UT examinations performed (References [9] and [1 O]). Both indications were located on the nozzle OD, with none of the indications originating on the nozzle ID . The first of the two indications, detected in 2007, was attributed to a lack-ef-fusion weld defect that ultimately allowed for initiation of PWSCC [7]. The second indication, detected in 2014, was attributed to PWSCC [8]. A summary of nondestructive examinations performed for the Byron Unit 2 RPVHPNs is provided in Table 2, which also shows the future schedule of examinations under the proposed alternative.

Both indications were located a substantial distance below the nozzle OD annulus, showing significant margin against a through-wall condition and pressure boundary leakage. The vertical ligament between the upper end of the indication and the bottom of the nozzle annulus (i.e., triple point) was at least 0.96 inch. Further growth in the axial (vertical) direction of at least 0.96 inches would be required for a through-wall condition and leakage to result.

Cracking within the Alloy 82/182 J-groove weld can lead to leakage without cracking within the nozzle base metal that would be detectable in the volumetric examination required by 10 CFR 50.55a(g)(6)(ii)(D) regardless of the frequency of that examination.

These volumetric examinations are intended to detect PWSCC affecting the nozzle base metal, with a leak path assessment also included to detect leakage due to a through-wall condition through the weld metal and/or base metal. The UT examination has the capability to detect PWSCC within the nozzle base metal prior to growth to the nozzle OD annulus and leakage occurring. The combination of the leak path assessment and the direct VE of the exterior of the head for evidence of leakage addresses the concern for boric acid corrosion due to PWSCC affecting the weld metal.

5.3. PWSCC Crack Growth Evaluation - Flaws at Limit of UT Detectability A PWSCC crack growth analysis specific to the Byron Unit 2 RPVHPNs, which was previously submitted to NRC in 2007 [11], determined the amount of time needed for a postulated flaw to result in a potential leak path. In the analysis, growth of axial flaws on both the nozzle OD and ID was simulated. The initial flaw depth in each case was taken as 0.075 inch (12% through the nominal nozzle thickness of 0.625 inch). This flaw depth, which is slightly larger than the minimum depth of flaws included in the demonstration specimen sets for qualification of UT volumetric examinations per Code Case N-729-6 (10% through-wall), is a reasonable measure of the limit of flaw detectability. Each flaw was assumed to have a semi-elliptical shape. Growth was simulated for PWSCC using the standard PWSCC crack growth rate equation of MRP-55 [12], which has since been included within Nonmandatory Appendix C of ASME Section XI versions that are incorporated by reference within 10 CFR 50.55a. A capacity factor of 98% was conservatively assumed in the crack growth calculations.

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Weld residual stress analysis results specifically produced for the Byron Unit 2 head, including the effect of normal operating pressure and temperature, were applied. Crack growth results were produced for five RPVHPNs incidence angles (0°, 25.4°, 42.8°,

43.8°, and 4 7°), spanning the complete range from the center nozzle to the outermost set of nozzles, as well as for the stress profiles applicable to both the uphill and downhill sides of the nozzle. A normal operating temperature of 558°F was applied in the PWSCC crack growth rate equation. This temperature produces conservatively faster crack growth rates (i.e., a factor of 1.28 faster applying the thermal activation energy of the MRP-55 equation) than the current Byron Unit 2 nominal closure head temperature of 549°F [13].

Growth of OD axial flaws was postulated on both the uphill and downhill sides of the weld, with the flaw centered in the area of the nozzle material with the highest tensile stresses. An initial total-length-to-depth aspect ratio of 2: 1 (length of 0.15 inch) was assumed. As shown in Table 1, this aspect ratio is similar to that observed for the two actual reported indications affecting the Byron Unit 2 RPVHPNs. Development of the aspect ratio over time was calculated as growth was simulated at both the deepest and surface points of the semi-elliptical crack profile. The flaws were grown in length and depth until the upper tip of the crack reached the top of the J-groove weld, at which point a leak path would be established.

Growth of ID axial flaws was evaluated on the uphill side of the J-groove weld for a flaw located at the weld. Growth was calculated in the depth direction only, under the assumption of a constant length-to-depth aspect ratio of 6:1 (i.e., an initial length of 0.45 inch). For ID axial flaws, this is a conservative assumption resulting in faster crack growth in the depth direction (when compared to modeling shorter flaws and considering development of the aspect ratio over time) . Crack growth was evaluated until the flaw depth reached 100% through-wall, at which point the flaw reaches the nozzle annulus and causes leakage.

Results of these crack growth evaluations are shown in Table 3. These results show that the bounding time for an axial crack to grow from 12% through-wall to leakage is 6.9 calendar years. These calculations demonstrate that leakage due to base metal cracking is unlikely to occur under the proposed alternative. This calculation takes no credit for the peening mitigation performed for these nozzles.

5.4. PWSCC Crack Growth Evaluation - Previously Detected Flaws in T co1d Heads The experience for unmitigated heads in the U.S . operating at reactor cold-leg temperature (Tco1d), including that for Byron Unit 2 prior to peening, shows that in practice and without taking credit for the peening SSI, a through-wall cracking condition and leakage are unlikely to occur during the additional operating cycle prior to the alternative N+4 follow-up inspection. A 2016 PVP conference paper [14] evaluated in detail the PWSCC indications detected in 25 RPVHPNs in Tcold heads by that time, all in the area of the toe of the J-groove weld on the nozzle OD. All of these PWSCC indications, including those affecting Byron Unit 2, were detected in Tcold heads having nozzle material supplied by B&W Tubular Products (B&WTP).

