View source for RS-21-008, 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)

{{Adams
| number = ML21025A417
| issue date = 01/25/2021
| title = 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)
| author name = Murray D
| author affiliation = Exelon Generation Co, LLC
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000456
| license number = NPF-072
| contact person = 
| case reference number = RS-21-008
| document type = Code Relief or Alternative, Letter type:RS
| page count = 35
}}

=Text=
{{#Wiki_filter:4300 Winfield Road Warrenville, IL 60555 Exelon Generation                                           630 657 2000 Office RS-21-008                                                                                    10 CFR 50.55a January 25, 2021 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Braidwood Station, Unit 1 Renewed Facility Operating License No. NPF-72 NRC Docket No . STN 50-456

==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 Braidwood Station (Braidwood) Unit 1 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 including 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 compl iance 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-13 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 Braidwood Unit 1 RPVHPNs (Item No. B4.60 of Table 4-3 of MRP-335, Revision 3-A) to the next refueling outage (A 1R23) , which is scheduled in fall 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 January 25, 2021 Page 2 Performing this examination in A 1R22 is a hardship due to expected challenges with obtaining and maintaining staffing levels sufficient for the examination in A 1R22. 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 spring 2021 refueling outage (A 1 R22) to the following refueling outage in fall 2022 (A 1R23). Performance of a volumetric examination of the Braidwood Unit 1 RPVHPNs during the spring 2021 (A 1R22) 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 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, to support this request.
EGC requests approval of the proposed alternative by March 26, 2021, which will support the Braidwood Unit 1 spring 2021 refueling outage (A 1R22). The requested approval date is based on ensuring adequate vendor support for A 1R22 scope.
There are no regulatory commitments in this submittal. Should you have any questions concerning this letter, please contact Mr. Phillip A. Henderson at (630) 657-4727.
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Sr. Manager - Licensing Exelon Generation Company, LLC Attachments:
: 1) 10 CFR 50.55a Relief Request 14R-13 for Braidwood Station, Unit 1, 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 - Braidwood Station NRC Project Manager - Braidwood Station Illinois Emergency Management Agency - Division of Nuclear Safety

ATTACHMENT 1 10 CFR 50.55a Relief Request 14R-13 for Braidwood Station, Unit 1 Revision 0

IS/ Program                                                            Unit 1, Fourth Interval 10 CFR 50.55a Relief Request 14R-13 for Braidwood Station, Unit 1 Revision 0 (Page 1 of 22)
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:              Unit 1, Reactor Vessel 1 RC01 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:                      84.20 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: 185313E 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, 2013 Edition [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.

IS/ Program                                                                  Unit 1, Fourth Interval 10 CFR 50.55a Relief Request 14R-13 for Braidwood Station, Unit 1 Revision 0 (Page 2 of 22)
: 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 75 nozzles peened in fall 2016 with the follow-up inspection for the 4 nozzles peened in spring 2018
[4]. The current authorized follow-up inspection for all 79 identified nozzles is in the spring 2021 refueling outage. This corresponds to a nominal 4.5 calendar years since

IS/ Program                                                                 Unit 1, Fourth Interval 10 CFR 50.55a Relief Request 14R-13 for Braidwood Station, Unit 1 Revision 0 (Page 3 of 22) the last volumetric examination for the 75 nozzles peened in fall 2016 (N+3 follow-up) and a nominal 3.0 calendar years since the last volumetric examination for the 4 nozzles peened in spring 2018 (N+2 follow-up).
The inspection frequency requirements for Item B4.60 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 spring 2021 (A 1R22) 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 B4.50 [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 spring 2021 (A 1R22) refueling outage.
3.2. Requirements Specific to Previously Repaired Penetrations In accordance with the NRC Safety Evaluation [5] of the Relief Request [6] addressing the embedded flaw repair method as applied to the Braidwood Unit 1 head, periodic ultrasonic testing (UT) and surface examinations have been performed of the previously repaired penetration 69. The UT examinations are performed from the nozzle ID to interrogate the nozzle tube wall in accordance with the same volumetric examination requirements that apply for unrepaired nozzles. The requirement for periodic surface examinations of the weld overlay surface has been satisfied through surface dye penetrant testing (PT). Once PT examination results have been verified to be acceptable for two consecutive refueling outages after the original installation of the embedded flaw repair, the ISi interval for the PT examination is every other refueling outage.
No indications attributed to PWSCC were detected during these examinations of the single repaired penetration. Rejectable rounded indications in the embedded flaw repair weld of penetration 69 were detected by PT in fall 2013 (A 1 R 17), spring 2015 (A 1R 18),
and fall 2016 (A1R19) (References [7], [8], and [9]). These indications were determined to be mechanical discontinuities I minor subsurface voids opening up to the weld surface due to thermal and/or pressure stresses during plant operation. The indications were either removed or reduced to an acceptable size, followed by weld repair as necessary, to restore the embedded flaw repair. Subsequent PT examinations of penetration 69 in spring 2018 (A 1R20) and fall 2019 (A 1 R21) detected no rejectable indications. Thus, the next required PT examination for penetration 69 is scheduled for fall 2022 (A 1R23),
satisfying the schedule of the NRC Safety Evaluation [5] . Hence, no relief is sought with regard to the timing of the PT examination requirement for the single repaired penetration at Braidwood Unit 1. In the same manner as for the unrepaired penetrations, under the proposed alternative the UT examination of the repaired penetration 69 would be deferred to the fall 2022 (A 1R23) refueling outage.

IS/ Program                                                                   Unit 1, Fourth Interval 10 CFR 50.55a Relief Request 14R-13 for Braidwood Station, Unit 1 Revision 0 (Page 4 of 22)
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 penetration 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 most recently on January 7, 2021. 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.
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, and avoiding crowds 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 January 22, 2021, all counties in the state of Illinois are currently in Phase 4 of the plan, with resurgence mitigations effective in Will county where Braidwood is located. This phase requires gatherings are limited and it requires face coverings and social distancing. To reach the fifth and final phase, a vaccine, highly effective treatment, or herd immunity would need to be established. Although the Food and Drug Administration has now granted emergency use authorization to two COVID-19 vaccines, vaccine supply is currently limited, with available doses being prioritized for critical populations. This hardship is considered to apply until the fifth and final phase is reached.
EGC has implemented practices beyond the CDC recommendations. These include self-screening questions of all workers prior to reporting to work, temperature monitoring, and questioning of all people prior to entering a company site or property and maximizing remote enabled employees. Braidwood has performed a detailed review of items for the spring 2021 refueling outage to minimize the number of people who have to travel to the site. Items that are not necessary for nuclear safety or reliable generation are being deferred from the scope of the spring 2021 refueling outage.
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 especially high numbers of active COVID-19 cases. The CDC has determined that COVID-19 poses a serious public health risk, with reports of community spread of COVID-19 throughout the country.

IS/ Program                                                                  Unit 1, Fourth Interval 10 CFR 50.55a Relief Request 14R-13 for Braidwood Station, Unit 1 Revision 0 (Page 5 of 22)
Twenty-four individuals from across the U.S. are required to mobilize on-site to support UT examination of the Braidwood Unit 1 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.
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 Braidwood Unit 1. 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 fall 2022. This alternative corresponds to a nominal 6.0 calendar years since the last volumetric examination for the 75 nozzles peened in fall 2016 (N+4 follow-up) and a nominal 4.5 calendar years since the last volumetric examination for the 4 nozzles peened in spring 2018 (N+3 follow-up). An interval of 6.0 calendar years is still less than the limit of RIY = 2.25, which corresponds to 7.0 calendar years for the Braidwood Unit 1 operating temperature of 556.71 &deg;F, conservatively assuming a 100% capacity factor. The limit of RIY = 2.25 (7.0 calendar years) would be acceptable under the requirements of 10 CFR