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Through an extension of the assessment of plant experience in the PVP paper, the evaluation per TN-4069-00-02, Revision 1 [15], demonstrates that the likelihood of leakage due to PWSCC affecting the nozzle tube base metal remains low assuming an examination frequency of once every four 18-month cycles (i .e., N+4, or 6.0 calendar years). Only one of the 44 total indications detected in a U.S. Tcold head is modeled to result in leakage per the methodology applied to consider the implications of an examination every four 18-month cycles. For all but three of the 44 indications, a large margin against leakage (i.e. , remaining vertical distance from the crack front to the nozzle annulus is greater than the distance equivalent to the nozzle tube thickness) would have been expected if the volumetric examinations had been performed every four 18-month cycles (6.0 calendar years) in each case.

5.5. Previously Repaired Nozzles The two nozzles previously repaired using the embedded flaw repair technique are normally examined volumetrically when the other 76 CROM nozzles are volumetrically examined . Under the proposed alternative, the examinations of the two previously repaired nozzles will also be deferred until the next refueling outage. This one-time extension of one additional refueling cycle between examinations of the repaired nozzles is justified for the following reasons.

The "embedded flaw repair" for a flaw connected to the nozzle outer surface involves applying PWSCC-resistant weld metal (e.g., Alloy 52) over the OD of the Alloy 600 nozzle tube and the wetted surfaces of the J-groove weld, overlapping the vessel cladding and extending to the bottom of the nozzle, to isolate the susceptible material from primary coolant. Without contact with coolant, further PWSCC-induced growth of the embedded flaw is prevented. This repair is unlikely to significantly affect the stress state at the nozzle ID, and to the extent there is an effect on the stress at the ID, the squeezing of the nozzle tube by shrinkage of the weld overlay upon cooling would tend to reduce the magnitude of the tensile stress at the nozzle ID.

Periodic UT examination on the nozzle ID, as regularly performed at Byron Unit 2 per 10 CFR 50.55a(g)(6)(ii)(D), monitors the potential for growth of an embedded flaw originally located in the nozzle tube , or checks for growth of an embedded flaw originally located in the weld into the nozzle tube (e.g., References [16) and [17)). As shown in Table 2, the indication detected in 2007 (penetration 68) has been examined volumetrically in seven refueling outages since repair. The indication detected in 2014 (penetration 6) has been examined volumetrically in one refueling outage since repair.

Neither of these indications has shown evidence of growth subsequent to the embedded flaw repair. Furthermore, as shown in Table 2, PT examinations of the embedded flaw repairs in penetrations 6 and 68 are planned for the fall 2020 (B2R22) refueling outage.

With the embedded flaw repair, the only mechanism for sub-critical crack growth is fatigue [16). The allowable flaw size is determined based on linear elastic fracture mechanics and elastic plastic fracture mechanics, in accordance with methodologies established in ASME Section XI. Predicted fatigue crack growth is subtracted from the calculated allowable crack sizes to determine the maximum crack sizes for the embedded flaw repair. The flaws repaired in penetrations 6 and 68 were smaller than this maximum crack size for the embedded flaw repair.

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The embedded flaw repair technique has been applied in over 45 different instances throughout the world, and the flaw being repaired has never come into contact with water after repair [18]. These repairs have been in place up to 10 years in some cases.

The standard UT examination on the nozzle ID also addresses the potential for new flaws initiating on the nozzle ID in the same manner as for an unrepaired nozzle. Plant experience has shown the embedded flaw repair to be a reliable approach, as long as consideration is given to ensuring that the overlay covers all of the susceptible material.

The major attributes and learnings from the recent Palisades experience [19] have been implemented in the EGC oversight process and were in place prior to the Palisades experience. EGC NOE personnel along with other utilities were involved with the review of the Palisades investigation and helped formulate the lessons learned for that event.

EGC implemented an oversight process in B2R19 and B2R20 that included comparison of the UT examination data from the inspection outage to earlier outages, not just the last previous outage. EGC has purchased the software used for the UT examination analysis by vendors performing the examinations. EGC NOE personnel have also received training and have been qualified at EPRI for both vendor procedures. With the purchase of the vendor UT examination analysis software, examination data was independently analyzed by qualified EGC NOE personnel using the data from earlier outages and not only the last outage to validate no significant changes were identified.

5.6. Peening Follow-Up Volumetric Examinations at Other Units To date, in the U.S., peening of RPVHPNs has been performed at Byron Unit 2 (spring 2016 I fall 2017), Braidwood Unit 1 (fall 2016 I spring 2018), Braidwood Unit 2 (spring 2017), and Byron Unit 1 (spring 2017). The post-peening N+2 follow-up volumetric examination was performed at Byron Unit 1 in spring 2020, with no indications accepted for continued service and no welded repairs required (i.e., no relevant indications) [20],

thereby qualifying for the MRP-335, Revision 3-A, Table 4-3, Item. No. B4.60 examination frequency of once per inspection interval (nominally 10 calendar years).

Post-peening follow-up volumetric examinations have not yet been performed for the other units. These examinations are scheduled for spring 2021 at Braidwood Unit 1 and spring 2023 at Braidwood Unit 2.