IS/ Program                                                                Unit 1, Fourth Interval 10 CFR 50.55a Relief Request 14R-13 for Braidwood Station, Unit 1 Revision 0 (Page 6 of 22) 50.55a(g)(6)(ii)(D) for the Braidwood Unit 1 head if PWSCC affecting the head nozzles was not previously detected.
During the initial peening application in fall 2016 (A1R19), 4 RPVHPNs at Braidwood Unit 1 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 [10]. Three control rod drive mechanism (CROM) penetrations and the vent line penetration were affected (4 total penetrations). EGC successfully peened the inner diameter (ID) or outer diameter (OD) surfaces of these 4 penetrations (as needed) during the spring 2018 refueling outage (A 1R20) to satisfy MRP-335, Revision 3-A requirements . Therefore, all 79 identified nozzles at Braidwood Unit 1 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 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 Braidwood Unit 1 RPVHPNs during the spring 2021 (A1 R22) 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 Braidwood Unit 1 RPVHPNs (Item No. B4.60 of Table 4-3 of MRP-335, Revision 3-A) to the next refueling outage (A 1R23), which is scheduled in fall 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.
Braidwood Unit 1 operating cycles are approximately 18 months in duration.
Performing this examination in A 1R22 is a hardship due to expected challenges with obtaining and maintaining staffing levels sufficient for the examination in A 1R22.
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 the previous indication of cracking that was repaired, direct VEs for evidence of leakage, and online leak detection capability. Based on these

IS/ Program                                                                 Unit 1, Fourth Interval 10 CFR 50.55a Relief Request 14R-13 for Braidwood Station, Unit 1 Revision 0 (Page 7 of 22) factors, EGC has identified performance of the volumetric examinations of the Braidwood Unit 1 RPVHPNs in A 1R22 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.
Operating experience demonstrates the susceptibility of Alloy 600 nozzles to PWSCC.
An indication of PWSCC has been detected during a previous examination affecting one of the RPVHPNs at Braidwood Unit 1 [11]. The affected CROM nozzle was repaired using the embedded flaw repair technique. Table 1 shows size and location information for the repaired PWSCC indication based on the data summary sheets for the UT examinations performed [12]. The indication was located on the nozzle OD, and thus did not originate on the nozzle ID. The indication, detected in 2012, was attributed to PWSCC [11]. A summary of nondestructive examinations performed for the Braidwood Unit 1 RPVHPNs is provided in Table 2, which also shows the future schedule of examinations under the proposed alternative.
The indication of PWSCC was 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 1.68 inches. Further growth in the axial (vertical) direction of at least 1.68 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.

IS/ Program                                                                  Unit 1, Fourth Interval 10 CFR 50.55a Relief Request 14R-13 for Braidwood Station, Unit 1 Revision 0 (Page 8 of 22) 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 [13), determined the amount of time needed for a postulated flaw to result in a potential leak path. This analysis is applicable to the Braidwood Unit 1 RPVHPNs, which shares a common head, penetration, and J-groove weld design configuration with Byron Unit 2. 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 [14), 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.
Weld residual stress analysis results specifically produced for the head, penetration, and J-groove weld design configuration common to the Braidwood and Byron heads, including the effect of normal operating pressure and temperature, were applied. Crack growth results were produced for five RPVHPNs incidence angles (0&deg;, 25.4&deg;, 42.8&deg;,
43.8&deg;, and 4 7&deg;), 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&deg;F was applied in the PWSCC crack growth rate equation. This temperature produces conservatively faster crack growth rates (i.e., a factor of 1.04 faster applying the thermal activation energy of the MRP-55 equation) than the current Braidwood Unit 1 nominal closure head temperature of 556.71&deg;F [Error! Reference source not found.].
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 actual reported PWSCC indication affecting the Braidwood Unit 1 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.