The RPVHPNs at Braidwood Unit 1, Braidwood Unit 2, Byron Unit 1, and Byron Unit 2 were all fabricated by B&W using Alloy 600 material supplied by B&WTP. The RPVHPNs at all four units were also peened using the UHPCP process. Thus, the nozzles in all four units have similar histories, with similar head materials, nozzle manufacturers, and peening processes. Hence, the favorable experience with the Byron Unit 1 post-peening follow-up volumetric examination is relevant to the peened head at Byron Unit 2.

ISi Program Plan Unit 2, Fourth Interval 10 CFR 50.55a Relief Request 14R-17 for Byron Station, Unit 2 Revision 0 (Page 11 of 20) 5.7. Maintenance of Defense in Depth The crack growth assessments described above show that the one-time alternative nominal interval of up to 6.0 years for volumetric examinations provides reasonable assurance of leak-tightness of the RPVHPN Alloy 600 base metal. In addition, axial cracking within the nozzle does not represent a credible rupture concern as the critical axial crack length is much longer than the region of high tensile weld residual stress that drives PWSCC growth [18] . Furthermore, example deterministic crack growth calculations summarized in MRP-395 [18] show large margins in the time until circumferential cracking within the RPVHPN tube could produce a nozzle ejection versus the up to 6.0-year alternative interval. Finally, the single head vent nozzle is of reduced safety significance because of its small diameter relative to the other RPVHPNs.

In accordance with ASME Code Case N-729-6, Inspection Item No. B4.20, for an unmitigated head without previously detected PWSCC, the reexamination interval is based on the reinspection years parameter (RIY). Specifically, the interval between volumetric examinations is limited to RIY =2.25 or 8 calendar years, whichever is sooner. At the current Byron Unit 2 operating temperature of 549°F, the allowable interval for heads without previously detected PWSCC is limited by the 8 calendar years requirement (RIY = 2.09 conservatively assuming a 100% capacity factor). Hence, the proposed alternative interval of 6.0 years between volumetric examinations is still substantially more frequent than the inspection frequency required for unmitigated heads without previously detected PWSCC. Note (8) of Table 1 of Code Case N-729-6 conservatively requires that the volumetric examination interval for unmitigated heads that operate at reactor cold-leg temperature and have previously reported PWSCC be limited to two 18-month operating cycles. Nevertheless, the crack growth analyses specific to Byron Unit 2 discussed in Section 5.3 show that leakage is unlikely to occur due to base metal cracking under the proposed alternative.

Boric acid corrosion degradation of the low-alloy steel head material cannot occur as a result of PWSCC of a head penetration unless the PWSCC results in a through-wall condition (i.e., cracking from a normally wetted surface to the nozzle OD annulus).

Hence, the crack growth assessments in Section 5.3 show that the proposed one-time alternative volumetric examination interval of up to 6.0 years is effective to prevent boric acid corrosion as a result of PWSCC affecting the nozzle base metal material.

Under the proposed alternative, the direct VE of the Byron Unit 2 closure head required by MRP-335, Revision 3-A, Table 4-3, Item No. B4.50, would still be performed during each refueling outage including the B2R22 refueling outage in fall 2020. This sensitive VE for evidence of pressure boundary leakage would provide defense in depth in the unlikely case that leakage was to occur due to base metal cracking . (The VE also addresses the potential for cracking through the J-groove weld to cause leakage. The volumetric examination for which deferral is proposed does not have the capability to detect PWSCC within the J-groove weld prior to leakage occurring.) The periodic VEs ensure that through-wall cracking is identified in a timely fashion, with insufficient time for substantial circumferential cracking in the nozzle tube at the top of the weld to develop and insufficient time for conditions to develop producing substantial low-alloy steel corrosion due to the concentration of boric acid.

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Byron Unit 2 implements a boric acid corrosion control program consistent with WCAP-15988, Revision 2 [21 ]. As part of the boric acid corrosion control program, periodic walkdowns for boric acid leaks are regularly performed. These walkdowns include careful inspection for evidence of boric acid deposits and are an effective method for detecting small leaks below the sensitivity of other detection methods.

Additional system walkdowns are performed just before shutdown with the plant hot but at zero power (i.e., Mode 3), as well as during plant startup in accordance with 10 CFR 50.55a and ASME Section XI IWB-5000 (VT-2 visual examinations), with particular attention paid to pressure boundary components or connections that were opened or replaced during the outage.

As for unmitigated heads, the demonstrated leak path assessment examinations required whenever a volumetric examination is performed (including immediately prior to peening as part of the pre-peening baseline inspection), provide defense-in-depth to identify leakage through both the J-groove weld and nozzle base metal. Recent industry experience [22) with a leaking CROM 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.

Moreover, a leak or increase in radiation levels within containment would be captured in the containment sump and detected by radiation monitoring during operation if a substantial leak were to develop. If an unidentified reactor coolant system (RCS) leak is greater than 1 gpm, the plant Technical Specifications (TS) 3.4.13, "RCS Operational LEAKAGE," outlines the timely actions required to maintain safe operability for recovery, including a shutdown. These detection methods ensure that leakage would not go undetected for long periods of time. Byron Unit 2 also implements a 0.1 gpm action level on unidentified leakage, consistent with WCAP-16465-NP, Pressurized Water Reactor Owners Group Standard RCS Leakage Action Levels and Response Guidelines for Pressurized Water Reactors [23).

The RCS leakage quantity is reviewed against the TS associated with RCS leakage criteria. Depending on the source identified, a shutdown could be required in accordance with TS Limiting Condition for Operation (LCO) 3.4.13 that has the following specific limits:

a. No pressure boundary LEAKAGE,

b. 1 gpm unidentified LEAKAGE,

c. 10 gpm identified LEAKAGE, and

d. 150 gallons per day primary to secondary LEAKAGE through any one steam generator (SG).

A through-wall leak from a reactor vessel closure head penetration nozzle or weld would constitute pressure boundary leakage. Should any of these limitations be exceeded the appropriate LCO condition would be entered, and the required actions performed within the specified completion time, including plant shutdown, if required.

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In summary, continued performance of the sensitive VE each refueling outage and online leak detection capabilities maintain defense in depth . Under the proposed alternative, reasonable assurance of structural integrity is provided.

5.8. Conclusion In summary, under the proposed alternative, the nuclear safety concern of nozzle ejection is addressed with large margins. In addition, the continued performance of sensitive direct VEs for evidence of pressure boundary leakage each refueling outage under the proposed alternative ensures that the nuclear safety concern for structurally significant boric acid corrosion of the low-alloy steel head material is addressed .

Furthermore, crack growth analyses demonstrate that leakage is unlikely to occur due to base metal cracking under the proposed alternative. Included are both a crack growth analysis specific to the Byron Unit 2 head for hypothetical initial flaws at the limit of detectability as well as extrapolation of growth of as-found indications in U.S. heads operating at Tcold* Conservatively, these analyses do not credit the effectiveness of the peening mitigation that has already been performed of the Byron Unit 2 head. In the unlikely case that leakage was to occur under the proposed alternative that would not occur if the alternative was not implemented, defense in depth is maintained by the direct VEs for evidence of leakage and the online leak detection capability. The proposed alternative provides reasonable assurance of structural integrity until the next volumetric examination is performed in spring 2022 (B2R23).

Due to the hardship presented by the COVID-19 pandemic, EGC requests approval for a one-time deferral of the next required volumetric examination from the fall 2020 refueling outage (B2R22) to the following refueling outage in spring 2022 (B2R23). Performance of a volumetric examination of the Byron Unit 2 reactor vessel closure head penetrations during the fall 2020 (B2R22) refueling outage would result in a hardship without a compensating increase in the level of quality and safety in accordance with 10 CFR 50.55a(z)(2).

EGC requests approval of the proposed alternative by September 15, 2020, which will support the Byron Unit 2 fall 2020 refueling outage (B2R22). The requested approval date is based on ensuring adequate vendor support for B2R22 scope.

6. Duration of Proposed Alternative The proposed alternative is requested for the duration of the next Byron Unit 2 operating cycle, until the spring 2022 refueling outage (B2R23). After performing the volumetric examination in B2R23, the approved frequency of in-service inspections per 10 CFR 50.55a(g)(6)(ii)(D) and MRP-335, Revision 3-A [3] would resume.

7. Precedent NRC has approved a previous alternative due to the hardship in mobilizing the inspection vendor during the COVID-19 pandemic and required augmented examinations of RPVHPNs per ASME Code Case N-729-4, as summarized in the table below.

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NRC ADAMS Accession No.

Plant Component Examination NRC Safety Relief Request Evaluation Reactor vessel hot leg N-770-2, Item A-2, nozzle dissimilar metal Volumetric welds Comanche N-722-1, Item 8MN ML201 OOG562 ML20106F235 Peak2 815.80 Visual N-729-4, Item 84 .10 RPVHPNs Visual

8. Acronyms ASME American Society of Mechanical Engineers BPV Boiler and Pressure Vessel BMN Bottom Mounted Nozzle B&WTP Babcock & Wilcox Tubular Products CDC Center for Disease Control and Prevention CFR Code of Federal Regulations COVID-19 Coronavirus Disease of 2019 CROM Control Rod Drive Mechanism EDY Effective Degradation Year EPRI Electric Power Research Institute ID Inner Diameter ISi lnservice Inspection LCO Limiting Condition for Operation MRP [EPRI] Materials Reliability Program NOE Nondestructive Examination NRC Nuclear Regulatory Commission OD Outer Diameter PHE Public Health Emergency PPE Personal Protective Equipment PT (Dye) Penetrant Testing PWR Pressurized Water Reactor PWSCC Primary Water Stress Corrosion Cracking RCS Reactor Coolant System RIY Reinspection years [parameter]

RPV Reactor Pressure Vessel RPVHPN Reactor pressure vessel [upper] head penetration nozzle RVON Reactor Vessel Outlet Nozzle

ISi Program Plan Unit 2, Fourth Interval 10 CFR 50.55a Relief Request 14R-17 for Byron Station, Unit 2 Revision 0 (Page 15 of 20)

SG Steam Generator SSI Surface Stress Improvement Tcold Reactor Cold-leg Temperature TS Technical Specification UH PCP Ultra High Pressure Cavitation Peening UT Ultrasonic Testing VE Visual Examination

9. References

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

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," Approval Date: March 3, 2016.

3. Letter from David Czufin (TVA) and Brian Burgos (EPRI) to U.S. Nuclear Regulatory Commission, "Transmittal of 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 ," dated November 8, 2016. (available at www.epri.com ) [NRC ADAMS Accession No. ML16319A282].

4. Letter from D. Wrona (U.S. Nuclear Regulatory Commission) to B. Hanson (Exelon Generation Company, LLC), "Byron Station, Unit No 2, Relief from the Requirements of the ASME Code (EPID L-2018-LLR-0118)," dated February 25, 2019. Relief Request 14R-16 Regarding Alternative Follow-Up Inspections for Reactor Pressure Vessel Head Penetration Nozzles. [NRC ADAMS Accession No. ML19035A294].