IS/ Program                                                                Unit 1, Fourth Interval 10 CFR 50.55a Relief Request 14R-13 for Braidwood Station, Unit 1 Revision 0 (Page 9 of 22)
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 cold Heads The experience for unmitigated heads in the U.S. operating at reactor cold-leg temperature (Tco1d), including that for Braidwood Unit 1 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 [16] 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 the one affecting Braidwood Unit 1, were detected in Tco1d heads having nozzle material supplied by B&W Tubular Products (B&WTP).
Through an extension of the assessment of plant experience in the PVP paper, the evaluation per TN-4069-00-02, Revision 1 [17], 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.
The evaluation in TN-4069-00-02 , Revision 1 provided in Attachment 2 was previously submitted to N RC [17] for a similar relief request for Byron Station, Unit 2, requesting a one-time deferral of the volumetric examination of the RPVHPNs planned for fall 2020 (B2R22) until the spring 2022 refueling outage (B2R23) [ML20199M304] . The relief request was approved in an NRC Safety Evaluation dated September 9, 2020 [18]. The evaluation of TN-4069-00-02, Revision 1 [17] remains valid as there have been no additional reports of PWSCC affecting RPVHPNs in heads operating at Tcold since this technical note was issued in July 2020. This technical note is directly applicable to Braidwood.
5.5. Previously Repaired Nozzle The nozzle previously repaired using the embedded flaw repair technique is normally examined volumetrically when the other 77 CROM nozzles are volumetrically examined.
Under the proposed alternative, the examination of the previously repaired nozzle will also be deferred until the next refueling outage. This one-time extension of one additional refueling cycle between examinations of the repaired nozzle is justified for the following reasons.

IS/ Program                                                                 Unit 1, Fourth Interval 10 CFR 50.55a Relief Request 14R-13 for Braidwood Station, Unit 1 Revision 0 (Page 10 of 22)
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 Braidwood Unit 1 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 [19] and [20]). As shown in Table 2, the indication detected in 2012 (penetration 69) has been examined volumetrically in three refueling outages since repair. This indication has not shown evidence of growth subsequent to the embedded flaw repair.
As discussed above in Section 3.2, PT examinations of penetration 69 in spring 2018 (A 1R20) and fall 2019 (A 1R21) detected no rejectable indications. In accordance with the NRC Safety Evaluation addressing the use of the embedded flaw repair technique at Braidwood [5], and as shown in Table 2, PT examination of the embedded flaw repair in penetration 69 will be performed in the fall 2022 (A 1R23) refueling outage. Hence, no relief is sought with regard to the timing of the PT examination requirement for the single repaired penetration at Braidwood Unit 1.
With the embedded flaw repair, the only mechanism for sub-critical crack growth is fatigue [19] . 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 flaw repaired in penetration 69 was smaller than this maximum crack size for the embedded flaw repair.
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 [21]. 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 Palisades experience in 2019 [22] 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 A 1R19 and A 1R20 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

IS/ Program                                                                Unit 1, Fourth Interval 10 CFR 50.55a Relief Request 14R-13 for Braidwood Station, Unit 1 Revision 0 (Page 11 of 22) 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), Byron Unit 1 (spring 2017), and Millstone Unit 3 (fall 2020). 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) [23], 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 2022 at Byron Unit 2, spring 2023 at Braidwood Unit 2, and fall 2023 at Millstone 3.
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. The RPVHPNs at Millstone Unit 3 were also peened using the UHPCP process, but were fabricated by Combustion Engineering (CE) using Alloy 600 material supplied by Huntington. Hence, the favorable experience with the Byron Unit 1 post-peening follow-up volumetric examination is relevant to the peened head at Braidwood Unit 1.
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 [21] . Furthermore, example deterministic crack growth calculations summarized in MRP-395 [21] 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 Braidwood Unit 1 operating temperature of 556.71&deg;F, the allowable interval for heads without previously detected PWSCC is limited by the RIY =

IS/ Program                                                                Unit 1, Fourth Interval 10 CFR 50.55a Relief Request 14R-13 for Braidwood Station, Unit 1 Revision 0 (Page 12 of 22) 2.25 requirement. Conservatively assuming a 100% capacity factor, the limiting time
                                      =
interval corresponding to RIY 2.25 is 7.0 calendar years . 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 the Braidwood and Byron heads 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 Braidwood Unit 1 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 A 1R22 refueling outage in spring 2021. 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.
Braidwood Unit 1 implements a boric acid corrosion control program consistent with WCAP-15988, Revision 2 [24] . 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. Industry experience in 2018 [25] 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.