5. Letter from K. J. Green (U .S. Nuclear Regulatory Commission) to B. Hanson (Exelon Generation Company, LLC), "Byron Station, Unit Nos. 1 and 2 - Request for Relief from the Requirements of the ASME Code (CAC Nos. MF8282 and MF8283)," dated March 6, 2017. Relief Request No. 14R-10, Revision 2, Regarding Reactor Vessel Head Penetrations. [NRC ADAMS Accession No. ML17062A428].

6. Letter from D. M. Gullatt (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "Response to Request for Additional Information for Byron Station and Braidwood Station Post Peening Relief Requests," dated July 14, 2017, including AREVA Licensing Report ANP 3601 NP Revision 0, "Response to Request for Additional Information for Byron Station Unit 2 and Braidwood Station Unit 1,"

dated July 2017 (Non-Proprietary). [NRC ADAMS Accession No. ML17200C952].

7. Letter from D. Hoots (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "Licensee Event Report (LER) 455-2007-001-00, Reactor Pressure Vessel Head Control Rod Drive Mechanism Penetration Nozzle Weld Indication Due to an Initial Construction Weld Defect Allows the Initiation of Primary Water Stress Corrosion Cracking," dated June 8, 2007. [NRC ADAMS Accession No. ML071590211].

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8. Letter from F. Kearney (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "Licensee Event Report (LER) 455-2014-004-00, Byron Unit 2 Reactor Pressure Vessel Head Control Rod Drive Mechanism Penetration Nozzle Weld Indication Attributed to Primary Water Stress Corrosion Cracking,"

dated December 5, 2014. [NRC ADAMS Accession No. ML14339A538].

9. Byron Unit 2, Penetration 68 Inspection Data Sheets:

a. CBE-R13-BP05-68-04, Ultrasonic Report Sheet, dated April 13, 2007 .

b. CBE-R13-BP01-68-01 , Ultrasonic Report Sheet, dated April 9, 2007.

c. CBE-R13-BP01-68-01 , Eddy Current Report Sheet, dated April 9, 2007.

d. Exelon Nuclear Liquid Penetrant Examination Data Sheet, Report Number 2007-276, dated April 10, 2007.

e. CBE-R12-2-068-R1, R2, R3, Ultrasonic Report Sheet, dated April 27, 2007.

f. CBE-R13-2-068-R1, Eddy Current Report Sheet, dated April 27, 2007.

g. "Review of CROM inspection Data from Byron Unit 2 Penetration #68," dated April 2007. KUW009 R1 Draft B.

h. "Review of CROM Inspection Data from Byron Unit 2 Penetration #68" dated April 2007. KUW009 R2 Issue 1-0.

i. Exelon Nuclear Liquid Penetrant Examination Data Sheet, Report Number 2007-315, dated April 23, 2007.

10. Byron Unit 2, Penetration 6 Inspection Data Sheets:

a. CBE-R18-CP02-06-03, 04, 05, Ultrasonic Report Data Sheet, dated October 17, 2014 .

b. CBE-R18-CP02-06-01, Ultrasonic Report Data Sheet, dated October 7, 2014.

11. Byron Unit 2 - Technical Basis for Reactor Pressure Vessel Head Inspection Relaxation, AM-2007-011 Revision 1, Exelon Nuclear, September 27, 2007.

[NRC ADAMS Accession No. ML091030445].

12. Materials Reliability Program (MRP) Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking (PWSCC) of Thick-Wal/ Alloy 600 Materials (MRP-55) Revision 1, EPRI, Palo Alto, CA: 2002. 1006695.

13. Exelon Engineering Change EC628148 Rev 000, "B2R21 Effective Degradation Years (EDY) Results and Effective Full Power Years Evaluation," dated August 7, 2019.

14. G. White, K. Fuhr, M. Burkardt, and C. Harrington, "Deterministic Technical Basis for Re-Examination Interval of Every Second Refueling Outage for PWR Reactor Vessel Heads Operating at Tcold with Previously Detected PWSCC," Proceedings of the ASME 2016 Pressure Vessels & Piping Conference, ASME, PVP2016-64032.

15. Technical Note TN-4069-00-02, Revision 1, "Experience for Unmitigated CROM Nozzles in U.S. PWRs Evaluated for Margin Against Leakage Considering Additional PWSCC Growth if Indications Had Remained in Service," Dominion Engineering, Inc., Reston, VA, July 2020.

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16. Letter from H. Berkov (U.S. Nuclear Regulatory Commission) to H. A. Sepp (Westinghouse Electric Company), "Acceptance for Referencing - Topical Report WCAP-15987-P, Revision 2, 'Technical Basis for the Embedded Flaw Process for Repair of Reactor Vessel Head Penetrations,"' dated July 3, 2003. [NRC ADAMS Accession No. ML031840237].

17. Letter from R. Pascarelli (U.S. Nuclear Regulatory Commission) to T. D. Gatlin (South Carolina Electric & Gas Company), "Virgil C. Summer Nuclear Station, Unit 1

- Alternative Request Weld Repair for Reactor Vessel Upper Head Penetrations (TAC NO. MF3546)," dated April 30, 2014. [NRC ADAMS Accession No. ML14107A332].