IS/ Program                                                                 Unit 1, Fourth Interval 10 CFR 50.55a Relief Request 14R-13 for Braidwood Station, Unit 1 Revision 0 (Page 13 of 22)
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. Braidwood Unit 1 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 [26] .
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.
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 Braidwood and Byron heads 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 Tco1d. Conservatively, these analyses do not credit the effectiveness of the peening mitigation that has already been performed of the Braidwood Unit 1 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

IS/ Program                                                                 Unit 1, Fourth Interval 10 CFR 50.55a Relief Request 14R-13 for Braidwood Station, Unit 1 Revision 0 (Page 14 of 22) proposed alternative provides reasonable assurance of structural integrity until the next volumetric examination is performed in fall 2022(A1 R23).
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 spring 2021 refueling outage (A 1R22) to the following refueling outage in fall 2022 (A 1R23).
Performance of a volumetric examination of the Braidwood Unit 1 reactor vessel closure head penetrations during the spring 2021 (A 1R22) 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 March 26, 2021, which will support the Braidwood Unit 1 spring 2021 refueling outage (A 1R22). The requested approval date is based on ensuring adequate vendor support for A 1R22 scope.
: 6.      Duration of Proposed Alternative The proposed alternative is requested for the duration of the next Braidwood Unit 1 operating cycle, until the fall 2022 refueling outage (A 1R23 ). After performing the volumetric examination in A 1R23, 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 several previous alternatives due to the hardship in mobilizing inspection vendors during the COVID-19 pandemic and required augmented examinations of RPVHPNs per ASME Code Case N-729-4 or ASME Code Case N-729-6, as summarized in the table below. The relief request [17] approved for Byron Unit 2 on September 9, 2020 [18] addressed the same type of hardship affecting the RPVHPNs at Braidwood Unit 1.

IS/ Program                                                         Unit 1, Fourth Interval 10 CFR 50.55a Relief Request 14R-13 for Braidwood Station, Unit 1 Revision 0 (Page 15 of 22)
NRC ADAMS Accession No.
Plant         Component              Examination                         NRC Safety Relief Request      Evaluation Reactor vessel hot N-770-2, Item A-2, leg nozzle dissimilar Volumetric metal welds Comanche                                N-722-1, Item BMN                                 ML201 OOG562      ML20106F235 Peak 2                                815.80 Visual N-729-4, Item RPVHPNs B4.10 Visual ML20196L828 Comanche                                N-729-4, Item RPVHPNs                                                  ML20224A055 Peak 1                                B4.30 Visual     ML20218A863 (supplement)
N-729-6, Item Byron 2         RPVHPNs            B4.20 Volumetric I  ML20199M304       ML20245E506 Surface
: 8.      Acronyms ADAMS        Agencywide Documents Access and Management System ASME         American Society of Mechanical Engineers BMN           Bottom Mounted Nozzle BPV           Boiler and Pressure Vessel B&WTP         Babcock & Wilcox Tubular Products CDC           Center for Disease Control and Prevention CE            Combustion Engineering CFR           Code of Federal Regulations COVID-19      Coronavirus Disease of 2019 CRDM          Control Rod Drive Mechanism EDY           Effective Degradation Year EGC           Exelon Generation Company, LLC EPRI          Electric Power Research Institute ID            Inner Diameter ISi           lnservice Inspection LCO           Limiting Condition for Operation MRP           [EPRI] Materials Reliability Program NDE           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]

IS/ Program                                                               Unit 1, Fourth Interval 10 CFR 50.55a Relief Request 14R-13 for Braidwood Station, Unit 1 Revision 0 (Page 16 of 22)
RPV              Reactor Pressure Vessel RPVHPN           Reactor pressure vessel [upper] head penetration nozzle 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," 2013 Edition.
: 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 L. Regner (U.S . Nuclear Regulatory Commission) to B. Hanson (Exelon Generation Company, LLC), "Braidwood Station, Unit 1 - Relief from the Requirements of the American Society of Mechanical Engineers Code (EPID L-2018-LLR-0126)," dated June 5, 2019. Safety Evaluation by the Office of Nuclear Reactor Regulation, Relief Request 14R-08 Regarding Repair and Examination of Reactor Vessel Head Penetration Nozzles. [NRC ADAMS Accession No. ML19155A060].