18. Materials Reliability Program: Reevaluation of Technical Basis for Inspection of Alloy 600 PWR Reactor Vessel Top Head Nozzles (MRP-395). EPRI, Palo Alto, CA:

2014. 3002003099. [NRC ADAMS Accession No. ML14307B007].

19. Letter from J. A. Hardy (Entergy) to U.S. Nuclear Regulatory Commission, "LER 2018-003 Indications Identified in Reactor Pressure Vessel Head Nozzle Penetrations, Palisades Nuclear Plant," dated January 3, 2019. [NRC ADAMS Accession No. ML19003A239]

20. Letter from H. Peterson (NRC) to B. C. Hanson (Exelon), "Byron Station -

Integrated Inspection Report 05000454/2020001 and 05000455/2020001," dated April 24, 2020. [NRC ADAMS Accession No. ML20115E528].

21. "Generic Guidance for an Effective Boric Acid Inspection Program for Pressurized Water Reactors," WCAP-15988-NP Revision 2, June 2012.

22. 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].

23. "Pressurized Water Reactor Owners Group Standard RCS Leakage Action Levels and Responses Guidelines for Pressurized Water Reactors," WCAP-16465-NP Revision 0, September 2006. [NRC ADAMS Accession No. ML070310082].

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Table 1: Summary of Byron Unit 2 CROM Nozzle PWSCC Indications Elevation of Remaining Depth- Upper Tip Vertical Primary to- Aspect Above As- Ligament to Outage Outage Nozzle Uphill/ Flaw Depth, a Thick., Length Ratio, Built Weld Leakage Detected Number Pen.# OD/ID Downhill Orient. (in.) alt 2c (in.) 2c/a Toe (in.) (in.)

Spring B2R13 68 OD Downhill Axial 0.304 47% 0.60 2.0 0.50 1.22 2007 Fall B2R18 6 OD Downhill Axial 0.222 34% 0.52 2.3 0.60 0.96 2014

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Table 2: Summary of Byron Unit 2 RPVHPN NOE Since Spring 2007 Volumetric Direct Visual Visual Mode 3 Examination and Leak Examination Walkdowns and Examination of Outage Outage Path Assessment per (VE) per VT-2 per ASME Relevant PT of Embedded Flaw Previously Season Number N-729-x N-729-x Section XI Indications Repairs Repaired Nozzles Spring UT indication , Performed on B2R13 Performed Performed Performed N/A 2007 penetration 68 penetration 68 Fall UT, PT of B2R14 Performed Performed Performed - -

2008 penetration 68 Spring Performed on UT, PT of B2R15 Performed Performed - -

2010 penetration 68 only penetration 68 Fall UT, PT of B2R16 Performed Performed Performed - -

2011 penetration 68 Spring Performed on UT, PT of 2013 B2R17 penetration 68 only Performed Performed - - penetration 68 Fall UT indication, Performed on UT, PT of B2R18 Performed Performed Performed 2014 penetration 6 penetration 6 penetration 68 UT of penetrations Spring 2016 B2R19 Performed Performed Performed - - 6 and 68 PT of penetration 6 UT of penetration Fall Performed on 9 nozzles 68 B2R20 Performed Performed - -

2017 being re-peened PT of penetrations 6 and 68 Spring 2019 B2R21 - Performed Performed - - -

Fall Deferral to B2R23 PT of penetration B2R22 Scheduled Scheduled N/A N/A 2020 Requested Herein 6 and 68 scheduled Spring 2022 B2R23 Proposed Alternative Scheduled Scheduled N/A N/A -

Note: Direct visual examinations were also performed in B2R10 (Fall 2002) and B2R12 (Fall 2005).

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Table 3: Results for PWSCC Growth of Hypothetical 12% Through-Wall Axial Flaws in Byron Unit 2 CROM Nozzle Calendar years from alt= 12% until leak (See Note 2)

Nozzle Incidence ID flaw on ID flaw on OD flaw on OD flaw on Anale (0 ) uphill side downhill side uphill side downhill side 0.0 9.1 11.0 11.0 25.4 8.6 9.1 13.6 42.8 7.2 See Note 1 9.6 17.5 43.8 7.1 9.6 18.4 47.0 6.9 10.0 20.6 Note 1: For axially oriented ID flaws on the inside surface of the nozzle, the most rapid growth is predicted on the uphill side of the J-groove weld for a flaw located at the weld . Thus, results for ID flaws on the downhill side are not presented in AM-2007-011 [11].

Note 2: Results shown conservatively assume a 98% capacity factor.

ATTACHMENT 2 Dominion Engineering, Inc. Technical Note TN-4069-00-02 Revision 1 July 2020

~

Dominion En~ineeri ":Y TECHNICAL NOTE Experience for Unmitigated CROM Nozzles in U.S.

PWRs Evaluated for Margin Against Leakage Considering Additional PWSCC Growth if Indications Had Remained in Service TN-4069-00-02 Revision 1 July 2020 Principal Investigators M. Burkardt G. White K. Fuhr A. Simon Prepared for Exelon Generation 200 Exelon Way Kennett Square, PA 19348 12100 Sunrise Valley Drive, Suite 220

• Reston, VA 20191

• PH 703.657.7300

• FX 703.657.7301

Dominion fn~ineerin~. Inc. TN-4069-00-02, Rev. 1 RECORD OF REVISIONS Prepared by Checked by Reviewed by Approved by Rev. Description Date Date Date Date M. Burkard! K. J. Fuhr G. A. White G. A. White 0 Original Issue 8/24/2018 8/24/2018 8/24/2018 8/24/2018 Senior Engineer Senior Engineer Principal Engineer Principal Engineer I Revised analysis to 11. 8-l,_,/- -IJ/~Jff} Cf.P. 0;1~ ({.Pr.Ori~

extrapolate growth to N+4 refueling outage

~/IS/ &.Ol.O (!)1/15/afj ~ \1t}'YJ1P -:r J5Jioio J

M. Burkard! A. J. Simon G. A. White G. A. White Senior Engineer Engineer Principal Engineer Principal Engineer The last revision number to reflect any changes for each section of the technical note is shown in the Table of Contents. Paragraphs including changes made in the latest revision, except for Rev.

0 and revisions which change the technical note in its entirety, are indicated by a double line in the right hand margin as shown here. The last revision numbers to reflect any changes for tables and figures are shown in the List of Tables and the List of Figures.

ii

Dominion fn~ineerin~, Inc. TN-4069-00-02, Rev. 1 CONTENTS Last Rev.

Page Mod.

INTRODUCTION ....................................................................................................................1 2 EXTENSION OF EXISTING PUBLISHED RES ULTS ..................................................................... . 1 3 RES ULTS AND CONCLUSION .................................................................................................3 4 REFERENCES ........... .... ............. .... ... .. .............. .... .......... .... .. ..... .. .... .. ... ...... .. ....... .. .............. 3 0 iii

Dominion fn~ineerin~, Inc. TN-4069-00-02, Rev. 1 LIST OF FIGURES Last Rev.

Page Mod.

Figure 1. Tcold PWSCC Indication Remaining Ligament Adjusted to Hypothetical Four-Cycle Examination Frequency ............................................................................. .4 Figure 2. Nozzle Schematic Illustrating the Remaining Ligament to Leakage for OD PWSCC ................................................................................................................4 0 Figure 3. Tcold PWSCC Indication Sizes Adjusted to Hypothetical Four-Cycle Examination Frequency ........................................................................................5 iv

Dominion fn~ineerin~, Inc. TN-4069-00-02, Rev. 1 1 INTRODUCTION A 2016 PVP conference paper [ 1] evaluated in detail the primary water stress corrosion cracking (PWSCC) indications detected in 25 reactor pressure vessel head penetration nozzles (RPVHPNs) in Tcold heads by that time, all in the area of the toe of the J-groove weld on the nozzle OD. All of these PWSCC indications, including those affecting the Byron Station Unit 2 head, were detected in Tcold heads having nozzle material supplied by B&W Tubular Products.

Figure 1, which is a modified version of Figure 7 from the PVP paper, illustrates the margin against growth upward to the nozzle annulus and against consequential leakage that would be expected with a 6.0-year follow up examination interval, i.e., four cycles for units with nominal 18-month fuel cycles. Figure 2 illustrates how the remaining ligament above the axial flaw quantifies the remaining margin against leakage.

Revision 1 of this Technical Note considers the implications of a reexamination periodicity of four 18-month cycles (6.0 calendar years), instead of the three 18-month cycles (4.5 calendar years) assessed by Revision 0.

2 EXTENSION OF EXISTING PUBLISHED RESULTS Using the same approach described in the PVP paper [1], Figure 1 was developed by plotting the nondestructive examination (NDE) dimensional data for PWSCC indications detected in heads operating at Tcold and adjusting the reported indications for a hypothetical examination interval of four 18-month cycles, or 6.0 years of calendar time. Figure 1 shows all 44 PWSCC nozzle tube indications in Tcold heads in the U.S. that have been sized using UT to date, including five indications in four CRDM nozzles at one PWR detected subsequent to the PVP paper. 1 As of July 2020, no additional indications have been detected in U.S. Tcold heads since Rev. 0 of this Technical Note (August 2018). A total of 16 flaws that were detected in less than 6.0 years since the prior examination are extrapolated to the size that would have been expected had each flaw not been detected until the fourth refueling outage (i.e., 6.0 years) after the prior examination.

1 Rejectable planar indications were detected in the nozzle material of three penetrations at tack welds attaching the guide funnels to the end of the nozzle during the spring 2017 pre-peening examination at one unit [2]. These are not included because they are not associated with the region of elevated stress generated by the J-groove weld and were corrected through grinding.

Dominion fn~ineerin~, Inc. TN-4069-00-02, Rev. 1 For context, the vertical weld height at the nozzle OD (i.e., between the bottom/toe and the top/annulus in Figure 2), as determined through UT examination, ranges from 29 to 62 mm (1.16 to 2.44 in.) at the locations of the extrapolated indications. The initial vertical elevation of the top of the indication relative to the bottom/toe of the weld varies from -10 to + 15 mm (-0.41 to

+0.60 in.), where a positive elevation applies when the top of the indication is above the bottom/toe of the weld.

The simulated growth for the extrapolated flaws is based on the assumption that the initial flaw depth was at the limit of detectability at the previous examination (for UT, assumed to be 10-15% through the nozzle wall thickness based on the range of flaw depths included in UT Performance Demonstration Initiative mockups). Alternatively, in some cases, the PWSCC indication was found to correspond to an indication reported during a previous outage but dispositioned as not service-related. In these cases, the depth of the earlier indication is applied to project the additional growth. An additional 36 or 54 months of service is projected for each extrapolated flaw, depending on whether the flaw was actually detected after two or after one 18-month cycle, respectively, since the previous UT examination. Similar to Figure 6 of the PVP paper [ 1], Figure 3 illustrates adjustment of the detected depth and length for each extrapolated PWSCC indication for the additional expected growth had detection been delayed.