IS/ Program                                                             Unit 1, Fourth Interval 10 CFR 50.55a Relief Request 14R-13 for Braidwood Station, Unit 1 Revision 0 (Page 17 of 22)
: 5. Letter from L. Regner (U.S . Nuclear Regulatory Commission) to B. Hanson (Exelon Generation Company, LLC), "Braidwood Station, Units 1 and 2 - Relief from the Requirements of the American Society of Mechanical Engineers Code (EPID L-2018-LLR-0131 ),"dated June 5, 2019. Safety Evaluation by the Office of Nuclear Reactor Regulation, Relief Request 14R-10 Regarding Repair and Examination of Reactor Vessel Head Penetrations Nozzles. [N RC ADAMS Accession No. ML19141A020].
: 6. Letter from D. Gullett (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "Proposed Alternative Requirements for the Repair and Examination of Reactor Pressure Vessel Head Penetration Nozzles for the Fourth lnservice Inspection Interval in Accordance with 10 CFR 50.55a(z)(1 ),"dated October 11, 2018. 10 CFR 50.55a Relief Request 14R-10, Alternative Requirements for the Repair of Reactor Pressure Vessel Head Penetration Nozzles in Accordance with 10 CFR 50.55a(z)(1). [NRC ADAMS Accession No. ML18284A445].
: 7. Letter from M. Kanavos (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "Licensee Event Report 2013-002 Indications in Control Rod Drive Mechanism Nozzle due to lmbedded Flaws Opening Up from Thermal and Pressure Stresses During Operation," dated November 13, 2013.
[NRC ADAMS Accession No. ML13318A096].
: 8. Letter from M. Kanavos (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "Licensee Event Report 2015-002 Indication in Control Rod Drive Mechanism Nozzle Weld due to Embedded Flaws Opening Up from Thermal and Pressure Stresses during Operation," dated June 1, 2015. [NRC ADAMS Accession No. ML15153A396].
: 9. Letter from M. Marchionda-Palmer (Exelon Generation Company, LLC) to U.S.
Nuclear Regulatory Commission, "Licensee Event Report 2016-003 Indication in Control Rod Drive Mechanism Nozzle Weld due to Embedded Flaws Opening Up from Thermal and Pressure Stresses during Operation," dated November 23, 2016.
[NRC ADAMS Accession No. ML16328A122].
: 10. Letter from D. M. Gullott (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].
: 11. Letter from D. Enright (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "Licensee Event Report (LER) 2012-002 Reactor Pressure Vessel Head Control Rod Drive Mechanism Penetration Nozzle Weld Indication Attributed to Primary Water Stress Corrosion Cracking," dated June 22, 2012. [NRC ADAMS Accession No. ML12174A227].
: 12. Braidwood Unit 1, Penetration 69 Inspection Data Sheets:

IS/ Program                                                            Unit 1, Fourth Interval 10 CFR 50.55a Relief Request 14R-13 for Braidwood Station, Unit 1 Revision 0 (Page 18 of 22)
: a. CCE-R16-CP02-069-01Y, Ultrasonic Report Data Sheet, dated April 23, 2012.
: b. CCE-R16-CP02-069-01 R,02R repair, Ultrasonic Report Data Sheet, dated May 6, 2012.
: 13. 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].
: 14. Materials Reliability Program (MRP) Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking (PWSCC) of Thick-Wall Alloy 600 Materials (MRP-55) Revision 1, EPRI, Palo Alto, CA: 2002. 1006695.
: 15. Exelon Engineering Change EC629913, Rev. 000, "Braidwood Unit 1 Effective Degradation Years (EDY) and Reinspection Years (RIY) Evaluation Per ASME Code Case N-729-4 Cycle 21," dated December 5, 2019.
: 16. 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.
: 17. Letter from D. Murray (Exelon Generation Company, LLC), "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), "dated July 17, 2020 . Attachment 1: 10 CFR 50.55a Relief Request 14R-17 for Byron Station, Unit 2, Revision 0. Attachment 2:
Dominion Engineering, Inc. Technical Note TN-4069-00-02 , Revision 1, July 2020.
[NRC ADAMS Accession No. ML20199M304].
: 18. Letter from N. Salgado (U.S. Nuclear Regulatory Commission) to B. C. Hanson (Exelon Generation Company, LLC), "Byron Station, Unit No. 2, Relief from the Requirements of the ASME Code [COVID-19] (EPID L-2020-LLR-0098)," dated September 9, 2020. Safety Evaluation by the Office Nuclear Regulation, Relief Request 14R-17 Regarding Alternative Follow-Up Inspections for Reactor Pressure Vessel Head Penetration Nozzles. [NRC ADAMS Accession No. ML20245E506].
: 19. Letter from H. Berkow (U .S. Nuclear Regulatory Commission) to H. 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].
: 20. Letter from R. Pascarelli (U .S. Nuclear Regulatory Commission) to T. 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].

IS/ Program                                                            Unit 1, Fourth Interval 10 CFR 50.55a Relief Request 14R-13 for Braidwood Station, Unit 1 Revision 0 (Page 19 of 22)
: 21. 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. [NRG ADAMS Accession No. ML143078007].
: 22. Letter from J. 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. [NRG ADAMS Accession No. ML19003A239].
: 23. Letter from H. Peterson (U.S. Nuclear Regulatory Commission) to B. Hanson (Exelon Generation Company, LLC), "Byron Station - Integrated Inspection Report 05000454/2020001 and 05000455/2020001," dated April 24, 2020. [NRG ADAMS Accession No. ML20115E528].
: 24. "Generic Guidance for an Effective Boric Acid Inspection Program for Pressurized Water Reactors," WCAP-15988-NP Revision 2, June 2012.
: 25. Letter from A 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.
[NRG ADAMS Accession No. ML18149A126].
: 26. "Pressurized Water Reactor Owners Group Standard RCS Leakage Action Levels and Responses Guidelines for Pressurized Water Reactors," WCAP-16465-NP Revision 0, September 2006. [NRG ADAMS Accession No. ML070310082].

ISi Program                                                                                            Unit 1, Fourth Interval 10 CFR 50.SSa Relief Request 14R-13 for Braidwood Station, Unit 1 Revision 0 (Page 20 of 22)
Table 1       s ummari   0f Braid      d Unit 1 CROM N       le PWSCC lndicaf Elevation of Depth-            Axial           Circ.           Upper Tip     Remaining Primary              to-    Axial   Aspect  Circ. Aspect           Above As-      Vertical Outage    Outage Pen. Nozzle     Uphill/   Flaw   Depth,    Thick., Extent   Ratio, Extent,  Ratio,    %      Built Weld   Ligament to Detected  Number  #    OD/ID     Downhill  Orient. a (in.)     alt   2c (in.)  2cla  2c (in.)  2cla  Circum. Toe (in.)   Leakaae (in.)
Spring A1R16   69    OD       Downhill   Axial  0.216     33.5%    0.60     2.78   0.314    1.45   2.5%        0.40          1.68 2012