The crack growth calculation considers stress intensity factors calculated both at the deepest and surface points on a semi-elliptical crack front. Crack growth is extrapolated forward in time by conservatively assuming a constant driving stress of 70 ksi (483 MPa) and using a crack growth rate percentile for the effect of material variability reflecting the elapsed time for growth from the assumed initial flaw depth to the detected depth, resulting in crack growth rate percentiles as high as the 92nd percentile.

Of the 16 flaws for which growth is extrapolated, six flaws are modeled to grow in depth through the nozzle wall thickness to penetrate the nozzle ID during the 6.0-year interval. These flaws are conservatively modeled to transition instantaneously to an idealized through-wall flaw having the same total length as the semi-elliptical flaw at the point of penetration to the nozzle ID.

Penetration to the nozzle ID does not affect leak tightness as growth to the nozzle annulus is necessary for pressure boundary leakage to occur. This fact is illustrated by Figure 2. The stress intensity factor for the idealized through-wall flaw is calculated per a standard published solution for an axial idealized through-wall flaw in a cylindrical pipe [3]. Figure 1 and Figure 3 each identify the point at which each of these six flaws is modeled to penetrate to the nozzle ID.

Of the six flaws which penetrate to the nozzle ID, one subsequently grows in length to beyond the top of the weld (illustrated in Figure 2). This specific flaw is modeled to grow at the 92nd 2

Dominion fn~ineerin~, Inc. TN-4069-00-02, Rev. 1 crack growth rate percentile. The flaw was detected after one cycle, so three additional cycles of growth were modeled for the extrapolation to a 6.0-year inspection interval. This flaw is modeled to result in leakage after about 5.4 calendar years of growth (i.e., about 0.6 year before the modeled N+4 exam would be performed).

3 RESULTS AND CONCLUSION Figure 1 shows the remaining vertical ligament to leakage (as defined in Figure 2) for the 44 detected indications, as well as the results for extrapolated growth of 16 indications detected in less than four 18-month cycles (6.0 calendar years). Figure 1 shows that for all but three of the 44 indications, a large margin against leakage (i.e., remaining vertical distance from the crack front to the nozzle annulus is greater than the distance equivalent to the nozzle tube thickness) would have been expected if the volumetric examinations had been performed every four 18-month cycles (6.0 calendar years). Furthermore, only one of the 44 total indications detected in a U.S. Tcotd head is modeled to result in leakage per the methodology applied to consider the implications of an examination every four 18-month cycles. These results show that the likelihood of leakage due to PWSCC affecting the nozzle tube base metal remains low assuming an examination frequency of once every four 18-month cycles (6.0 calendar years) for heads operating at Tcotd and with nozzles fabricated with nozzle material supplied by B&W Tubular Products.

4 REFERENCES

1. G. White, K. Fuhr, M. Burkardt, and C. Harrington, "Deterministic Technical Basis for Re-Examination Interval of Every Second Refueling Outage for PWR Reactor Vessel Heads Operating at Tcotct with Previously Detected PWSCC," Proceedings of the ASME 2016 Pressure Vessels & Piping Conference, ASME, PVP2016-64032.

2. U.S. NRC, "Safety Evaluation by the Office of Nuclear Reactor Regulation, Relief Request I4R-15 Regarding Examination of Reactor Pressure Vessel Head Penetration Nozzles, Exelon Generation Company, LLC, Byron Station, Unit 1, Docket No. 50-454,"

dated January 10, 2018. [NRC ADAMS Accession No. MLl 7325B571]

3. S. Marie, et al., "French RSE-M and RCC-MR code appendices for flaw analysis:

Presentation of the fracture parameters calculation - Part III: Cracked Pipes,"

International Journal of Pressure Vessels and Piping, 84, pp. 614-658, 2007.

3

Dominion fn~ineerin~, Inc. TN-4069-00-02, Rev. 1 125 -..===============================::;---------,----------, ..

+ Actual Size at Detection after One Cycle C Extrapolation for Additional Three Cycles Vertical Ligament ..**... ***

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• Actual Size at Detection after Two Cycles Ii Extrapolation for Additional Two Cycles Equal to Nozzle 100 XActual Size at Detection after Four Cycles I ,..-"

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Figure 1. Tcold PWSCC Indication Remaining Ligament Adjusted to Hypothetical Four-Cycle Examination Frequency R V Head (Low Alloy Steel) n Top of Weld Nozzle (A lloy600)

Cladding Remaining Ligam ent (Stainless Steel)-__, ~~

to L eakage Butteni1g (Alloy 821182)

Axial Length of Flaw a

Bottom of W eld Figure 2. Nozzle Schematic Illustrating the Remaining Ligament to Leakage for OD PWSCC 4

Dominion fn~ineerin~, Inc. TN-4069-00-02, Rev. 1 100

• Actua l Size at Detection after One Cycle C Extrapolation for Additional Three Cycles

• Actua l Size at Detection after Two Cycles O ne flaw continues A Extrapolation for Additional Two Cycles to grow in length XActua l Size at Detection after Four Cycles to228 mm 75 I-

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Figure 3. Tcold PWSCC Indication Sizes Adjusted to Hypothetical Four-Cycle Examination Frequency 5