ISi Program                                                                                          Unit 1, Fourth Interval 10 CFR 50.55a Relief Request 14R-13 for Braidwood Station, Unit 1 Revision 0 (Page 21 of 22)
Table 2      s           f Braid     d Unit 1 RPVHPN NOE s*     Sorina 2012 Volumetric                           Visual Mode 3 Examination and         Direct Visual     Walkdowns                                       Examination Leak Path           Examination      and VT-2 per   Relevant            PT of         of Previously Outage      Outage     Assessment     per N-      (VE) per      ASME Section Indications of      Embedded            Repaired Season     Number             729-x                N-729-x             XI        PWSCC          Flaw Repairs         Nozzles Spring                                                                         UT indication,   Performed on A1R16           Performed            Performed         Performed                                            N/A 2012                                                                         penetration 69   penetration 69 Fall                                                                                                               UT, PT of A1R17           Performed            Performed         Performed         -                -
2013                                                                                                             penetration 69 Spring                                                                                                               UT, PT of 2015 A1R18           Performed            Performed         Performed         -                -           penetration 69 Fall                                                                                                               UT, PT of A1R19           Performed            Performed         Performed         -                -
2016                                                                                                             penetration 69 Performed on 4 Spring                                                                                                                 PT of A1R20       nozzles being re-        Performed         Performed         -                -
2018                                                                                                             penetration 69 peened.
Fall                                                                                                                 PT of 2019 A1R21                -               Performed         Performed         -                -           penetration 69 Spring                  Deferral to A 1R23 A1R22                                Scheduled         Scheduled        N/A              N/A                 -
2021                   Requested Herein Fall                                                                                                               UT, PT of A1R23      Proposed Alternative      Scheduled         Scheduled        N/A              N/A 2022                                                                                                             penetration 69 Note: Volumetric examination and leak path assessment were also performed in Spring 2006 (A 1R12). Direct visual examinations were also performed in Spring 2003 (A1 R10) and Fall 2007 (A1 R13).

ISi Program                                                                                                 Unit 1, Fourth Interval 10 CFR 50.SSa Relief Request 14R-13 for Braidwood Station, Unit 1 Revision 0 (Page 22 of 22)
Table 3:     Results for PWSCC Growth of Hypothetical 12% Through-Wall Axial Flaws in Braidwood and Byron CROM Nozzles Calendar years from alt = 12% until leak (See Note 2)
Nozzle Incidence         ID flaw on             ID flaw on           OD flaw on              OD flaw on Angle ( 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 [13].
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 En~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 I
Senior Engineer Senior Engineer   Principal Engineer Principal Engineer I     Revised analysis to          11. TJ-l_,,/- -/J;~Jjf) (f.P . cr,IJ.t.. &-.t\-'.~~
extrapolate growth to N+4     ://IS/ &.o&.O      fJ1/15/W            ~ 11~)11)1SJ 1J15}'lD-u>
refueling outage                 M. Burkardt     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                                 ..**... ***
Tube Thickness .
* 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      ,..-"
::KActual Size at Detection at First Inservice Exam
                  - Modeled Size U on Penetratin Nozzle ID                                                                                       .........                  ....a
                                                                                                                                          ... **         I            *.. **
                                                                                                                                .. ...***                               Flaw continues to
                                                                                                                      ,..-*                              I grow in length
                                                                                                                                       ..a ......
                                                                                                               ,.. ..                                    1 I
                                                                                                     ,... ..
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                                                                                                                                                              .a . . _________,
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                                *         .....             xxr;iM-* * . .... x ~                                                    x 0 +--'---'-___.,--''--+--_.___.__,____.,--1_..___.___.__,_--+---'-'---'--'--+---'---''--''---'---+--'--'--'---''--i 60                      50                            40                                  30                                20                             10                            0 Estimated Remaining Ligament to Leakage (mm)
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-
:i<Actua l Size at Detection at First Inservice Exam
                         - Modeled Size Upon Penetrating Nozzle ID
          ......c                                                                    Flaws grow to penetrate nozz le
            =
e.11 QJ ID (nominal 16 mm depth) and
          ~        50                                                                transitions to assumed idealized
          .:.::                                                                              through-wall flaw Col x
            ~
u
          ~
25 x
Ill 0
0          2              4               6            8      10              12            14        16 Maxi mum Depth (mm)
Figure 3.         Tcold PWSCC Indication Sizes Adjusted to Hypothetical Four-Cycle Examination Frequency 5}}